Contents
Results of Operation and Utilization of the Dalat Nuclear Research Reactor
Nguyen Nhi Dien, Luong Ba Vien, Le Vinh Vinh, Duong Van Dong, Nguyen Xuan Hai,
Pham Ngoc Son, Cao Dong Vu..............................................................................................1
Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor
Luong Ba Vien, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien Cuong........................10
Conceptual Nuclear Designof a 20 MW Multipurpose Research Reactor
Nguyen Nhi Dien, Huynh Ton Nghiem, Le Vinh Vinh, Vo Doan Hai Dang, Seo Chulgyo,
Park Cheol, Kim Hak Sung..................................................................................................26
Some Main Results of Commissioning of The Dalat Research Reactor with Low Enriched Fuel
Nguyen Nhi Dien, Luong Ba Vien, Pham Van Lam, Le Vinh Vinh, Huynh Ton Nghiem,
Nguyen Kien Cuong, Nguyen Minh Tuan, Nguyen Manh Hung, Pham Quang Huy,
Tran Quoc Duong, Vo Doan Hai Dang, Trang Cao Su, Tran Tri Vien...............................36
Production of Radioisotopes and Radiopharmaceuticals at the Dalat Nuclear Research Reactor
Duong Van Dong, Pham Ngoc Dien, Bui Van Cuong, Mai Phuoc Tho, Nguyen Thi Thu,
Vo Thi Cam Hoa..................................................................................................................46
The gamma two-step cascade method at Dalat Nuclear Research Reactor
Vuong Huu Tan, Pham Dinh Khang, Nguyen Nhi Dien, Nguyen Xuan Hai, Tran Tuan
Anh, Ho Huu Thang, Pham Ngoc Son, Mangengo Lumengano..........................................57
Progress of Filtered Neutron Beams Development and Applications at the Horizontal
Channels No.2 and No.4 of Dalat Nuclear Research Reactor
Vuong Huu Tan, Pham Ngoc Son, Nguyen Nhi Dien, Tran Tuan Anh, Nguyen Xuan Hai….62
Characterization of neutron spectrum parameters at irradiation channels for neutron
activation analysis after full conversion of the Dalat nuclear research reactor to low
enriched uranium fuel
C.D. Vu, T.Q. Thien, H.V. Doanh, P.D. Quyet, T.T.T. Anh, N.N. Dien.............................70
Some results of NAA collaborative study in white rice performed at Dalat Nuclear Research Institute
T.Q. Thien*, C.D. Vu, H.V. Doanh, N.T. Sy........................................................................76
A new rapid neutron activation analysis system at Dalat nuclear research reactor
H.V. Doanh, C.D. Vu, T.Q. Thien, P.N. Son, N.T. Sy, N. Giang, N.N. Dien.....................84
Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 01-09
Results of Operation and Utilization of
the Dalat Nuclear Research Reactor
Nguyen Nhi Dien, Luong Ba Vien, Le Vinh Vinh, Duong Van Dong,
Nguyen Xuan Hai, Pham Ngoc Son, Cao Dong Vu
Nuclear Research Institute (NRI), Vietnam Atomic Energy Institute (VINATOM)
01 Nguyen Tu Luc, Dalat, Vietnam
(Received 5 March 2014, accepted 26 March 2014)
Abstract: The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW was
reconstructed and upgraded from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The
renovated reactor was put into operation on 20th March 1984. It was designed for the purposes of
radioisotope production (RI), neutron activation analysis (NAA), basic and applied researches, and
nuclear education and training. During the last 30 years of operation, the DNRR was efficiently
utilized for producing many kinds of radioisotopes and radiopharmaceuticals used in nuclear medicine
centers and other users in industry, agriculture, hydrology and scientific research; developing a
combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods
for quantitative analysis of about 70 elements and constituents in various samples; carrying out
experiments on the reactor horizontal beam tubes for nuclear data measurement, neutron radiography
and nuclear structure study; and establishing nuclear training and education programs for human
resource development. This paper presents the results of operation and utilization of the DNRR. In
addition, some main reactor renovation projects carried out during the last 10 years are also mentioned
in the paper.
Keywords: DNRR, HEU, LEU, RRRFR, RERTR, WWR-M2, NAA, INAA, RNAA, PGNAA.
I. INTRODUCTION
The DNRR is a 500-kW pool-type
reactor loaded with the Soviet WWR-M2 fuel
assemblies. It was reconstructed and upgraded
from the USA 250-kW TRIGA Mark-II reactor
built in early 1960s. The first criticality of the
renovated reactor was on the 1st November
1983 and its regular operation at nominal
power of 500 kW has been since March 1984.
The first fresh core was loaded with 88 fuel
assemblies enriched to 36% (HEU- Highly
Enriched Uranium).
In the framework of the program on
Russian Research Reactor Fuel Return
(RRRFR) and the program on Reduced
Enrichment for Research and Test Reactor
(RERTR), the DNRR core was partly
converted from HEU to Low Enriched
Uranium (LEU) with 19.75% enrichment in
September 2007. Then, the full core conversion
of the reactor to LEU fuel was also performed
from 24th November 2011 to 13th January 2012.
Recently, the DNRR has been operated with a
core configuration loaded with 92 WWR-M2
LEU fuel assemblies and 12 beryllium rods
around the neutron trap.
The reactor is used as a neutron source
for the purposes of radioisotopes production,
neutron activation analysis, basic and applied
researches and training. As a unique research
reactor in Vietnam, the DNRR has been
playing an important role in the research and
development of nuclear technique applications
as well as in nuclear power programme
development of the country. Safe operation and
©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR
effective utilization of the reactor expected at
least to the year 2030 are a long-term objective
of the DNRR. For this reason, so far the
Government has strongly supported for many
specific projects in order to upgrade the facility
and improve its operation and utilization.
some main reactor renovation projects carried
out during the last 10 years are also
mentioned, too.
The results of operation and utilization
of the DNRR are presented in this paper and
Main specifications of the DNRR are
shown in Table I.
II. BRIEF REACTOR DESCRIPTION
AND IT’S OPERATION
Table I. Specifications of the DNRR.
Reactor type
Swimming pool TRIGA Mark II, modified to Russian
type of IVV-9
Nominal thermal power
500 kW, steady state
Coolant and moderator
Light water
Core cooling mechanism
Natural convection
Reflector
Beryllium and graphite
Fuel types
WWR-M2, dispersed UO2-Al with 19.75% enrichment,
aluminium cladding
Number of control rods
7 (2 safety rods, 4 shim rods, 1 regulating rod)
Materials of control rods
B4C for safety and shim rods, stainless steel for
automatic regulating rod
Neutron measuring channels
6 combined in 3 housings with 1 CFC and 1 CIC each
Vertical irradiation channels
4 (neutron trap, 1 wet channel, 2 dry channels) and 40
holes at the rotary rack
Horizontal beam-ports
4 (1 tangential - No #3 and 3 radial - No #1, #2, #4)
Thermal column
1
Maximum thermal neutron
flux
2.1x1013 n.cm-2.s-1 (in the neutron trap at core center)
Main utilizations
RI, NAA, PGNAA, NR, basic and applied researches,
nuclear training
The reactor consists of a cylindrical
aluminum tank 6.26 m high and 1.98 m in
diameter of the original TRIGA Mark II
reactor. The reactor core, positioned inside the
graphite reflector, is suspended from above by
an inner cylindrical extracting well so as to
increase the cooling efficiency for copping
with higher thermal power of the reactor. The
vertical section view of the reactor is shown in
Fig. 1 and the cross-section view of the reactor
core is shown in Fig. 2.
2
NGUYEN NHI DIEN et al.
~ 2000 mm
Rotating top lid
SR
Pool tank
Sh
Upper
cylindrical
shell
R
Sh
~ 6840 mm
Extracting
well
RgR
Concrete
shielding
Spent fuel
storage tank
Thermal
column door
A
Core
Graphite
Sh
Sh
Door plug
SR
(ex bulk-shielding
experimental tank)
Fig. 1. Vertical section view of the DNRR reactor.
The reactor core has a cylindrical shape
with a height of 60 cm and a diameter of 44.2
cm, that is constituted of 92 LEU fuel
assemblies, 7 control rods, a neutron trap at the
core center and 3 in-core irradiation facilities.
Fig. 2. Cross-section view of the core with 92
fuel assemblies.
At present, the DNRR is operated
mainly in continuous runs of 100 or 130 hrs,
once every 3-4 weeks, for radioisotope
production, neutron activation analyses, basic
and applied researches and training. The
remaining time between two consecutive runs
is devoted to maintenance activities and also to
physics experiments. From the first start-up to
the end of 2013, it totaled about 37,800 hrs of
operation, namely a yearly average of 1300
hrs, and the total energy released was about
760 MWd. Detailed yearly operation time of
the DNRR is given in Fig. 3.
Type of fuel with a 235U enrichment of
19.75% of UO2+Al covered by aluminum
cladding is used. Each LEU fuel assembly
contains about 50.5 g of U-235, distributed on
three coaxial fuel tubes, of which the outermost
one is hexagonal shaped and the two inner ones
are circular.
Fig. 3. Yearly operation time of the DNRR.
3
disease therapeutics and 32P in injectable
solution, 99mTc generator of gel type by 98Mo(n,
)99Mo reaction have regularly been produced
and supplied once every 2 weeks. Other
radioisotopes as 51Cr, 60Co, 65Zn, 64Cu, 24Na,
etc. were also produced in a small amount
when requested. 53Sm in solution form was
ready for labelling. Totally, about 5,500 Ci of
radioisotopes have been produced and supplied
to medical uses so far with a yearly average in
the last 5 years of about 400 Ci (Fig. 4)
correspondingly.
So far, the reactor has proved to be safe
and reliable, as it has never suffered from any
incident, which significantly affected the
environment, and annual operation schedules
have been rigorously respected. The
unscheduled shutdowns were mainly due to
unstable working of the city electric network.
III. MAIN RESULTS OF REACTOR
UTILIZATION
A. Radioisotopes and radiopharmaceuticals
production
In order to support the application of
Tc, 113mIn and 53Sm radioisotopes in clinical
diagnosis and therapeutics, the preparation of
radio-pharmaceuticals in Kit form for labelling
was carried out in parallel with the
development of 99mTc generator systems.
About 17 labeled compounds kits have been
regularly prepared and supplied including
Phytate, Gluconate, Pyrophosphate, Citrate,
DMSA, HIDA, DTPA, Macroaggregated HSA
and EHDP, etc.. The annual production rate is
about 1000 bottles for each Kit which is
equivalent to 5000 diagnostic doses.
Research
on
radioisotope
and
radiopharmaceutical
production
serving
nuclear medicine and other users such as
industry, agriculture, hydrology, scientific
research, etc. is oriented towards efficient use
of the reactor. Via such research a variety of
products including 131I, 32P applicators and
solutions, 99mTc generators, etc. were produced.
99m
For medicine applications, radioisotopes
and radiopharmaceuticals have been delivered
to 25 hospitals throughout the country. The
main radioisotopes, such as 131I in NaI solution
and 131I capsule type, 32P applicators for skin
Fig. 4. Total radioactivity of RI produced annually at Dalat Nuclear Research
Institute for medicine.
4
Other applications of radioisotopes
produced at the DNRR are radiotracer
technique in sediment studies, oil exploitation,
chemical industry, biology, agriculture and
hydrology. Some main products are 46Sc, 192Ir,
198
Au, 131I, 140La, etc. In addition, some small
sources of 192Ir and 60Co with low radioactivity
have also been produced for industry
applications.
channel. An auto-pneumatic transfer system
installed in 2012 at the DNRR can transfer a
sample from irradiation position to measuring
detector about 3 seconds.
The k-zero method for INAA has been
also developed to analyse airborne particulate
samples for investigation of air pollution; crude
oil samples and base rock samples for oil field
study. Based on developed k-zero-INAA
method, a multi-elements analysis procedures
have been applied to simultaneously determine
concentration for about 31 elements including
Al, As, Ba, Br, Ca, Cl, Cr, Cu, Dy, Eu, Fe, Ga,
Hf, Ho, K, La, Lu, Mg, Mn, Na, Sb, Sc, Sm,
Sr, Th, Ti, V, Yb, Zn.
B. Neutron activation analysis
Research on analytical techniques based
on neutron activation and other related
processes consists of the elaboration of
analytical processes and the design and
construction of analytical instruments.
C. Neutron beam utilization
Requests of many branches of the
national economy for various types of samples
have quickly been responded. NAA at the
DNRR has always been met the demand of
analytical services for geology exploration, oil
prospecting,
agriculture,
biology,
environmental studies, etc.
The reactor has four horizontal beam
ports, which provide beams of neutron and
gamma radiation for a variety of experiments.
They also provide irradiation facilities for
large specimens in a region close to the
reactor core. Besides, the reactor also has a
large thermal column with outside dimensions
of 1.2m by 1.2m in cross section and 1.6m in
length (Fig. 5).
The relatively high neutron flux in
irradiation channels of the reactor allows
elemental analysis using various neutron
activation approaches, such as Instrumental
NAA (INAA), Radiochemical NAA (RNAA),
Delayed NAA (DNAA) and Prompt gamma
NAA (PGNAA). By the end of 2013, a total
of about 60,000 samples have been irradiated
at the reactor with a yearly average of 2000
samples. It can be estimated that those make
up 60% of geological samples, 10% of
biological samples, 20% of environmental
samples, 5% of soil and agriculture materials,
3% of industrial materials.
Up to now, only three beam ports (No.2,
No.3 and No.4) and the thermal column have
been used for reseaches and applications. At
the beam port No.2, a BGO-HPGe gamma-rays
Compton suppression spectrometer has been
recently
installed
for
PGNAA
and
experimental researches on neutron capture
reactions. The filtered thermal neutron beams
extracted from the tangential beam port No.3
are used for nuclear structure studies,
especially for experimental determination of
nuclear energy levels and level density in
regions below neutron binding energy. The
filtered neutron beams at the piercing beam
port No.4 with quasi-monoenergies of 24keV,
54keV, 59keV, 133keV and 148keV are used
In order to determine the elements
having short-lived radionuclides, the method of
cyclic INAA with the alternation of irradiation
and measurement was implemented by using
the thermal column and vertical irradiation
5
for the mesurements of neutron total and
capture cross sections. In addition, these
neutron beams are also applied for practical
study on radiation shielding design. Typical
research activities using neutron beam of the
DNRR are listed below.
Thermal column
No. 2: Gamma spectrometry
system with BGO detector for
PGNAA and neutron capture
reactions study
No. 3: Nuclear
structure study
Colum n
door
Be am port # 2
Be am port # 3
The rm al
Colum n
Stainle s s s te e l
Alum inum
Graphite re fle ctor
Be llow s
as s e m bly
Be am port # 1
This port is closed
Pool tank w all
Core
Be am port # 4
The rm alizing colum n (clos e d)
Concre te
s hie lding
No. 4: Nuclear data
measurement
Spe nt fue l s torage tank
Fig. 5. Horizontal section view of the DNRR.
Neutron physics and nuclear data measurement
- Measurement of isomeric ratio created
in the reaction 81Br(n, )82Br on the 55 keV and
144 keV neutron beams;
In the keV energy region, filtered
neutron beams are the most intense sources,
which can be used to obtain neutron data for
reactors and other applications. The following
experiments have been carried out at the
DNRR including:
- And other investigations, such as
average resonance capture measurements,
using the - coincidence spectrometer for
study on the (n, 2) reaction, etc.
Application of neutron capture gamma ray
spectroscopy
- Total neutron cross section measurement
for 238U, Fe, Al, Pb on filtered neutron beams
at 144 keV, 55 keV, 25 keV and evaluation of
average neutron resonance parameters from
experimental data;
- Development of PGNAA technique using
the filtered thermal neutron beam in
combination with the Compton-suppressed
spectrometer for analyzing Fe, Co, Ni, C in
steel samples; Si, Ca, Fe, Al in cement
samples; Gd, Sm, Nd in uranium ores, Sm, Gd
in rare earth ores; etc.;
- Gamma ray spectra measurement from
neutron capture reaction of some reactor
materials (Al, Fe, Be, etc.) on filtered neutron
beam at 55 keV and 144 keV;
- Utilization of the PGNAA method for
investigating the correlation between boron
and tin concentrations in geological samples as
a geochemical indication in exploration and
assessment of natural mineral resources;
analyzing boron in sediment and sand samples
- Measurement of average neutron
radioactive capture cross section of 238U, 98Mo,
151
Eu, 153Eu on the 55 keV and 144 keV
neutron beams;
6
to complement reference data for such samples
from rivers;
Besides, the DNRR has been used as a
main tool for practical training, a set of
equipment was supported under IAEA TC
project, bilateral projects with the Japan
Atomic Energy Agency and Bhabha Atomic
Research Center of India. The measuring
systems for practices at the Training Center
can meet the fast increasing demand and is
expected to move forward to the regional
standard in the field of nuclear training.
- Development of the spectrometer of
summation of amplitudes of coinciding pulses
for (n, 2) reaction research and for measuring
activity of activated elements with high
possibility of cascade transitions.
D. Eduacation and training activities
Training Center at Dalat Nuclear
Research Institute which was established in
1999 is responsible for organizing training
courses and training in reactor engineering,
nuclear and radiation safety, application of
nuclear techniques and radioisotopes in
industry, agriculture, biology and environment,
etc. Training courses on non-destructive
evaluation (NDE) including radiographic
testing, ultrasonic testing as well as on security
of nuclear facilities and radiation sources have
also been done. The center also is the training
facility for expertise students from local
universities and foreign postgraduate students.
Thereby, the human resource development is
conducted annually so that it can deal with
scientific works of higher and higher quality
and meet a huge demand in the field of nuclear
science and technology in Vietnam in the
future. Thanks to the bilateral co-operation
with the Japan Atomic Energy Agency, US
Department of Energy, Bhabha Atomic
Research Center of India, and Korea Atomic
Energy Research Institute, we have conducted
a variety of training courses in the four
following key areas:
E. Other applications
Research on sediment using radiotracer
techniques was carried out to investigate bed
load layers displacement at estuaries
navigation channel region and to explain the
sediment deposition phenomenon causing
frequent dredging activities.
Research on radio-biology consists of
using gamma radiation associated with other
factors for improving agricultural seeds and
applying radioactive tracers for studying
biological metabolism, especially nutrition
problems. These studies are to investigate
phosphorus absorption and other nutritional
problems during the growing processes of rice
and other plants. Irradiation effects on some
plants to gain higher yield or environment
adapted varieties were also studied.
Gemstone colorizing experiments of
topaz and sapphire in the reactor core, in the
rotary rack as well as in horizontal channels
has been done.
As research purpose, silicon monocrystals have been irradiated at the central
neutron trap of the reactor. Irradiated products
of good quality, appropriate for fabrication of
power diodes and thyristors have been created
thanks to proper neutron distribution in this
irradiation facility and suitable cadmium ratio.
- Reactor engineering for nuclear power
programme;
- Research and development activities;
- State management in the field;
- And University lecturer training program.
7
IV. SOME MAIN REACTOR RENOVATION
PROJECTS PERFORMED
B. Reactor control and instrumentation system
modification
A. Reactor conversion from HEU to LEU fuels
The Control and Instrumentation System
In the framework of the program on
Russian Research Reactor Fuel Return
(RRRFR) and the program on Reduced
Enrichment for Research and Test Reactor
(RERTR), the DNRR core was partly
converted from HEU to LEU in September
2007.
(C&I) of the DNRR was designed and
manufactured by the former Soviet Union and
put into operation in November 1983. Due to
the spare part procurement problem was
suspected and using technology of the 1970’s
with discrete and low-level integrated
electronic components, the system technology
was somewhat obsolete and un-adapted to
tropical climate.
After this success, the full core
conversion study from HEU to LEU of the
DNRR was also carried out during years 2008
- 2010. The results of neutronics, thermal
hydraulics and safety analysis showed that a
LEU core loaded with 92 fuel assemblies and
12 beryllium rods around the neutron trap
satisfies the safety requirements while
maintaining the utilization possibility similar
to that of the previous HEU and recent mixed
fuel cores.
The first renovation work was
implemented during 1992-1993 period and the
renovated C&I system was commissioned at
the end of 1993. The most important
renovation task was to redesign and construct a
number of electronic systems/blocks, which
play the key role in enhancing the reliability of
the system. This renovation work was focused
mainly on the process and instrumentation
system, but not on the neutron measurement
and data processing parts. Because of that, it
was necessary to fulfill the second renovation
and modernization during the years of 20052007 to replace neutron measurement and
signal processing parts of the existing C&I
system by the digital system named ASUZ14R. The main items replaced under the
second modification are neutron detector
channels; neutron flux control system
(NFCS), reactor protection system, control
console and control panels, reactor protocol
and diagnostic system, etc.
Physics and energy start-up of the
DNRR for full core conversion to low
enriched uranium (LEU) fuel were performed
from November 24 th, 2011 until January 13 th,
2012 according to a planned program that
was approved by Vietnam Atomic Energy
Institute (VINATOM). At 15:35 on
November, 30 th, 2011 the reactor reached
criticality with core configuration including
72 LEU FAs and neutron trap in center. Then
the fuel loading for working core and power
ascension test were also carried out from
December, 6th, 2011 to January, 13 th, 2012.
Experimental results of physical and thermal
hydraulics parameters of the reactor during
start up stages and long operation cycles at
nominal power showed very good agreement
with calculated results and met the safety
requirements.
The commissioning of the new I&C
system was finished in August 2007 and
operating license was approved in October
2007.
8
V. CONCLUSIONS
REFERENCES
The DNRR has been safely operated and
effectively utilized for 30 years. To achieve
that, maintaining and upgrading the reactor
technological facilities have been done with a
high quality. The reactor physics and thermal
hydraulics studies have also provided the
important bases for safety evaluation and incore fuel management to ensure its safe
operation and effective exploitation. The
safety and security for the reactor are one of
the main issues that national and local
authorities are particularly interested in and
strongly support up.
[1] Nguyen Nhi Dien, Dalat Nuclear Research
Reactor - Twenty five years of safe operation
and efficient exploitation, Dalat, (March 2009).
[2] Duong Van Dong, Status of Radioisotope
Production and Application in Vietnam, Dalat
Sym. RR-PI-09, Dalat, (2009).
[3] V. V. Le, T. N. Huynh, B. V. Luong, V. L.
Pham, J. R. Liaw, J. Matos, Comparative
Analyses for Loading LEU Instead of HEU
Fuel Assemblies in the DNRR, RERTR Int’l
Meeting, Boston, (2005).
[4] P.V. Lam, N.N. Dien, T.D. Hai, L.B. Vien,
L.V. Vinh, H.T. Nghiem, N.M. Tuan and N.K.
Cuong, Results of the reactor control system
During 30 years of operation, the DNRR
has been playing an important role in the use
of atomic energy for peaceful purpose in
Vietnam. The reactor has been used for
radioisotope production for medicine and
industry purposes, NAA of geological, crude
oil and environment samples, performance of
fundamental and applied researches on
nuclear and reactor physics, as well as
creation of a large amount of human resource
with high skills and experiences on
application of nuclear techniques in the
country. A strategic plan and long-term
working plan for the DNRR has been set up in
order to continue its safe operation and
effective utilization at least to 2025.
replacement and reactor core conversion at the
Dalat nuclear research reactor, The 12th Annual
Topical Meeting on Research Reactor Fuel
Management, Hamburg, Germany, (2008).
[5] P.V. Lam, N.N. Dien, L.V. Vinh, H.T.
Nghiem, L.B. Vien and N.K. Cuong,
Neutronics and thermal hydraulics calculation
for full core conversion from HEU to LEU fuel
of the Dalat nuclear research reactor, RERTR
Int’l Meeting, Lisbon, Portugal, (2010).
[6] L.B. Vien, L.V. Vinh, H.T. Nghiem and N.K.
Cuong, Transient/ accident analyses for full
core conversion from HEU to LEU fuel of the
Dalat nuclear research reactor, RERTR Int’l
Meeting, Lisbon, Portugal, (2010).
[7] C.D. Vu, Study on application of k0-IAEA at
Dalat research reactor, Project report (code
CS/09/01-01), (2010).
It should be mentioned that a project for
establishment of a new nuclear science and
technology center with a high power research
reactor expected to put into operation between
2020-2022 is now under preparation and
consideration. Therefore, the DNRR will be
necessary and keep playing an important role
in scientific research, applications and human
resource development for Vietnam in the
coming time.
[8] N.N. Dien, L.B. Vien, P.V. Lam, L.V. Vinh,
H.T. Nghiem, N.K. Cuong, N.M. Tuan, N.M.
Hung, P.Q. Huy, T. Q. Duong, V.D.H. Dang,
T.C. Su, T.T. Vien, Some main results of
commissioning of the Dalat Nuclear Research
Reactor with low enriched fuel, Nuclear
Research Institute, (2012).
[9] Safety Analysis Report (SAR) for the Dalat
Research Reactor, Dalat, (2012).
9
Design Analyses for Full Core Conversion of
The Dalat Nuclear Research Reactor
Luong Ba Vien, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien Cuong
Reactor Center – Nuclear Research Institute – Vietnam Atomic Energy Institute
01 Nguyen Tu Luc, Dalat, Lamdong
Email: reactor@hcm.vnn.vn
Abstract: The paper presents calculated results of neutronics, steady state thermal hydraulics and
transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low
Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the
characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated
by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for
steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for
maximum hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power
distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters
of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB
margin were estimated in steady state thermal hydraulics investigation. The working core was also
analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure,
earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as
HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.
Keywords: HEU, LEU, neutronics, thermal hydraulics, safety analyses
I. INTRODUCTION
In this full core conversion study,
neutronics, thermal hydraulics and safety
analysis were carried out to investigate
characteristics of LEU working core fully
loaded with LEU fuel. All computer codes
were validated with HEU and mixed cores.
Using MCNP [6], REBUS-PC [5] and
VARI3D computer codes, a series of static
reactor physics calculation were performed to
obtain neutronics parameters of the working
core (see Fig. 1). Some parameters included in
the design of working core with shutdown
margin, excess reactivity taking into account of
irradiated Beryllium poisoning, control rod
worths, detailed power peaking factors,
neutron performance at the irradiation
positions, reactivity feedback coefficients, and
kinetics parameters. Because the higher content
of 235U in a LEU FA compared to HEU FA, it
is needed to rearrange the fuel assemblies and
berrylium rods with the different way to the
first HEU core to meet the safety requirements.
Thermal hydraulics parameters at steady
state condition were obtained by using
PLTEMP3.8 code [11] introduced models and
correlations that suitable for the concentric
tube fuel type and natural convection regime of
the DNRR.
Based on the neutronics analysis
parameters of the LEU core, the postulated
transients and accidents selected for the DNRR
are analyzed. The RELAP5/MOD3.2 code [15]
was used for analysis of RIA (Reactivity
Initiated Accident), LOFA (Loss Of Flow
Accident) transients.
LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG
These study results showed that a LEU
core loaded with 92 fuel assemblies and 12
beryllium rods around the neutron trap satisfies
the safety requirements while maintaining the
utilization possibility similar to that of the
previous HEU and recent mixed fuel cores.
REBUS-MCNP Linkage [7] was used
to calculate burnup distribution using “two
way” linking option in which MCNP is used
for calculating neutron flux and cross section
in one group neutron energy and burn up
calculation is implemented by REBUS-PC.
The MCNP5 code using an ENDF-B/VI
cross section library was used to construct a
detailed geometrical model of each reactor
component and calculate control rod worths,
multiplication coefficient, power distribution,
neutron flux performance in irradiation
positions, reactivity feedback coefficients, and
kinetics parameters (prompt neutron life time
and delayed neutron fraction).
A detailed geometrical model of reactor
components including all fuel assemblies,
control rods, irradiation positions, beryllium
and graphite reflectors, horizontal beam tubes
and thermal column was made in the MCNP
model, except in the axial reflectors above and
below the fuel assembly where some materials
were homogenized. Fig. 2a provides the radial
and axial models of the reactor for Monte
Carlo Calculations.
Fig. 1. The new designed working core loaded with
92 LEU FA and 12 Beryllium rods.
II. CALCULATION MODELS AND
COMPUTER CODES
A. Neutronics
Calculation
and
Thermal
Hydraulics
The diffusion code REBUS-PC with
finite difference flux solution method was used
to perform core calculation for reactor physics
characteristics and operation cycle calculations
with micro neutron cross sections according to
7 energy groups (collapsed from 69 energy
groups) that were generated by WIMS-ANL
code [4].
The FA cross sections were
generated in a radial geometry with each fuel
element depleted based upon its unique neutron
spectrum in the WIMS-ANL model. The
REBUS-PC fuel depletion chains included
production of six Pu isotopes, Am-241, Np237, and lumped fission product. Isotopic
precursors of Xe-135 and Sm-149 were also
included in the depletion chains so that Xe and
Sm transients during periods of shutdown and
startup could be modelled.
The kinetics parameters were calculated
also by VARI3D code. The real and adjoint
fluxes which are required to compute these
parameters were provided by DIF3D-a main
module of REBUS-PC code.
In diffusion theory, the reactor was
modeled in hexagonal geometry with a
heterogeneous representation of the fuelled and
non-fueled portions (see Fig. 2a). Each
homogenized fuel assembly was modelled
using five equal volume axial depletion zones.
The beam tubes were modeled using a
homogenized mixture of air or concrete,
graphite and aluminum.
The reactor models for diffusion and
Monte-Carlos computer codes were validated
by comparing with good agreement not only to
11
DESIGN ANALYSES FOR FULL CORE CONVERSION OF …
the fresh HEU configuration cores but also to
the HEU burnt cores. These models were then
applied for partial core conversion analyses of
DNRR [3]. The measured data collected during
the deployment of partial core conversion
project showed that the predicted calculation
results are quite acceptable [8,9].
for PLTEMP code. A fuel assembly was
modelled as three concentric cylindrical tubes.
Before using PLTEMP code to
calculate for DNRR with fully LEU fuel
assemblies, the code was validated by
comparing
analytical
results
with
experimental results of mixed-core.
The PLTEMP/ANL3.8 [15] thermalhydraulics code for plate and concentric-tube
geometries with capability of calculating
natural circulation flow was used for thermalhydraulics analyses. A chimney model as well
as Collier heat transfer correlation and CHF
Shah’s correlation have been recently
implemented make the code suitable DNRR
thermal-hydraulics calculation.
B. Transient/Accidents analyses
The DNRR has three barriers as other
research reactors that prevent or limit the
transport of fission products to the
environment, which are fuels and cladding,
reactor pool water and reactor confinement.
The safety system settings are showed in
Table I.
Fig. 2b shows the model of WWR-M2
fuel assembly, core and chimney of the DNRR
Table I. Safety system settings.
Parameters
Maximum thermal power (Pmax)
Minimum reactor period (Tmin)
Deficient level of pool water
Primary coolant flow rate
Secondary coolant flow rate
Safety system settings
550 kW (110% FP)
20s
60 cm
40 m3/h
70 m3/h
In the Safety Analysis Report (SAR) for
the DNRR [1], the possible initiating events
were classified by groups. The initiating events
in each group are then analyzed and justified in
order to identify the limiting event that will be
selected for further detail quantitative analysis.
The limiting event in each group has potential
consequences that exceed all others in that
group. Limiting events were selected for
detailed analyzed are as follows: (1)
Uncontrolled withdrawal of a control rod; (2)
Primary/Secondary
Pump
Failure;
(3)
Earthquake; (4) Fuel cladding failure. A
summary of the core parameters used for the
safety analysis is given in Table II.
Fig. 2a. Radial and Axial models for Monte
Carlo calculations (upper) and Radial model
for Diffusion Theory calculations (under).
12
To ensure the fuel clad integrity in
operational condition and to protect the public
and the environment in case of accident, in the
SAR for the DNRR, the following acceptance
criteria were defined:
For
occurrences:
anticipated
1)
operational
3)
(1) Minimum margin to departure from
nucleate boiling (DNB) shall be over 1.5;
(2) Maximum temperature of fuel cladding
shall not exceed 400oC;
(3) Fuel
assured.
cladding
integrity
shall
be
- For accident conditions:
(1) Core covering shall be maintained;
(2) Core
damaged;
shall
not
be
remarkably
(3) Release of fission products into the
environment shall not be remarkable.
The RELAP5 code was used for
analyzing the events of excess reactivity
insertion by uncontrolled withdrawal of a
control rod and earthquake. The piping of the
primary cooling system and pool volume were
divided into nodes with similar dynamic
characteristics. The reactor core was divided
into 2 channels with axial nodes. The hot
channel represents the hottest channel in the
2)
Fig. 2b. DNRR model for PLTEMP
(1-fuel assembly cross-section; 2-FA model
for PLTEMP; 3-reactor coolant system model).
core corresponding to a cooling channel with
maximum heat flux. The average channel
represents the rest of the cooling channels.
Each channel was modelled as three fuel
element plates and four coolant flow gaps. The
nodding diagram of the DNRR for
RELAP5/3.2 is presented in Fig. 2c.
The MACCS2 code [19] was used to
estimate the radiological impact of the
hypothetical accident on the surrounding
public. The core radiation inventories were
calculated by ORIGEN2 code [20] using
neutron cross-sections of the actinides obtained
from MCNP5 code.
Fig. 2c. Nodding diagram of DNRR for RELAP5/3.2.
13
11
Table II. Core parameters used for safety analysis.
Parameters
Power, kW
Coolant inlet temperature, oC
Peaking factor (shim rods at 300 mm)
- Axial peaking factor
- Radial peaking factor
- Local peaking factor
Reactor kinetics
- Prompt neutron life, s
- Delayed neutron fraction (1$)
Temperature reactivity coefficients
- Moderator, %/K; (293-400oK)
- Fuel, %/oC;
(293-400oK)
(400-500oK)
(500-600oK)
- Void, %/% of void
(0-5%)
(5-10%)
(10-20%)
Reactivity control
- Shutdown worth, % (2 safety rods)
- Maximum withdrawal speed of one shim
rod, mm/s
and of the regulating rod, mm/s
Reactor protection characteristics
- Response time to overpower scram, s
- Response time to fast period scram, s
Start-up range
Working range
- Drop time of control rods, s
Values
500
32
1.363
1.376
1.411
8.92510-5
7.55110-3
- 1.26410-2
- 1.8610-3
- 1.9210-3
- 1.5610-3
-0.2432
-0.2731
-0.3097
3.7
3.4
20
0.16
9.1
6.7
0.67
results in large negative reactivities which alter
flux and power distributions.
III. RESULTS AND DISCUSSIONS
A. Neutronics and Thermal Hydraulics
Program Beryl [10] has been modified
to calculate the 3He, 6Li and 3H concentrations.
The MCNP5 was then used to determine the
poisoning effect of 3He, 6Li and 3H
concentrations on reactor core reactivity. The
comparison of reactivities between calculation
results and measured data of some beryllium
blocks irradiated in DNRR (Table III) shows
that the negative reactivity of irradiated
beryllium determined by above-mentioned
3
He and 6Li Poisoning of Irradiated
Beryllium [10]
Since 1984, the DNRR has been put into
operation with a considerable amount of
Beryllium used for neutron trap at the core
center and periphery for improving neutron
reflection around. Because Beryllium has large
thermal neutron absorption cross sections, the
buildup of 3He, 6Li and 3H concentrations
14
method is reliable. Six beryllium rods were
used for measurement, two fresh beryllium
rods and four irradiated beryllium rods (two
beryllium rods at the end 1994 and two at the
end 2002). 9-6 and 5-6 positions were chosen
to measure reactivity of couple beryllium rods
through changing position of control rod
(Regulating Rod). The error of control rod
position is estimated about 0.4 cent.
Following calculation scheme for beryllium
poisoning above, reactivity of the poisoning
process in new configuration cores about -1$.
All calculation for design LEU cores,
beryllium poisoning is included in the model
for MCNP code.
Table III. Comparison of calculated and measured of reactivities of irradiated beryllium rods in DNRR.
Measured reactivity
(Cent)
Calculated reactivity
(Cent)
Error
(%)
2 Beryllium Rods at
the end 1994
-3.89  0.4
-4.65  0.0038
16.34
2 Beryllium Rods at
the end 2002
-6.28  0.4
-7.19  0.0039
12.66
The working core characteristics
From the calculation results of shutdown
margins, excess reactivities, power peaking
factors, and neutron performance at the
irradiation positions of 4 candidates cores, the
working core with the better features from the
safety and utilization point of view was chosen
for detailed analysis. The main calculated
characteristics of working core is showed in
the Table IV. The shutdown margins of the
core is met the safety requirement of -1.0%.
Calculated neutron flux at the neutron trap of
the core is nearly the same as that of mixed
core (92HEU+12LEU). Table V shows the
control rod worths. Detailed neutron flux
performance at the main irradiation positions
are presented in Table VI.
Table IV. Calculation results of working core compared with current mixed core.
Parameters
LEU Core
Excess Reactivity (%) – Fresh
Excess Reactivity (%) – After 600FPDs
Shutdown Margin (%) – Fresh
Shutdown Margin (%) – After 600 FPDs
Radial Power Peaking Factor
Control Rods Out
Control Rods In
Thermal Neutron Flux at Neutron Trap Center (n/cm2)
Control Rods Out
Control Rods In
Fast Neutron Flux at Neutron Trap Center (n/cm2)
Control Rods Out
Control Rods In
6.63
3.79
-2.92
-6.62
15
11
1.398
1.434
Current Mixed
Core
-4.56
1.431
2.22E+13
2.14E+13
2.22E+13
1.95E+12
1.92E+12
3.15E+12
Table V. Control Rods worths (%k/k).
Control Rods
Shim rod 1
Shim rod 2
Shim rod 3
Shim rod 4
Regulating rod
Safety rod 1
Safety rod 2
Core1
Fresh
2.5896
2.6100
2.7784
2.4687
0.4363
2.1955
2.2356
MCNP
error
0.000091
0.000111
0.000118
0.000122
0.000126
0.000106
0.000119
Core1
Burnt
2.3539
2.4033
2.5381
2.2604
0.3629
2.3084
2.3579
MCNP
error
0.000091
0.000124
0.000122
0.000117
0.000119
0.000115
0.000105
Table VI. Neutron flux performance.
Maximum
Average
Maximum
Average
Maximum
Average
Maximum
Average
Fresh
2.07E+13
1.45E+13
9.45E+12
7.00E+12
5.41E+12
4.11E+12
9.24E+12
6.85E+12
Burnt
2.20E+13
1.49E+13
9.86E+12
7.12E+12
5.66E+12
4.18E+12
9.71E+12
7.01E+12
Epithermal,
<0.821MeV
(n/cm2.s)
Fresh
Burnt
6.79E+12 7.12E+12
6.00E+12 6.04E+12
8.19E+12 8.42E+12
6.53E+12 6.51E+12
9.63E+12 9.76E+12
7.23E+12 7.15E+12
8.02E+12 8.22E+12
6.41E+12 6.40E+12
Average
3.55E+12
3.56E+12
7.58E+11 7.56E+11
Irradiation positions
Neutron
Trap
Channel
13-2
Channel
7-1
Channel
1-4
Rotary
Specimen
Thermal, <0.625eV
(n/cm2.s)
Power Distribution and Power Peaking
Factors
Fast, <10MeV
(n/cm2.s)
Fresh
1.83E+12
1.62E+12
2.98E+12
2.46E+12
4.22E+12
3.19E+12
2.92E+12
2.42E+12
Burnt
1.92E+12
1.63E+12
3.02E+12
2.44E+12
4.26E+12
3.15E+12
2.99E+12
2.40E+12
1.93E+11
1.93E+11
of control rods at 250 mm. Detailed axial
power distribution according to control rod
position was also calculated. Radial power
distributions at different control rod position
are showed in Fig. 3.
Power peaking factors of the core with
different position of control rods were
calculated and presented in Table VII. The
maximum power peaking factor is in position
Table VII. Power peaking factor according to control rod positions
Position
(mm)
0
150
200
250
300
350
600
F.A. Radial
1.378
1.378
1.375
1.377
1.376
1.378
1.378
Peaking Factor
Core Radial
Axial
1.398
1.296
1.399
1.343
1.403
1.356
1.409
1.365
1.411
1.363
1.415
1.336
1.434
1.284
16
Total
2.498
2.589
2.615
2.648
2.646
2.605
2.537
0.973
0.947
0.913
0.901
0.877
0.875
0.917
0.906
0.962
0.939
1.138
1.090
0.998
0.974
1.018
0.991
1.020
0.992
0.979
0.964
1.004
1.008
1.090
1.126
1.164
1.129
1.165
1.126
1.107
1.085
SR
1.410
1.370
1.006
1.001
1.021
1.023
1.406 1.082
1.368 1.124
1.281
1.296
1.259
1.283
ShR
0.810
0.863
1.016
0.994
1.106
1.145
0.860
0.910
0.918
0.917
1.421 1.408
1.381 1.368
1.038
1.038
1.031
1.019
1.198
1.156
SR
1.124
1.097
1.005
1.006
0.986
0.970
1.180
1.137
1.079
1.114
0.816
0.866
0.858
0.857
0.985
0.960
ShR
0.745
0.803
0.843
0.846
0.959
0.933
RgR
0.843
0.837
0.808
0.818
0.755
0.808
0.919
0.903
0.865
0.855
0.911
0.900
0.918
0.919
0.868
0.919
1.252
1.273
0.929 1.296
0.975 1.312
0.921
0.915
0.842
0.850
0.930
0.929
1.139
1.122
1.307
1.284
0.908
0.913
0.787
0.843
0.882
0.937
1.313
1.289
1.353
1.321
0.963
0.958
0.775
0.833
ShR
1.220
1.192
0.830
0.881
0.968
0.959
0.885
0.895
0.835
0.889
0.996 1.364
0.987 1.332
0.903
0.904
0.996
0.977
0.858
0.858
0.825
0.876
ShR
0.906
0.953
0.983
0.975
1.056
1.027
1.167
1.117
0.762
0.817
0.790
0.841
0.872
0.873
0.849
0.851
0.980
0.957
0.911
0.887
0.845
0.835
0.841
0.836
0.831
0.836
0.895
0.889
0.973
0.949
Fig. 3. Radial power distribution (Upper values: Fresh Core; Under values: Burnt Core)
Reactivity Feedback Coefficients and
Kinetics Parameters
kinetics parameters of the LEU cores
calculated using the VARI3D and MCNP5
codes. The calculated results from the two
computer codes are in good agreement.
These data will be used in transient
calculation for safety analysis of fully LEU
core of DNRR.
Reactivity
feedback
coefficients
calculated with the MCNP5 are depicted in
Table VIII. The negative results of reactivity
feedback coefficients show the inherent
safety of the LEU core. Table IX shows the
Table VIII. Feedback reactivity coefficients.
Parameter
DATA
±σ
-0.01317
0.00005
-0.00192
-0.00182
0.00005
0.00003
-0.00154
0.00002
-0.2514
-0.2784
-0.3255
0.0011
0.0012
0.0006
o
Moderator Temperature Reactivity Coefficient (%/ C)
293 oK to 400 oK
Fuel Temperature (Doppler) Reactivity Coefficient
(%/oC)
293 oK to 400 oK
400 oK to 500 oK
500 oK to 600 oK
Moderator Density (Void) Reactivity Coefficient (%/%
of void)
0 to 5 %
5% to 10 %
10 % to 20 %
17
Table IX. Calculated results of kinetics parameters for LEU core.
Family, i
Decay Const.
λi (s-1)
Relative Yield
ai
1
2
3
4
5
6
1.334E-02
3.507E-02
3.273E-02
1.804E-01
1.208E-01
1.742E-01
3.030E-01
3.843E-01
8.503E-01
1.594E-01
2.856E+00
6.666E-02
Total delayed neutron fraction, β
VARI3D
MCNP5 – Fresh
MCNP5 – Burnt
Prompt neutron life time, ℓ
Burn up calculation
Fraction
βi
2.648E-04
1.363E-03
1.315E-03
2.902E-03
1.204E-03
5.033E-04
7.551E-03
7.761E-03
7.762E-03
8.925E-05
extended about 11 years (calculated with 1300
hours per year) or 600 full power days (FPDs).
The burn up of U-235 reached average value of
8.2% and maximum value of 11.4%. In the
next cycle, about 8 fuel assemblies will be
inserted so the reactor core will operate with
100 fuel assemblies. The Fig. 4 shows burn up
distribution after 600 FPD operation.
The first cycle length was estimated by
REBUS-MCNP Linkage system code. Burn up
calculations were performed by assuming that
shim rods and regulating rod were in critical
position following each burn-up step. The
value of reactivity for Xe-135 poisoning was
estimated about 1.2% k/k. The result of
depletion shows that operating time may be
Fig. 4. Burn up distribution using REBUS-MCNP Linkage system after 600 FPD.
18
working core meets the requirements of
thermal hydraulics safety. At the power of
500kW with systematic errors, maximum
cladding temperatures are below the
permissible value of 103oC [2] and far below
the ONB temperature (estimated about 116oC
using Forster-Greif correlation). The maximum
outlet coolant temperature is calculated about
60oC, much lower than saturated temperature
(108oC).
The PLTEMP code was used for
calculating cladding temperature, coolant
temperature and safety margins for the
candidate cores. The calculated results are
presented in Table X and Fig. 5. At nominal
power without uncertainties and maximum
permissible inlet temperature (32oC), the
maximum cladding temperature is 90.50oC.
Calculation was carried out for nominal power
with systematic errors (equivalent to 70kW
power) and the maximum cladding temperature
is 95.69oC. In this case, by using Shah’s
correlation, the obtained minimum DNBR is
9.9. The minimum flow instability power ratio
(MFIPR) is 2.04. From above-mentioned
calculated results, it may conclude that the
Fig. 6 shows the comparison of cladding
temperature of 92FA LEU cores and 89FA
fresh HEU core. Compared to the 89FA fresh
HEU core established in 1984, cladding
temperature of working core is about 2oC
lower.
Table 10. Cladding temperature and ONB margin by PLTEMP Code.
Distance
(cm)
100.0
550kW
with sys. error
TTc(oC)
ONB(oC)
68.95
47.39
76.58
40.14
85.63
31.51
92.89
24.48
97.33
20.06
99.23
17.97
98.43
18.41
94.41
21.94
89.94
25.90
84.43
30.74
78.79
35.57
74.64
38.98
600kW
with sys. error
TTc(oC)
ONB(oC)
70.96
45.59
78.97
37.97
88.46
28.91
96.05
21.57
100.68
16.95
102.65
14.80
100.76
16.40
96.22
20.48
91.57
24.60
85.89
29.59
80.13
34.52
75.94
37.93
100
Temperature ( o C)
Temperature ( o C)
2.5
7.5
12.5
17.5
22.5
27.5
32.5
37.5
42.5
47.5
52.5
57.5
500kW
without sys. error
with sys. error
TTTc(oC)
ONB(oC)
Tc(oC)
ONB(oC)
63.91
51.89
66.89
49.24
70.56
45.59
74.13
42.36
78.46
38.07
82.71
34.18
84.83
31.90
89.61
27.50
88.77
27.95
93.85
23.26
90.50
26.05
95.69
21.25
89.86
26.34
94.95
21.63
87.10
28.58
91.91
24.13
83.98
31.14
88.24
27.24
79.67
34.76
82.92
31.91
74.91
38.73
77.42
36.64
71.21
41.70
73.32
40.02
90.0
80.0
T-clad
70.0
60.0
92 LEU FA Core2
95
89 HEU FA Core
90
92 LEU FA Core1
85
50.0
80
40.0
T-coolant
75
30.0
70
20.0
DT-ONB
65
10.0
0.0
0
10
20
30
40
50
60
60
0
Distance from core bottom (cm)
Fig. 5. T/H parameters at 500kW without
uncertainties.
10
20
30
40
50
60
Distance from core bottom (cm)
Fig. 6. Comparison of calculated cladding temperature
between 92FA LEU cores and HEU core.
19
The event of one shim rod inadvertently
withdrawal with speed of 3.4 mm/s from stable
operation of 100%FP (500 kW) are showed in
Fig. 7 and Table XI. In this case, the reactor
power increases and reaches to the over-power
setting value (110%FP) within 3.39 seconds
generating a scram signal. After a delay time of
0.16 seconds the reactor power is rapidly
suppressed because of the control rods
insertion. The peak power of the reactor is only
attained 0.553 MW with a slight increase of the
maximum fuel cladding temperature. With the
assumption of no overpower scram signal
appeared, a fast period scram signal is
generated after 8.33 seconds from the initiation
of transient event. The reactor will be
shutdown after 6.7 second delay with a peak
power of 0.957 MW. The maximum fuel
cladding temperature is predicted to be 113.0oC
without any nucleate boiling occurrences. The
minimum DNBR (Departure from Nucleate
Boiling Ratio) estimated about 6.5 is much
higher than the acceptance criterion of 1.5.
2. Transient/Accidents analyses
Uncontrolled withdrawal of one shim
rod or the regulating rod
In this event, it is assumed that one of
the shim rods or the regulating rod is
withdrawn in the most effective part from 200
mm to 400 mm at the speed for 3.4 mm/s of
shim rod and 20 mm/s for regulating rod. The
initial conditions are as follows:
a) Start-up case:
(1) -1% k/k sub-critical; Power level: 105
%FP; Coolant inlet temperature: 32oC.
(2) Critical state; Power level: 10-3%FP;
Coolant inlet temperature: 32oC.
b) Steady-state operation:
Power level: 100%FP; Coolant inlet
temperature: 32oC.
In sub-critical status, when one shim rod
is inadvertently withdrawn with the speed of
3.4 mm/s, from the core, the reactor power
only increases to the maximum value of
2.7810-7 MW while the fuel cladding
temperature is unchanged. With initial
conditions of criticality at the power level of
10-3%FP (510-6 MW) if there is no fast period
signal and the overpower trip setting is
110%FP, the fuel clad temperature reaches to
97.8oC, but still far below ONB (Onset of a
Nucleate Boiling) temperature.
With the same initial conditions, the
calculated results for the event of withdrawal
of the regulating rod are slightly different from
those of above-mentioned event, when one
shim rod is withdrawn. This can be explained
by the similar insertion rate of reactivity in the
two cases (about 0.02$/s). The regulating rod
has lower reactivity worth but higher
withdrawal velocity compared to those of a
shim rod.
Table XI. Transient results of one shim rod withdrawal from 100%FP.
Values
Parameters
110%FP Scram
Period Scram
3.6
15.1
0.553
0.957
3.7
15.2
91.9
113.0
Time to Peak Power, s
Peak Power, MW
Time to Peak Clad Temperature, s
o
Peak Clad Temperature, C
Minimum DNBR
6.5
20
20
Period scram
0.9
18
0.8
16
0.7
14
0.6
12
0.5
10
DNBR
0.4
8
0.3
6
DNBR
1.0
Temperature (oC)
Power (MW)
120
Period scram
110
100
90
80
70
60
0.2
4
Overpower scram with set point 110% FP
0.1
2
0.0
0
0
2
4
6
8
10
12
14
16
18
20
Time (s)
50
Overpower scram with set point 110% FP
40
0
2
4
6
8
10
12
14
16
18
20
Time (s)
Fig. 7. Reactor power and cladding temperature transient of one shim rod withdrawal
from a stable operation of 100%FP.
Cooling pump failure
measures
undertaken
in
design
and
construction, the removal of all control rods
would not exceed 10 mm and insert a step
positive reactivity estimated of 0.3$. With this
reactivity insertion, the scram set-point of
reactor overpower is attained almost
instantaneously. If the reactor scram is
initiated by overpower signal with a delay of
0.16 sec, the fuel surface temperature
increases slightly before decreases with the
power, the residual heat after shutdown is
sufficiently removed from the fuel by natural
convection of pool water without considerable
increase of the temperature.
In the event of in-service primary or
secondary cooling pumps stopped working, the
reactor is automatically shutdown by an
abnormal technological signal on low flow rate
(the setpoint is 40 m3/h for the primary flow,
and 70 m3/h for the secondary flow). The
residual heat after shutdown is about 6% FP
(30 kW) in maximum and the natural
convection process can itself assure the good
cooling of the core.
If the reactor is purposely maintained at
full power operation, failure of cooling pumps
leads to loss of heat removal from the pool
water, and thus gradually increases of the pool
water temperature. The results show that the
clad temperature reaches the maximum
allowable operating clad temperature of 103 oC
at about 55 min; i.e. the reactor could continue
its operation for 55 minutes within the envelope
of the limiting conditions of operation. The
results also show that even at the end of the
simulation (7000 s) the clad temperature has
been well below that of the acceptance criterion
for anticipated operational occurrences.
Fig. 8 shows the analyzing results of
the earthquake event assuming the
protection system fails to shutdown the
reactor, and Because of the loss of offsite
power due to the earthquake, the primary and
secondary pumps stop operating. In this case,
the reactor power increases to the max value of
1.525 MW after 200 seconds from the
initiation of this event. The reactor power then
rapidly decreases because the significant
increasing of core water temperature so that the
positive reactivity insertion is overtaken by the
negative reactivity feedback (about -0.44$).
The reactor is then kept at subcritical state. The
cladding temperature reaches a maximum
value of 118.2oC, then decreases with no
Earthquake
The postulated event of an earthquake of
intensity grade VI is assumed to occur while
the reactor is at full power. Owing to the
21
1.6
16
1.4
14
DNBR
1.2
12
DNBR
Power (MW)
In case the cooling pumps remain
working after the earthquake event (very
unlikely); the peak power reaches 1.57 MW
within 300 seconds and decreases due to
negative temperature feedback to a stable
value of about 1.12 MW. The cladding
temperature reaches to a maximum value of
118.38oC then gradually decreases to a stable
value of 115oC without nucleate boiling. The
maximum temperature of outlet water is 89 oC
at the peak power then decreases and
stabilizes at about 82 oC, well below the
saturation point. The minimum DNBR in this
case estimated about 4.74 is still far from the
acceptance criterion.
Temperatute (oC)
significant overheating of the fuel. The
maximum outlet water reaches 89oC and
gradually decreases to a value at about 60oC,
which is still far below the saturation
temperature. The minimum DNBR of 4.79 is
much higher than the acceptance value.
120
Max. Cladding Temperature
110
100
90
1.0
10
0.8
8
0.6
6
60
0.4
4
50
0.2
2
40
80
70
Max. Water Temperature at Outlet
Water Temperature at Inlet
Power
0.0
0
500
1000
1500
2000
2500
3000
3500
4000
0
4500
5000
Time (s)
30
0
500
1000
1500
2000
2500
3000
3500
4000
4500
5000
Time (s)
Fig. 9. Power and Temperature responses to earthquake event while cooling pumps are stopped functioning.
Fuel cladding failure (MHA)
For the derivation of source term of this
event, it is assumed that no core melting occurs
but cladding rupture of one fuel assembly is
involved. It is also assumed that the damaged
fuel assembly is irradiated at the maximum
neutron flux position in the core and the fuel
damage occurs immediately at the end of
operating cycle of 100 hrs with no decay.
From the damaged fuel assembly, 100%
of noble gases (Xe, Kr), 25% halogens (I), and
1% of other radionuclides (Cs, Te) [21] are
released directly to the reactor building with
the assumption of no retention of volatile
fission products in the pool water. During the
accident evolution, the emergency ventilation
system is not in place, the normal ventilation
system V1 is in operation but HEPA filter with
22
11
95% efficiency is not available, and there are
no decay and deposition of radionuclides
within the reactor building.
The evaluation of dose to a member of
the public is calculated by code MACCS2
version
1.13.1,
using
the
following
assumptions: (1) The radionuclides are
released to the environment through the 40 m
stack; (2) The Gaussian plume model is used to
calculate air concentration of radioactivity; (3)
Tadmor and Gur parameterization is used for
this analysis; (4) No building in the vicinity (an
open area release), plume rise mechanics only
due to momentum rise (non-buoyant plume)
and no wet deposition are assumed; (5) The dry
deposition velocity is assumed to be 0.01 m/s,
which corresponds to a particle with an
aerodynamic equivalent diameter of 2 m to 4
m (for unfiltered particulate releases) [15]; (6)
Surface roughness length is specified as 50 cm;
(7) Mixing layer height is assumed to be 500 m
(see Table 36 in Appendix VII of Ref. 21); (8)
The breathing rate is 3.3x10-4 m3/s; (9) No
shielding and sheltering are assumed; (10)
Doses at each downwind distance are
calculated for one year after the arrival of the
plume (11). The environmental release is
assumed to begin at the start of the weather
conditions: Pasquill class D2.0 (most frequent
stability class and most frequent wind speed).
The effective equivalent doses, including
cloudshine dose, inhalation dose and
groundshine dose, as a function of the distance
from the source are shown in Table XII and
Fig. 10. It is seen that radiation exposure to the
general public with the maximum effective
dose of 0.64 mSv/year at distance from 400 m
to 500 m from the stack. This value is lower
than the annual dose limit of 1.0 mSv specified
for the public [22].
Table XII. The annual effective dose to the public vs distance for the MHA.
Distance
(m)
Effective Dose
(mSv)
Distance
(m)
Effective Dose
(mSv)
50
150
250
350
450
550
650
750
850
950
4.80E-02
1.43E-01
4.95E-01
6.42E-01
6.44E-01
5.94E-01
5.33E-01
4.74E-01
4.21E-01
3.75E-01
1100
1300
1500
1700
1900
2250
2750
3250
3750
4250
3.18E-01
2.59E-01
2.16E-01
1.83E-01
1.57E-01
1.23E-01
9.14E-02
7.08E-02
5.66E-02
4.64E-02
7.00E-01
Effective Equivalent Dose, mSv
6.00E-01
5.00E-01
4.00E-01
3.00E-01
2.00E-01
1.00E-01
0.00E+00
0
500
1000
1500
2000
2500
3000
3500
4000
4500
5000
Downwind Distance, m
Fig. 10. The annual effective dose to the public in MHA event within 5 km.
23
11
- If one of the cooling pumps stopped
working, the reactor is automatically shutdown
by a scram signal on low flow rate. The decay
heat is removed from the fuel by natural
convection of pool water. In this event, if the
reactor was purposely maintained at full power,
it could be safely operated for 55 minutes when
maximum cladding temperature is still lower
than the permissible value of 103oC.
IV. CONCLUSIONS
Neutronics,
steady-state
thermalhydraulic and transient/accidents analyses for
Dalat Nuclear Research Reactor show that with
a slight change in arrangement of Be rods, the
main features of 92 LEU WWR-M2 FA cores
are equivalent to those of HEU and current
mixed fuel cores.
- The postulated earthquake event of
MSK intensity grade VI would cause a step
reactivity insertion of 0.3$. Even if the reactor
fails to be scrammed, this positive reactivity
can be covered by negative temperature
feedback if the cooling pumps are stopped
simultaneously, keeping the reactor sub-critical.
In case the cooling pumps continue operating
after earthquake event, the negative temperature
feedbacks act to bring the reactor power to a
stable level of about 1.12 MW without nucleate
boiling. The minimum DNBR is much higher
than the acceptance criterion of 1.5.
The negative values of reactivity feedback
coefficients show the inherent safety feature and
shutdown margin of both candidate cores meets
the safety required value of -1% k/k. The
working core with 92 fresh LEU fuel assemblies
can be operated for 600FPDs or about 11 years
based on the current operating schedule without
shuffling. The neutron fluxes at the irradiation
positions are not much different from those of the
current mixed fuel core.
In thermal hydraulics aspect, the
requirement of thermal-hydraulic safety margin
for two candidate cores in normal operational
condition is satisfied. The calculated maximum
cladding temperature in operational condition
is below the permissible value of 103oC.
- The maximum hypothetical accident
assumes 100% of noble gases (Xe, Kr), 25%
halogens (I), and 1% of other radio-nuclides (Cs,
Te) in a most power fuel assembly after a long
run are released into the environment through
40m high stack. This event is considered to be
very unlikely to occur for the DNRR. Even so, it
would not cause undue radiological risk to the
environment or the public.
In transient/accidents aspect, some
postulated initiating events and accident related
to the conversion of the DNRR to full LEU
core were selected and analyzed. Based on the
calculated results, conclusions might be
withdrawn as following:
ACKNOWLEADGMENTS
- The excess reactivity insertions when
inadvertent withdrawals of control rod from
start-up or nominal power operation are
prevented by safety settings to initiate the
reactor scram at overpower and fast period.
None of these initiators would lead to the ONB
and DNB, ensuring the integrity of the fuel
cladding. The residual heat after shutdown is
sufficiently removed from the fuel by natural
convection of pool water.
The authors would like to express their
gratitude to experts from the Reduced
Enrichment for Research and Test Reactors
(RERTR) program of Argonne National
Laboratory for financial support as well as very
useful discussions during design calculation of
full core conversion for the Dalat Nuclear
Research Reactor.
24
REFERENCES
[1] Safety Analysis Report for the Dalat Nuclear
Research Reactor, (2003).
[2] VVR-M2 and VVR-M5 Fuel AssembliesOperation Manual, 0001.04.00.000 PЭ, (2006).
[3] Report on the 7th Core Loading for the Dalat
Nuclear Research Reactor, NRI, Dalat, (in
Vietnamese), (2009).
[4] J.R. Deen, W.L. Woodruff, C.I. Costescu, and
L. S. Leopando, “WIMS-ANL User Manual
Rev. 5”, ANL/RERTR/TM-99-07, Argonne
National Laboratory, (February 2003).
[13] Technical Design for Reconstruction and
Enlargement of the Dalat Nuclear Research
Reactor - Volume 3. State Design Institute,
USSR State Committee for the Utilization of
Atomic Energy, Moscow, (in Russian) (1979).
[14] Additional Physics and Thermal-Hydraulic
Data for Reactor IVV-9, State Design Institute,
USSR State Committee for the Utilization of
Atomic Energy, Moscow, (in Russian) (1980).
[15] RELAP5/MOD3 Code Manual, SCIENTECH,
Inc. Rockville, Maryland, (1999).
[5] P. Olson, “A Users Guide for the REBUS-PC
Code, Version 1.4,” ANL/RERTR/TM02-32,
(December 21, 2001).
[16] W. L. Woodruff, et al., A comparison of the
PARET/ANL and RELAP5/MOD3 Codes for
the Analysis of IAEA Benchmark Transients
and the SPERT Experiments, RERTR
Program, ANL.
[6] J. F. Briesmeister, Ed., “MCNP – A General
Monte Carlo N-Particle Transport Code,
Version 4C”, LA-13709-M (April 2000).
[17] V. V. Le and T. N. Huynh, Application of
RELAP5/MOD3.2 for the DNRR, Proceedings
of JAEA Conf. 2006-001.
[7] John G. Stevens, “The REBUS-MCNP Linkage”,
Argonne National Laboratory, (2007).
[18] M. M. Shah, Improved General Correlation for
Critical Heat Flux during Upflow in Uniformly
Heated Vertical Tubes, International Journal of
Heat and Fluid Flow, Vol. 8, No. 4, pp. 326335 (1987).
[8] N.A. Hanan, J.R. Deen, J.E. Matos, “Analyses
for Inserting Fresh LEU Fuel Assemblies
Instead of Fresh Fuel Assemblies in the DNRR
in Vietnam, 2004 International Meeting on
RERTR, Vienna, (2004).
[9] V. V. Le, T. N. Huynh, B. V. Luong, V. L.
Pham, J. R. Liaw, J. Matos, “Comparative
Analyses for Loading LEU Instead of HEU
Fuel Assemblies in the DNRR”, RERTR Int’l
Meeting, Boston, (2005).
[10] Teresa Kulikowska et al., Raport IAE-40/A,
(1999).
[11] Arne P. Olson, M. Kalimullah, “A users guide
to
the
PLTEMP/ANL
V3.8
Code”,
ANL/RERTR, Argonne National Laboratory,
(June, 2009).
[12] Le Vinh Vinh, Huynh Ton Nghiem and
Nguyen Kien Cuong, “Preliminary results of
full core conversion from HEU to LEU fuel of
the Dalat Nuclear Research Reactor”, RERTR
Int’l Meeting, Beijing, (2009).
[19] MACCS2 Computer Code Application
Guidance for Documented Safety Analysis,
U.S. Department of Energy, (June 2004).
[20] A.G. Croff, A User Manual for the ORIGEN2
Computer Code, Oak Ridge National
Laboratory, (1980).
[21] INTERNATIONAL
ATOMIC
ENEGY
AGENCY, Derivation of the Source Term and
Analysis of Radiological Consequences for
Research Reactor Accidents, SAFETY
REPORTS SERIES No. 53, VIENNA (2008).
[22] Governmental Decree for the Implementation
of the Ordinance on Radiation Protection and
Control, Government of the Socialist Republic
of Vietnam, No. 50/1998/ND-CP, Hanoi, (in
Vietnamese) (1998).
1125
Conceptual Nuclear Design
of a 20 MW Multipurpose Research Reactor
Nguyen Nhi Dien, Huynh Ton Nghiem, Le Vinh Vinh, Vo Doan Hai Dang
Reactor Center – Nuclear Research Institute – Vietnam Atomic Energy Institute
01 Nguyen Tu Luc, Dalat, Lamdong, Vietnam
Seo Chulgyo, Park Cheol, Kim Hak Sung
Korean Atomic Energy Institute,
150 Dukjin-dong, Yuseong-gu, Taejeon 305-353, Korea
Abstract: This paper presents some of studied results of a pre-feasibility project on a new research
reactor for Vietnam. In this work, two conceptual nuclear designs of 20 MW multi-purpose research
reactor have been done. The reference reactor is the light water cooled and heavy water reflected
open-tank-in-pool type reactor. The reactor model is based on the experiences from the operation and
utilization of the HANARO. Two fuel types, rod and flat plate, with dispersed U 3Si2-Al fuel meat are
used in this study for comparison purpose. Analyses for the nuclear design parameters such as the
neutron flux, power distribution, reactivity coefficients, control rod worth, etc. have been done and the
equilibrium cores have been established to meet the requirements of nuclear safety and performance.
Keywords: HANARO, AHR, MTR, MCNP, MVP, HELIOS, dispersed U3Si2-Al, open-tank-in-pool,
equilibrium core, BOC, EOC, shutdown margin.
I. INTRODUCTION
Research reactor has been widely
utilized in various fields such as industry,
engineering,
medicine,
life
science,
environment, etc., and now its application
fields are gradually being expanded together
with the development of its technology. The
utilization of a research reactor is related to the
necessary and essential technologies of
information technology, nano-technology,
biotechnology, environmental technology and
space technology. Hence, R&D in the area of
research reactor utilizations has a large effect
on the growth of a national industry.
has considerable experience in the research
reactor technology through the design,
construction, operation and utilization of the
High-flux Advanced Neutron Application
Reactor (HANARO) of 30 MWth. Therefore,
in the framework of the joint study on the prefeasibility of MRR with KAERI, a model of
Advance HANARO Reactor (abbreviated as
AHR) has been developed to meet the
requirements for use in the future [3,4]. Based
on the model of AHR, a similar reactor model
with plate fuel type MTR (abbreviated as
MTR) has been also developed for the purpose
of comparison between the two fuel types.
II. NUCLEAR DESIGN REQUIREMENTS
Vietnam has a plan to construct a high
performance multipurpose research reactor
(MRR) to satisfy increasing utilization
demands. So, the pre-feasibility studies to build
a new MRR have been set [1,2]. The Korea
Atomic Energy Research Institute (KAERI)
A. General
A research reactor should be designed in
conformity with user's requirements. The
reactor type, power, and core configuration,
systems and the installed experimental
NGUYEN NHI DIEN, HUYNH TON NGHIEM, LE VINH VINH, VO DOAN HAI DANG
facilities depend on the application purposes
and on the construction and operation costs as
well. Hence, a flexible design is an
indispensable feature when considering a
future expansion of its experimental facilities.
1) The neutron flux variation at the irradiation
sites and the nose of the beam tubes should be
stable with a 5% variation regardless of a
loading or unloading of samples.
2) The axial neutron flux gradient in the
reflector region should be within ±20% over a
length of 50 cm.
The major basic principles to develop
models of the conceptual design are as follows.
1) Multipurpose
medium power
research
reactor
with
a
3) The maximum fast and thermal neutron
fluxes at an irradiation site inside the core
should be greater than 1.3x1014 and 4.0x1014
n/cm2-s, respectively. The maximum thermal
neutron flux at the reflector region should be
greater than 4.0x1014 n/cm2-s.
2) High ratio of flux to power
3) High Safety and Economics
4) Sufficient spaces and expandability of the
facility for various experiments
4) The maximum local power peaking factor
should be less than 3.0.
Fundamentally, a research reactor should
be designed to achieve the established safety
objectives such as the IAEA standards. The
nuclear design requirements for the AHR and
MTR are considered in two parts, functional
and performance requirements.
5) The average discharge burn-up of the fuel
assembly should be higher than 50% of the
initial fissile heavy material, U-235.
6) The reactor operating cycle should be
longer than 30 days.
B. Funtional Requirements
The functional requirements aim to
ensure the safety of the reactor and ready to
operate in all conditions.
III. CORE CONCEPT
The basic concepts of the reactor are the
light water cooled and moderated, heavy water
reflected, open-tank-in-pool type research
reactor and 20 MW power cores loaded with
two typical geometric kinds of fuel elements as
rod or flat plate.
1) The power coefficient and temperature and void
coefficients of the reactivity should be negative for
all operational and accident conditions.
2) The shutdown margin should be at least 10
mk (1mk = k/k  1/1000) regardless of any
changes in the reactor condition.
A. Fuel
3) The second reactor shutdown system should
be prepared to improve the reactor safety and
its shutdown margin should be at least 10 mk
for all relevant design basis fault sequences.
Fuels selected for the design are
commercial or commercial available. The fuel
meat is fabricated by a dispersion of high
density U3Si2 particles into pure Al with its
uranium enrichment 19.75 wt%. Two kinds of
fuel assemblies in the core are standard fuel
assembly and control fuel assembly (including
control rods inside fuel assembly). Some
specifications of the fuel elements and
assemblies are listed in Table I and their cross
sectional views are showed in Figure 1.
4) The excess reactivity should be at least 10
mk at the end of cycle for conducting
experiments and 15 mk for the Xe override.
C. Performance requirements
The performance requirements aim to
ensure meeting the requirements of use and
high economic efficiency.
27
CONCEPTUAL NUCLEAR DESIGN OF A 20 MW MULTIPURPOSE RESEARCH REACTOR
Table I. Specifications of the fuel element and assembly
Fuel element
Meat content
Fuel length (mm)
Fuel diameter/widththickness (mm)
66.0w/%U, 5.2w/%Si,
28.8w/%Al
700.0
6.35/5.49 (In/Out)
72.8w/%U, 6.0w/%Si,
21.2w/%Al
700.0
64.0/51.4x0.61 (S/C)*
6.06
0.76/1.19 (In/Out)
Al
6.6
0.37/0.445 (In/Out)
Al
Hexagonal
36/18 (S/C)
Square
21/17 (S/C)
Fuel density (g/cm3)
Cladding thickness (mm)
Cladding material
Fuel assembly
Shape
Element number
* S/C: Standard fuel assembly / Control fuel assembly
a) AHR standard
b) AHR Control
c) MTR standard
d) MTR control
Fig. 1. Cross sectional view of AHR and MTR standard and control fuel assemblies
B. Core Arrangement
core. The reactor regulating system shares
control rods with the reactor protection system.
Fig. 2 shows the horizontal cross sectional
view of the AHR and MTR cores. Some
specifications of the cores are listed in Table II.
The core has 23 lattices that consist of
fourteen standard assemblies, four control
assemblies and three in-core irradiation sites.
The heavy water reflector tank of 200 cm in
diameter and 120 cm in height surrounds the
Fig. 2. The horizontal cross sectional view of the AHR and MTR cores
28
Table II. The specifications of the cores
Reactor type
Core volume (cm3)
Fuel assembly Number
Control rod Number
Absorber material
Total weight U-235 (kg)
In-core irradiation sites
AHR
1199.5 x 70
16 S + 4 C
4
Hf
9,87
3
MTR
1527.7 x 70
16 S + 4 C
4
Hf
10,12
3
Fig. 3. The layout of the experimental sites of the AHR and MTR
IV. NUCLEAR ANALYSIS
distribution, the reactivity of the core and the
reactivity worth of control rods were also
assessed to meet the requirements. Two core
configurations with one and three in-core
irradiation sites were proposed. Although the
first configuration (with one irradiation site) is
better in the fuel saving point of view, the
configuration with three in-core irradiation
sites was selected to meet predicted utilization
of in-core irradiation in the future.
To confirm that the conceptual cores
satisfy the functional and performance
requirements, nuclear analyses are performed
for fresh core and equilibrium core with
several code systems such as MCNP [5], MVP
[6], HELIOS [7], etc.
A. Fresh Core
The basic analysis of the core
characteristics was performed for the fresh core
with and without irradiation facilities.
As the ultimate goal of a research reactor
is its utilization, the irradiation facilities should
be designed in conformity with the user's
requirements. The required irradiation facilities
should be located at proper positions to
maximize neutron utilization and minimize
reactivity effect. Based on the neutron flux
distribution of the reflector region, the
arrangement by their purposes has been studied
to achieve the objectives above. Their
The core configuration should be
designed to meet the functional and
performance requirements. The neutron flux at
the in-core irradiation sites and the reflector
region of the cores without irradiation facilities
was calculated by the MCNPX code [8] using a
mesh tally. On the other hand, the power
29
reactivity worth is considered as a priority
because of the influence to the reactor core.
Various layouts of the irradiation facilities
were proposed, and one of them was selected.
To evaluate the stability of neutron flux at the
irradiation sites, their neutron fluxes were
calculated when the control rods are located at
300 mm and fully withdrawn.
The reactivity effect by the irradiation
facilities was estimated to be 20.2 mk and 28.9
mk and the total control rods worth 182.4 mk
and 217.7 mk for AHR and MTR, respectively.
Table III shows the neutron fluxes at the
irradiation facilities. Figure 4 presents the
thermal and fast neutron distribution of the
AHR fresh core.
Table III. Neutron fluxes at the experimental sites
Neutron flux [n/cm2/sec]( Thermal<0.625eV, Fast>1.0MeV)
AHR
MTR
Maximum
Thermal
Fast
Average
Thermal
Maximum
Fast
Thermal
Fast
Average
Thermal
Fast
CT 4.46E+14 1.46E+14 3.04E+14 9.80E+13 4.01E+14 1.13E+14 2.87E+14 8.06E+13
IR1 3.21E+14 1.18E+14 2.23E+14 8.29E+13 3.37E+14 9.31E+13 2.49E+14 6.76E+13
IR2 3.16E+14 1.20E+14 2.23E+14 8.26E+13 3.33E+14 9.16E+13 2.46E+14 6.65E+13
CNS 8.71E+13 1.15E+12 7.01E+13 8.76E+11 8.49E+13 1.69E+12 6.48E+13 1.24E+12
ST1 1.37E+14 1.96E+12
-
-
1.40E+14 3.23E+12
-
-
ST2 2.40E+14 3.47E+12
-
-
1.79E+14 1.01E+13
-
-
NR 1.28E+14 3.20E+11
-
-
1.28E+14 1.32E+12
-
-
NTD1 4.74E+13 1.13E+11 4.31E+13 8.12E+10 4.93E+13 4.19E+11 4.26E+13 3.19E+11
NTD2 4.63E+13 9.91E+10 4.24E+13 7.60E+10 5.29E+13 4.84E+11 4.57E+13 3.56E+11
NTD3 5.16E+13 2.43E+11 4.70E+13 2.04E+11 4.64E+13 5.21E+11 3.93E+13 3.78E+11
HTS1 6.96E+13 3.42E+11 5.96E+13 2.71E+11 7.02E+13 6.30E+11 5.79E+13 5.03E+11
HTS2 2.23E+13 2.07E+10 1.93E+13 1.39E+10 2.25E+13 2.81E+10 1.97E+13 2.29E+10
NAA1 1.39E+14 4.96E+11 1.20E+14 3.88E+11 1.22E+14 8.15E+11 1.05E+14 6.27E+11
NAA2 4.11E+13
-
3.59E+13
-
4.00E+13
-
3.55E+13
-
NAA3 1.74E+13
-
1.52E+13
-
1.53E+13
-
1.35E+13
-
RI1 3.53E+14 1.47E+13 2.60E+14 9.05E+12 2.31E+14 1.49E+13 1.69E+14 9.28E+12
RI2 3.44E+14 1.42E+13 2.57E+14 8.91E+12 2.18E+14 1.47E+13 1.58E+14 9.13E+12
RI3 2.46E+14 4.03E+12 1.85E+14 2.54E+12 2.10E+14 1.53E+13 1.58E+14 9.56E+12
RI4 2.48E+14 4.23E+12 1.86E+14 2.82E+12 2.03E+14 1.45E+13 1.52E+14 8.92E+12
RI5 2.24E+14 3.12E+12 1.67E+14 2.10E+12 2.15E+14 1.55E+13 1.58E+14 9.51E+12
30
a) Thermal Neutron Flux
b) Fast Neutron Flux
Fig. 4. Neutron flux profile at the AHR fresh core
B. Equilibrium Core
6-batch core) are assessed. The 9-batch cores
show a high discharge burnup and a good fuel
economy, but the cycle lengths are less than 30
days. They look proper for a low utilization
condition of the reactor. The 6-batch cores with
a cycle length greater than 30 days are suitable
for the design requirements, so they are
selected for evaluating in detail. In the 6-batch
core, three of the standard fuel assemblies or
two of the standard fuel assemblies and two of
the control fuel assemblies are replaced for an
operation cycle, so the whole core will be
replaced for 6 cycles according to the loading
strategy. There are many loading patterns that
they depend on the fuel management strategy.
The loading pattern showed in Table IV is
evaluated in detail.
An equilibrium core is dependent on an
operation strategy, so there may be various
equilibrium cores according to a reactor
operating strategy. In this report, an
equilibrium core is proposed and analyzed to
meet the established design requirements.
Fuel Management
A candidate model for an equilibrium
core can be easily obtained by considering
target discharge burnup, cycle length and
excess reactivities at begin of cycle (BOC) and
end of cycle (EOC). There are many candidate
models according to the number of reloaded
fuel assemblies and the loading pattern. The
equilibrium cores with 2 or 3 fuel assemblies
reloaded for an operation cycle (the 9-batch or
Table IV. Loading location of the fuel assemblies for 6-batch cores
Cycle
Assembly Number
(standard+control)
AHR
Loading Location
MTR
1
2+2
H14,H16,C1,C3
H9,H12,C1,C3
2
3+0
H8,H10,H12
H14,H15,H7 (move H14,H15,H7 to
H2,H4,H6)
31
3
3+0
H7,H9,H11
4
2+2
H13,H15,C2,C4
5
3+0
H2,H4,H6
6
3+0
H1,H3,H5
H13,H10,H16 (move H13,H10,H16 to
H3,H5,H1)
H8,H11,C2,C4
H14,H15,H7 (move H14,H15,H7 to
H2,H4,H6)
H13,H10,H16 (move H13,H10,H16 to
H3,H5,H1)
Once a cycle length and a loading
pattern are determined, an equilibrium core is
obtained by numerical iterations. The initial
core is loaded with the new FAs then the
burnup calculations are iterated by the loading
pattern until the parameters of burnup and
reactivity are stable over 6 cycles. Table V
presents the calculated results of the average
burnup and reactivity of 6 cycles for different
cycle lengths. From these results, it can be
concluded that the 36 days cycle for AHR and
34 days cycle for MTR meet the performance
requirements.
Table V. Burnup and reactivity of the equilibirum cores
Reactor type
Cycle Length (days)
AHR
MTR
35
36
37
33
34
35
- BOC
23.43
24.02
24.61 22.38 23.04 23.70
- EOC
31.82
32.65
33.47 29.08 29.94 30.81
- Discharge
50.35
51.77
53.18 48.65 49.91 51.17
- BOC (no Xe)
111.9
109.9
107.8
87.8
85.8
83.6
- Fuel Depletion
37.5
38.7
39.9
15.1
16.7
18.3
- Xenon Buildup
38.1
38.1
38.0
36.2
36.3
36.3
- Power Defect
3.0
3.0
3.0
3.0
3.0
3.0
- EOC (eq. Xe)
33.4
30.1
26.9
33.5
29.8
26.0
- Shutdown Margin
15.0
17.1
19.6
22.2
24.2
26.4
Average Burnup (%U-235)
Reactivity (mk)
Power Distribution
The power distribution is strongly
dependent on the positions of the control rods
and it was checked for all possible positions at
5 cm intervals. The largest maximum linear
power of the equilibrium cores was observed at
a 300 mm position of the control rods. The
power distribution for the equilibrium cores of
30
32
6 cycles at a 300 mm position of the control
rods was calculated. Table VI shows maximum
total peaking factors for the 6 cycles equilibrium
cores and Table VII shows the power
distributions and peaking factors at the cycle
that total peaking factor reaches the maximum
value. The maximum local power peaking factor
for AHR and MTR are 2.56 and 2.79
respectively.
Table VI. Maximum total peaking factor for the equilibrium cycles
Reactor
type
AHR
Cycle
Parameter
1
2
3
4
5
6
Position of FA
H02
C2
H01
H03
C3
H04
Fq(peaking factor)
2.47
2.56
2.5
2.46
2.56
2.49
Vị trí FA
Fq
H09
2.69
H04
2.77
H01
2.74
H11
2.76
H04
2.76
H01
2.79
MTR
Table VII. Power distribution and peaking factor for equilibrium cores
(cycle 5 for AHR, cycle 6 for MTR)
Location
H01
H02
H03
H04
H05
H06
H07
H08
H09
H10
H11
H12
H13
H14
H15
H16
C1
C2
C3
C4
AHR
Total Power (kW)
1117
1061
1314
897
1064
1305
811
1071
1329
1063
1088
1298
1224
991
1184
1050
577
491
589
476
MTR
Total Power (kW)
1167
1158
1229
1092
1223
1143
1033
1179
985
1066
1213
987
1137
1093
1170
1174
445
527
449
531
Fq
1.93
1.72
2.21
1.58
1.73
2.18
1.38
1.64
2.37
1.76
1.65
2.29
1.78
1.4
1.57
1.42
2.44
2.04
2.56
1.95
Reactivity Coefficients
Fq
2.79
1.82
1.96
2.51
2.02
1.82
2.12
2.21
1.67
2.31
2.26
1.66
2.01
1.91
2.05
1.96
1.28
1.55
1.29
1.56
region is to cool the fuel assemblies, and so
called a ‘coolant’ and the light water in the
gaps of the flow tubes is called a ‘moderator’.
Nuclear characteristics of these two light water
regions are somewhat different, and a heat
transfer between them is small. Therefore, their
temperature variations following a power
change are also different, thus the respective
temperature coefficients were computed
separately. The effect of a spectrum hardening
To affirm the inherent safety, the
reactivity coefficients should be determined.
They include temperature coefficients of fuel,
light water and heavy water. Physical changes
of water due to a temperature change could be
considered in two ways: one is a density
change, and the other is a cross section change
for a nuclear reaction. There are the gaps of the
flow tubes for AHR. The light water in the fuel
33
of neutrons following a temperature increase
for heavy water is so small that it can be
negligible. Table VIII presents the result of
temperature and void coefficients. From this
result, they are negative (except temperature
coefficient of moderator. where almost of
arriving neutrons are slowed down) and meet
the functional requirements. The temperature
variation of moderator is so small, therefore its
contribution to power coefficient is small.
Table VIII. Reactivity coefficients of temperature and void
Parameter
Fuel temperature coefficient (mk/K)
Light water temperature coefficient (mk/K)
- Coolant
- Moderator
Light water void coefficient (mk/%)
0- 5%
5 - 10 %
10 - 20 %
Heavy water void coefficient (mk/%)
0- 5%
V. CONCLUDING REMARKS
AHR
MTR
<-0.002 <-0.02
-0.059
0.06
-0.11
-1.23
-1.37
-1.48
-1.79
-1.97
-2.25
-1.26
-0.79
the temperature coefficients are negative
showed the inherent safety feature. The
parameters for utilization and for the safety
aspects of the reactor respectively meet the
performance and functional requirements.
From the functional and performance
requirements, two reactor models AHR and
MTR were proposed and investigated. The
reference reactors are the light water cooled
and moderated, heavy water reflected and
open-tank-in-pool type research reactors with a
20 MW power.
The comparison of cores loaded with 2
different fuel types, AHR and MTR, shows that
the AHR fuel type core has a little longer
operation cycle and higher discharge burn up as a
result. In the safety point of view, the MTR core
has an advantage because of shutdown margin,
temperature and coolant void coefficients are
higher compared to those of AHR core.
The maximum fast and thermal neutron
flux in the core region are greater than 1.0×1014
n/cm2s and 4.0×1014 n/cm2s, respectively. In
the reflector region, the thermal neutron peak
occurs about 28 cm far from the core center
and the maximum flux is estimated to be
4.0×1014 n/cm2s.
REFERENCES
[1] Luong Ba Vien and C. Park et.al., Joint
KAERI/VAEC pre-possibility study on a new
research reactor for Vietnam, KAERI/TR2756/2004, (May, 2004).
For the equilibrium cores, the cycle
length is greater than 30 days, the whole core
will be replaced for 6 cycles, and the assembly
average discharge burnup is greater than 50%.
For the proposed fuel management scheme, the
maximum peaking factor Fq is less than 3. The
shutdown margins by the 1st and 2nd
shutdown systems are greater than 10 mk and
[2] Nguyen Nhi Dien et al., Report on Study Project
No BO/06/01-04, (in vietnamese), (2008).
[3] Seo Chul Gyo, Huynh Ton Nghiem et al.,
Conceptual Nuclear Design of a 20 MW
34
Multipurpose Research Reactor - KAERI/TR3444/(2007).
[7] E. A. Villarino, R. J. J. Stamm'ler, A. A. Ferri,
J. J. Casal, HELIOS: Angularly Dependent
Collision Probabilities, Nucl. Sci. & Eng., 112,
16, (1992).
[4] Hee TaekChae, Le Vinh Vinh et al., Conceptual
Thermal Hydraulic Design of a 20MW
Multipurpose Research Reactor - KAERI /TR3443/(2007).
[8] Denise B. Pelowitz (Editor), MCNPX User's
Manual, LA-CP-05-0369, Los Alamos
National Lab, (2005).
[5] J. F. Briesmeister (Editor), MCNP-A General
Monte Carlo N-Particle Transport Code, LA12625-M, Los Alamos National Lab, (1993).
[6] Yasunobu NGAYA et al., MVP/GMVP II:
General Purpose Monte Carlo Codes for
Neutron and Photon Transport Calculations
based on Continuous Energy and Multigroup
Methods, JAERI 1348, (2005).
35
Some Main Results of Commissioning
of The Dalat Research Reactor with Low Enriched Fuel
Nguyen Nhi Dien, Luong Ba Vien, Pham Van Lam, Le Vinh Vinh, Huynh Ton Nghiem,
Nguyen Kien Cuong, Nguyen Minh Tuan, Nguyen Manh Hung, Pham Quang Huy,
Tran Quoc Duong, Vo Doan Hai Dang, Trang Cao Su, Tran Tri Vien
Nuclear Research Institute –Vietnam Atomic Energy Institute
01-Nguyen Tu Luc, Dalat, Vietnam
(Received 5 March 2014, accepted 13 April 2014)
Abstract: After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for
conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the
commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried
out from 24 November 2011 to 13 January 2012. The experimental results obtained during the
implementation of commissioning programme showed a good agreement with design calculations and
affirmed that the DNRR with LEU core have met all safety and exploiting requirements.
Keywords: HEU, LEU, physics start up, energy start up, effective worth, Xenon poisoning, Iodine pit.
I.
INTRODUCTION
Physics and energy start-up of the Dalat
Nuclear Research Reactor (DNRR) for full
core conversion to low enriched uranium
(LEU) fuel were performed from November
24th, 2011 until January 13th, 2012 according to
an approved program by Vietnam Atomic
Energy Institute (VINATOM). The program
provides specific instructions for manipulating
fuel assemblies (FAs) loading in the reactor
core and denotes about procedures for carrying
out measurements and experiments during
physics and energy start-up stages to guarantee
that loaded LEU FAs in the reactor core are in
accordance with calculated loading diagram
and implementation necessary measurements
to ensure for safety operation of DNRR.
Main content of the report is a brief
presentation of performed works and achieved
results in the physics and energy start up stages
for DNRR using LEU fuel assemblies, that is
from starting loading LEU fuel to the reactor
core (November, 24th, 2011) until finishing 72
hours testing operation without loading at
nominal power (December, 13rd, 2011).
II. PHYSICS START UP
Physics startup of reactor is the first
phase of carrying out experiments to confirm
the accuracy of design calculated results,
important physical parameters of the reactor
core to meet safety requirements. Physics
startup includes fuel loading gradually until to
approach criticality, loading for working core
and implementing experiments to measure
parameters of the core at low power such as
control rods worth, shutdown margin,
temperature effect,…
A. Fuel loading to approach criticality
The loading of LEU FAs to the reactor
core was started on November, 24th, 2011
following a predetermined order in which each
step loaded one or group LEU FAs to the
reactor core. After each step, the ratio of
N0
(N0 is initial number of neutron count rate,
Ni
LUONG BA VIEN et al.
Ni is that to be obtained after step ith) was
for working core, effective worth of loaded
fuel assemblies and shutdown margin were
preliminarily evaluated to ensure shutdown
margin limit not be violated. Fig. 3 shows the
current working core of DNRR, including 92
LEU FAs (80 fresh LEU FAs and 12 partial
burnt LEU FAs, the burn up about 1.5 to 3.5
%) and neutron trap at the center. Total mass of
U-235 that was loaded to the reactor core is
about 4246.26 g. Shutdown margin (or
subcriticality when 2 safety rods are fully
withdrawn) is 2.5 $ (about 2% k/k), smaller
than calculated value (3.65 $) but still
completely satisfy the requirement >1% for
the DNRR. Excess reactivity of the core
configuration is about 9.5 $, higher than
calculated value (8.29 $), ensuring operation
time of the reactor more than 10 years with
recent exploiting condition. So, it can be said
that the current working core meets not only
safety requirements but reactor utilization
also (ensure about shutdown margin and
sufficient excess reactivity for reactor
operation and utilization).
evaluated to estimate critical mass. At 15h35
on November, 30th, 2011 the reactor reached
critical status with core configuration including
72 LEU FAs and neutron trap in center (see
Fig. 1 and 2).
Established critical core configuration
with 72 LEU FAs having neutron trap is in
good agreement with design calculated
results. With 72 LEU FAs, by changing
position of some fuel assemblies, all new
criticality conditions were achieved with
lesser inserting position of regulating rod. It is
concluded that the above critical configuration
(Fig. 1) is the minimum one among
established configurations. The critical mass
of Uranium is 15964.12 g in which Uranium235 is 3156.04 g.
B. Fuel loading for the working core
After completion of fuel loading to
approach criticality, fuel loading for working
core was carried out from December, 6th, 2011
to December, 14th, 2011. During fuel loading
Fig. 1. Critical core configuration and
order of loaded fuel assemblies
Fig. 2. N0/Ni ratio versus number of FAs
loading to the core
37
SOME MAIN RESULTS OF COMMISSIONING OF …
LEU Fuel
Wet channel
Berrylium
Dry channel
Aluminum
Empty cell
Neutron trap
Fig. 3. Working core configuratiom with 92 LEU FAs
C. Performed experiments in the working core
configuration
working core in configuration with 82 fresh
LEU FAs and 92 LEU FAs.
Determination of control rod worth
Control rods worths and integral
characteristics in core configuration with 92
LEU FAs are presented in Table I, Fig. 4
and 5. Measured results were smaller than
design calculated results about 12% in
average.
To calibrate control rod worth, doubling
time method was applied for regulating rod
while reactivity compensation method was
used for shim rods and safety rods. The
calibration of control rods of DNRR were
implemented two time during fuel loading for
Table I. Effective worth of regulating rod, 4 shim rods and 2 safety rods
in core configuration with 92 LEU FAs.
Effective reactivity ($)
Control Rod
Measured value
Calculated value
Regulating rod
0.495
0.545
Shim rod 1
2.966
3.237
Shim rod 2
3.219
3.263
Shim rod 3
2.817
3.473
Shim rod 4
2.531
3.086
Safety rod 1
2.487
2.744
Safety rod 2
2.195
2.795
38
Reactivity ($)
Reactivity ($)
Position (mm)
Position (mm)
Fig. 5. Integral characteristics of 4 shim rods
Fig. 4. Integral characteristics of regulating rod
Thermal neutron flux distribution
measurement in the reactor core
From the measured results, it can be
seen that the maximum peaking factor of
1.49 is achieved at outer corner of hexagonal
tube of the fuel assembly in cell 6-4. Neutron
distribution of working core has large
deviation from North (thermal column) to
South (thermalizing column). Neutron flux
in southern region of the core (cell 12-1 and
12-7) is about 28 % smaller than those in
Northern region (cell 2-1 and 2-7). The
asymmetry of the reactor core has reason
from the not identical reflector that was noted
from the former HEU fuel core.
Measurement of thermal neutron flux
distribution following axial and radial in the
reactor core was carried out by Lu metal foils
neutron activation. A number of positions in
the reactor core were chosen to measure
thermal neutron flux distribution including
neutron trap, irradiation channels 1-4 and 13-2,
and 10 FAs at the cells: 1-1, 2-2, 2-3, 2-7, 3-3,
3-4, 4-5, 6-4, 12-2 and 12-7. Figs 6 to 9 present
the measured results of axial and radial neutron
flux distributions of the reactor core.
1.1
1.1
1
1
0.9
0.9
FA cell 2-3
0.8
0.8
FA cell 3-3
0.7
0.7
Relative Unit
FA cell 4-5
Relative Unit
0.6
0.5
-30
-25
-20
-15
-10
-5
0.5
0.4
0.4
0.3
0.3
0.2
0.2
0.1
0.1
0
0
-35
0.6
0
5
10
15
20
25
30
0
35
5
10
15
20
25
30
35
40
45
50
55
60
Position from the bottom to the top (cm)
Position from the bottom to the top (cm)
Fig. 6. Axial thermal neutron flux distribution in
the fuel assemblies
Fig. 7. Axial thermal neutron flux distribution in the
neutron trap
39
Fig. 8. Thermal neutron flux distribution of FAs
and irradiation positions in comparison with
neutron trap.
Fig. 9. Thermal neutron flux distribution of the
FA’s corners in comparison
with neutron trap
Determination of effective worth of FAs,
beryllium rods and void effect
determined by comparing position change of
control rods before and after withdrawing FA
or beryllium rod or before and after inserting
watertight aluminum tube. Reactivity worth
values were obtained using integral
characteristics curves of control rods.
The measurements of effective worth of
FAs, beryllium rods and void effect (by
inserting an empty aluminum tube with
diameter of 30 mm) were also performed.
These are important parameters related to
safety of the reactor. Positions for
measurement of effective reactivity of FAs,
Be rods and void effect were chosen to
examine the distribution, symmetry of the
core and the interference effects at some
special positions. Effective reactivity of FAs,
beryllium rods and void effect were
Figs 10÷12 show the measured results
of effective worth of 14 FAs in the reactor
core at different positions; effective worth of
beryllium rods around neutron trap and a
new beryllium rod at irradiation channel 1-4;
void effect at neutron trap, irradiation
channel 1-4 and cell 6-3, which surrounded
by other FAs.
Fig. 10. Effective worth of FAs in the reactor core
Fig. 11. Effective worth of Be rods in the reactor core
40
Reactivity ($)
.
Temperature (0C)
Fig. 13. Negative reactivity insertion
dependent on pool water temperature
temperature
Fig. 12. Measured results of void effect at
some positions in the reactor core
The most effective worth of fuel
assembly measured at cell 4-5 is 0.53 $.
Measured results of effective reactivity of fuel
assemblies and Be rods show a quite large
tilting of reactor power from North to South
direction. Void effect has negative value in the
reactor core (cell 1-4 and 6-3) while positive in
the neutron trap. Void effect in neutron trap
has positive value because almost neutrons
coming in neutron trap are thermalized, that is
absorption effect of water in neutron trap is
dominant compared to moderation effect. The
replacement of water by air or decreasing of
water density when increasing steadily of
temperature introduces a positive reactivity.
With the core using HEU fuel also has positive
reactivity of void in neutron trap.
established after each increased step of pool
water temperature about 2.50C. Basing on the
change of regulating rod position (due to
change of temperature in the reactor core) the
temperature coefficient of moderator was
determined.
Heating process of water in reactor pool
by operating primary cooling pump took long
time so water in neutron trap also heated up
and inserted positive reactivity (as explanation
in measurement of void effect), as opposed to
temperature effect in the reactor core. So, a
hollow stainless steel tube 60 mm diameter
was inserted in neutron trap to eliminate
positive temperature effect of neutron trap.
Fig. 13 shows measured results of
temperature coefficient of moderator with
initial temperature of 17.7 oC. Based on these
results, the temperature coefficient of
moderator is determined about -9.110-3 $/oC.
Measured result without steel pipe containing
air at neutron trap was about -5.210-3 $/oC.
Thus, temperature coefficient of moderator
including neutron trap still has negative value.
Temperature coefficient of moderator of the
core loaded with 88 HEU FAs measured in
1984 was -8.010-3 $/oC.
Determination of temperature coefficient
of moderator
Temperature coefficient of moderator is
the most important parameter, demonstrating
inherent safety of reactor. To carry out
experiment, the temperature inside reactor pool
was raised about 100C by operating primary
cooling pump without secondary cooling
pump. To measure temperature coefficient of
moderator, criticality of the reactor was
41
III. ENERGY START-UP
13-2 and rotary specimen was measured by
using Au foil activation method. Also, on
January 17th, 2012 thermal neutron flux of
positions mentioned above was measured at
power level 100%. Measured results of thermal
neutron flux at several irradiation positions in
the reactor core with different power levels are
presented in Table II.
A. Power ascension test
On January 6th, 2012 reactor power has
been increased at levels of 0.5% nominal
power, 10% nominal power and 20% nominal
power. At each power level, thermal neutron
flux in neutron trap, irradiation channels 1-4,
Table II. Measured results of thermal neutron flux at several irradiation positions at different reactor power levels
Irradiation
positions
Neutron trap
0,5
1.143E+11
Power (% Nominal power)
10
20
2.063E+12
4.174E+12
100
2.122E+13
Channel 1-4
5.288E+10
9.719E+11
1.965E+12
8.967E+12
Channel 13-2
4.749E+10
8.542E+11
1.682E+12
N/A
Rotary Specimen
N/A
N/A
N/A
4.225E+12
Based on the reactor power determined
by thermal neutron flux measurements at low
power levels, on January 9th, 2012 the reactor
was ascended power: 0.5%, 20%, 50%, 80%
and then operated at 80% nominal power
during 5 hours for determination of thermal
power and examination of technological
parameters and gamma dose before raising the
reactor power to nominal level.
system parameters was about 372 kW. This
value enables us to raise the reactor power to
full power level. 15h32 on January, 9th, 2012 the
reactor was raised to 100% nominal power and
maintained at this power about 65 hours before
decreasing to 0.5% nominal power to measure
Xenon poisoning transient. Table III presents
the values of thermal power of the reactor
during the first 8 hours after the reactor reached
100% nominal power. The data indicate that
thermal power is just only about 460 kW, lower
than design nominal power about 10%.
Thermal power of the reactor
corresponding to 80% nominal power level
(based on indication of control system) after 5
hours calculated based on primary cooling
Table III. Thermal power of the reactor with operation time after the reactor reached 100% nominal power
Time
15h30
16h00
17h00
18h00
1h00
20h00
21h00
22h00
23h00
24h00
Tin (1)
[oC]
29,2
30,3
31,0
31,0
30,9
30,8
30,8
30,7
30,6
30,5
Tout (1)
[oC]
22,4
22,9
23,1
23,0
22,9
22,9
22,9
22,8
22,7
22,6
42
GI
[m3/h]
49,4
49,3
49,8
49,8
49,8
49,6
49,8
50,5
50,1
49,6
PI
[kW]
390
423
456
462
462
454
456
457
459
455
B. Xenon poisoning transient and Iodine hole
C. Power adjustment
The experiment to determine the curve
built up of Xenon poisoning and then
calculating its equilibrium poisoning was
conducted from January 9th, 2012 to January
12th, 2012 when the reactor was in 100%
nominal power (indicating of control system
without adjusting power) . Next, Iodine hole
was also determined from 12 to January 13th,
2012 after reducing power of the reactor from
100% to 0.5% nominal power by monitoring
the shift position of regulating rod.
In the process of gradually raising power
in energy start-up, although power indication
on control system was 100% but calculated
thermal power of the reactor through flow rate
of primary cooling system and difference
between inlet and outlet temperatures of the
heat exchanger was only 460 kW, smaller than
nominal power about 10%. The reason was
mainly due to power density of the core using
92 LEU FAs were higher than the mixed core
using 104 FAs before. The adjustment to
increase thermal power of the reactor was
performed by changing the coefficients on the
control panel. After adjusting, the reactor was
operated to determine thermal power at power
setting 100%. The results of thermal power
obtained from the next long operation was
about 510.5 kW. This value includes 500 kW
thermal power of the reactor and about 10 kW
generated by primary cooling pump.
Fig. 14 presents measured results of
Xenon poisoning curve and Iodine pit of the
above
experiment.
Xenon
equilibrium
poisoning and other effects is totally about -1.1
eff and the maximum depth of Iodine pit
determined about -0.15 eff after 3.5 hours
since the reactor was down to 0.5% nominal
power. After adjusting thermal power up to
500 kW, during the long operation from
March, 12-16, 2012, after the reactor was
operated 68 hours at nominal power, total
value of poisoning and temperature effects is
about -1.32 eff.
D. Measurement of neutron flux and neutron
spectrum after power adjustment
Negative reactivity ($)
After carrying out reactor power
adjustment, thermal neutron flux at some
Xe
poisonning
Iodine Pit
PpppPithole
Time (hour)
Fig. 14. Negative reactivity insertion by Xenon poisoning with operation time and Iodine pit
43
irradiation positions in the reactor core and
neutron spectrum in neutron trap were
measured again by neutron activation foils.
Measured maximum neutron flux at neutron
trap was 2.23  1013 n/cm2.s (compared with
calculated result was 2.14÷2.22  1013 n/cm2,
depending on shim rods position). Those in
channel 1-4 and 13-2 were 1.07  1013 n/cm2.s
and 8.611012 n/cm2.s,
respectively. The
experimental error of neutron flux was
estimated about 7%.
start up were carried out successfully. DNRR
was reached criticality at 15:35 on November,
30th, 2011 with 72 LEU FAs, consistent with
calculated results. Then, the working core with
92 LEU FAs has been operating 72 hours for
testing at nominal power during from January,
9th, 2012 to January, 13th, 2012.
Experimental results of physical and
thermal hydraulics parameters of the reactor
during start up stages and long operation cycles
at nominal power showed very good agreement
with calculated results. On the other hand,
experimental results of parameters related to
safety such as peaking factor, axial and radial
neutron flux distribution of reactor core,
negative temperature coefficient, temperature
of the reactor tank, temperature at inlet/outlet
of primary cooling system and secondary
cooling system,…it could be confirmed that
current core configuration with 92 LEU FAs
meets the safety and exploiting requirements.
From reaction rate measured by foils
irradiation method in neutron trap, neutron
spectrum obtained by SAND-BP computer
code. Obtained results of neutron spectrum in
neutron trap (Fig. 15) showed that comparing
with mixed-core HEU-LEU fuel, when
neutron trap having thinner Beryllium layer,
thermal neutron flux increased while epithermal and fast neutron flux decreased with a
significant percentage.
Measured neutron flux at irradiation
positions and actual utilization of the
reactor after full core conversion also
showed that the reactor core using LEU fuel
is not much different than previous core
using HEU fuel.
IV. CONCLUSIONS
Neutron flux/Lethagy, n/cm2.sec
After completing design calculation and
preparation, start up of DNRR with entire LEU
FAs core was implemented following a
detailed plan. As a result, physics and energy
Fig. 15. Measured neutron spectrum in neutron trap before and after conversion
44
ACKNOWLEDGMENTS
REFERENCE
The NRI’s staffs that performed start up
work of DNRR with entire LEU fuel core
would like to express sincere gratitude to the
leadership of Ministry of Science and
Technology, Vietnam Atomic Energy Institute,
Vietnam Agency for Radiation and Nuclear
Safety, who have regularly regarded, guided
and created the best condition for us to
implement our works. We also express our
thanks to Argonne National Laboratory and
experts from RERTR program (Reduced
Enrichment for Research and Test Reactors)
and specialists, professionals in program
RRRFR (Russian Research Reactor Fuel
Return) has supported in finance as well as
useful discussions during design calculation of
full core conversion, upgrading equipments
and carrying out start up of DNRR.
[1] P. V. Lam, N. N. Dien, L. V. Vinh, H. T.
Nghiem, L. B. Vien, N. K. Cuong, “Neutronics
and Thermal Hydraulics Calculation for Full
Core Conversion from HEU to LEU of the
Dalat Nuclear Research Reactor”, RERTR Int’l
Meeting, Lisbon, Portugal, 2010.
[2] L. B. Vien, L. V. Vinh, H. T. Nghiem, N. K.
Cuong, “Transient Analyses for Full Core
Conversion from HEU to LEU of the Dalat
Nuclear Research Reactor”, RERTR Int’l
Meeting, Lisbon, Portugal, 2010.
[3] “Process of physics and energy start up for full
core conversion using LEU fuel of the Dalat
Nuclear Research Reactor”, Nuclear Research
Institute, 2011.
[4] “Operation logbook of DNRR”, 2011-2012
45
Production of Radioisotopes and Radiopharmaceuticals
at the Dalat Nuclear Research Reactor
Duong Van Dong, Pham Ngoc Dien, Bui Van Cuong, Mai Phuoc Tho,
Nguyen Thi Thu, Vo Thi Cam Hoa
Nuclear Research Institute,
01 Nguyen Tu Luc Street, Dalat City, Vietnam
Abstract: After reconstruction, the Dalat Nuclear Research Reactor (DNRR) was inaugurated on
March 20th, 1984 with the nominal power of 500 kW. Since then the production of radioisotopes and
labelled compounds for medical use was started. Up to now, DNRR is still the unique one in Vietnam.
The reactor has been operated safely and effectively with the total of about 37,800 hrs (approximately
1,300 hours per year). More than 90% of its operation time and over 80% of its irradiation capacity
have been exploited for research and production of radioisotopes. This paper gives an outline of the
radioisotope production programme using the DNRR. The production laboratory and facilities
including the nuclear reactor with its irradiation positions and characteristics, hot cells, production
lines and equipment for the production of Kits for labelling with 99mTc and for quality control, as well
as the production rate are mentioned. The methods used for production of 131I, 99mTc, 51Cr, 32P, etc.
and the procedures for preparation of radiopharmaceuticals are described briefly. Status of utilization
of domestic radioisotopes and radiopharmaceuticals in Vietnam is also reported.
Keywords: Radioisotope; Radiopharmaceutical; Labelled KIT, Nuclear Medicine.
I.
INTRODUCTION
During the last 30 years of operation, the
DNRR has been successfully used for producing
many
kinds
of
radioisotopes
and
radiopharmaceuticals used in medicine and other
economic and technical fields. Providing about
400Ci per year of radioisotopes including I-131,
P-32, Tc-99m generator, Kit in-vivo and in-vitro,
Sr-46, Cr-51, etc. Each year, about 300,000
patients have been diagnosed and treated by
radioisotopes produced at DNRR that contributed
to push forward the development of nuclear
medicine in Vietnam.
In a developing country of low economic
level, the benefit of establishment of a nuclear
research center with a research reactor of low
power will be recognized by society only when
its contributions to social progress become
evident. This point of view has oriented us to
put forward a limited radioisotope production
programme to support radioisotope application
in medicine, agriculture and industry. For this
objective the core of the present 500-kW
reactor reconstructed from the previous 250kW TRIGA MARK II reactor is equipped with
more neutron irradiation channels and with a
neutron trap for improving thermal neutron
flux. In addition, the reactor characteristics are
more useful as far as radioisotope production is
concerned, i.e. of higher excess reactivity, the
cadmium ratio in neutron irradiation channels
being rather high in the thermal neutron trap
and rather low in the fast neutron channels.
The establishment of a laboratory for routine
production of radioisotopes was carefully
considered by balancing the investment
requirement and the production technology of
choice, as well as the radioactive waste
treatment problem and radiation protection.
DUONG VAN DONG et al.
II. PRODUCTION OF RADIOISOTOPE
AND RADIOPHARMACEUTICALS
effective irradiation volume of trap will
increase a factor of 1.5.
A. Production laboratory and facilities
Construction of neutron irradiation positions
Due to the low power of DNRR we must
take into consideration of the irradiation
position construction, core management and
reactor operation mode in order to improve the
neutron flux, to maximize the volume available
for target irradiation and to balance the
formation-decay of activated radionuclides.
For the purpose of augmentation of
thermal neutron flux, the central irradiation
channel (so called neutron trap) was
surrounded by beryllium metal block of
thickness of 1.7 cm and height of 60 cm.
Fig. 1. Neutron plux depletion in target.
The effect of beryllium gave an
improving in flux and quality of thermal
neutron. As cited in Figs. 1 & 2, this neutron
trap has a diameter of 64 mm originally and
has only one guiding tube of diameter of 38
mm in the centre for holding the target
containers. This construction of neutron trap
has been found inconvenient in exhaustive
exploitation of irradiation volume. So it has
been proposed for reconstruction. The design
work is based on the fact of self shieldingeffect of targets and cooling water circulation,
as an example of this, neutron flux depletion in
TeO2 and MoO3 target under reactor irradiation
was noted in Fig. 1.
65cm
65cm
Old trap
New trap
Fig. 2. Neutron trap construction for the
optimization of effective irradiation
volume exploitation.
Irradiation techniques
As shown in the case of target sample of
diameter of 2 cm, the neutron flux in its center
dropped about 10 percent. This fact leads us to
design a neutron trap which is composed of
two channels of 24-mm diameter. The
sectional cut of old and new neutron trap was
shown in Fig. 2. With this new construction,
The targets held in the quartz ampoule
were irradiated with thermal neutron either in
the neutron trap at the center of the reactor core
or in the rotary specimen rack. For fast neutron
irradiation, it was carried out in a dry channel
inserted between fuel elements of the reactor
47
PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS AT …
core. Before irradiation, the targets were
purified to remove traces of impurities.
radioisotopes of higher specific radioactivity,
such as 131I and 99Mo. 99Mo with high specific
radioactivity used for 99Mo-99Tcm generator.
99
Mo was produced by neutron irradiation of
MoO3 target at the centre of neutron trap, where
thermal neutron flux is of highest value. The
distribution of neutron flux in an irradiation
position is a very important parameter for the
management of target irradiation.
Reactor operation schedule
The schedule of reactor operation mainly
depends on the kinds of radionuclide produced
and their role. The formation rate of these
kinds of radionuclides and the required
minimal specific radioactivity of radioisotopes
are indispensable factors to decide on the
option of reactor operation mode. The DNRR
was offered to produce some important
radionuclides for nuclear medicine application.
Among these radioisotopes, 131I and 99Tcm
isotopes are most highly evaluated. So the
reasonable schedule of reactor operation must
be chosen, taking into consideration of the
production yield and quality of 131I and 99mTc
radioisotope products. Basing on the formation
rate under reactor activation and half life of
99
Mo and 131I radionuclides a reactor operation
schedule of 130-150 hrs of continuous run
every three weeks has been applied.
Fig. 3. Quartz ampoule and aluminum container
for containing target.
Reactor core management for the irradiation of
targets
The core management plays an
important role in the optimization of research
reactor utilization for production of
radioisotopes. The core management is based
on the nuclear reaction applied to produce a
predescribed radionuclides, the neutron
activation cross section and/or requested
specific radioactivity of a specified radioactive
products. Besides, the neutron flux and
characteristics of irradiation position such as
Rcd, neutron flux distribution were also taken
into consideration.
At the DNRR, the irradiation channel of
lowest cadmium ratio, Rcd =1.90 is used for
fast neutron irradiation to produce the
radionuclide with (n, p) nuclear reaction, such
as 32S(n, p)32P. 32P isotope produced in this
channel is of high specific radioactivity and is
used for preparation of injectable 32P solution.
Meanwhile the rotary specimen rack of highest
cadmium ratio, Rcd = 4.5 is used for production
of 32P of low specific radioactivity with 31P(n,
γ)32P reaction. This 32P product was used to
prepare the 32P applicators for skin disease
treatment. In the neutron trap of highest
thermal neutron flux and of Rcd = 2.93, the (n,
γ) nuclear reaction was applied to produce the
Fig. 4. Annual operation time at DNRR
since 1984 to 2013.
48
B. Production laboratory
in 1990 with 2 shielded cells ball-joint
manipulators (Fig. 7).
The main utilization of the DNRR is the
production of radioisotopes for nuclear
medicine, agriculture, sedimentology and other
scientific research. About 90 percent of time is
used for radioisotope production.
An area of 200 sq.m is reserved for a
rather limited programme of isotope
production. The facilities available for the
isotope production consist of one hot cell with
master slave manipulator (Fig. 5).
Fig. 6. 131I isotope production line equipped
in 2008 with 2 shielded cells.
Fig. 5. Hot cell with master slave manipulator
One 131I isotope production line
equipped by the IAEA TC Project VIE/0/002
in 1987 with 4 shielded cells, one 131I isotope
production line equipped in 2008 by the
National Project with 2 shielded cells ball-joint
manipulators (Fig. 6), and five shielded fume
Fig. 7. 99mTc generator production line.
All these facilities are connected with
the existing ventilation system of the reactor.
hoods for isotope labelling and -emitted
isotope processing.
Equipment for the production of Kits to
be labeled with 99mTc isotope and for the
quality control of radioisotopes and
radiopharmaceuticals was also supplied by the
National Projects (Figs. 8 and 9).
One 99mTc generator production line
(using fission 99Mo solution) equipped
under the IAEA TC Project No. VIE/6/016
49
- Other radioisotopes such as 60Co, 65Zn,
64
Cu, 24Na, 86Rb, 46Sc, 71Ge, 55Fe, etc., were also
produced in a small amount when requested.
C. Radiochemical processing of activated
targets
Iodine-131:
Iodine-131 is produced from the
irradiated tellurium dioxide in neutron trap.
The target of tellurium dioxide contained in a
welded aluminum capsule, according to the
nuclear reaction as follows:
Fig. 8. Sterile hot cell.
The irradiated tellurium dioxide powder
is transferred to a Vycor distillation vessel and
connected to the iodine-131 tellurium
processing system. The processing furnace is
heated up to 750oC in order to distill the
iodine-131 over to a charcoal column trap
connected in-line of the distillation system.
The charcoal column trap is rinsed with the deionized water then eluted with sodium
hydroxide 0.05N to form the final product of
iodine-131 solution. The scheme in Fig. 10
shows the flow chart of the operation and
procedures.
Fig. 9. Clean room.
Since the beginning of 1984 (the year of
reactor inauguration) up to now the
radioisotope production at the DNRR has
concentrated on the following radionuclides:
The target used in the production is an
analytical grade material of natural tellurium
as tellurium dioxide obtained from Fluka Inc.
The chemical purity of the target as TeO 2 is
>95%. The specification of the target before
being fired in a muffle furnace through
analysis by emission spectrograph should
contain of selenium less than 0.05% and
heavy metals less than 0.1%. After being fired
in the muffle furnace the analysis should give
selenium less than 0.005% and heavy metal
less than 0.1%.
- 32P in injectable orthophosphate
solution and 32P applicator for skin disease
therapeutics.
- 131 I in NaI solution.
- 99Mo-99Tcm generator.
- 51Cr in injectable sodium chromate
solution and Cr-EDTA.
50
Fig. 10. The flow chart of the operation and procedures of I-131.
99m
Final product specification for use
Tc generator:
The final product as sodium iodide, 131I
solution in NaOH, without reducing agents will
be used as 131I bulk solution for
radiopharmaceuticals
production.
The
specification of the final product is as follows:
Among the two reactions of choice for
production of 99Mo parent isotope, the large
investment for use of 235U(n, fission)99Mo
reaction let us to opt for the 98Mo(n, ) 99Mo
reaction to produce 99mTc generator.
Physical appearance: Colorless solution.
In order to separate 99mTc from its parent
99
Mo we first used the MEK extraction method.
Radioactivity of 131I: more than 11.1
GBq (300 mCi) I-131/mL.
133
I content: less than 0.80% of the
content at assay time.
The inherent disadvantages of this
method compelled us to start our studies on the
preparation of gel type generators in late 1984
in the framework of the IAEA-CRP on the
“Development of 99mTc generators using low
power research reactor”. This represents the
state-of-the-art for generator technology and
promises opportunities for both developed and
131
I
pH: more than 11
Radionuclidic purity:
than 99.9%.
131
I content more
Radiochemical purity: Iodide more than 95%.
51
99
developing countries particularly with respect
to eliminating the need for fission 99Mo. Two
directions of preparation of gel type generators
were studied:
Mo could be used to produce portable,
chromatographic type 99mTc- generators which
have a good performance for application in
clinical investigations. Among the established
procedures the column loading procedure was
highly evaluated, because it proved to be
prominent figures for easy and safe operation,
for low cost of technology facilities, equipment
and for the capability to match the traditional
technology of the fission 99Mo based 99mTcgenerator production.
- Preparation
of
chromatographic
generators using zirconium molybdate (ZrMo) or
titanium molybdate (TiMo) column packing
materials synthesized from the neutron irradiated
molybdenum trioxide and the zirconium chloride
and/or titanium chloride, respectively.
- Preparation
of
chromatographic
generators using TiMo column packing material
(preformed TiMo) synthesized from the inactive
molybdenum compound and TiCl4 and
subsequently neutron activated in the reactor.
DNRI had been proposed attending in
these studies program. The commercial
production of PZC generator through the
establishment of national project stage 20062008 for the routine production of 99Mo/99mTc
generator. In this project the 99mTc-generator
used PZC coming locally synthesis and from
KAKEN - Japan as the column material, 99Mo
formed from MoO3 irradiated at DNRR, the
semi-automatic loading and adsorption machine
had studied, designed and installed in the hot
cells available. The generator assembly had also
been designed and fabricated, (Figs. 11, and 12).
In both modes of preparation we have
carried out studies on three different options of
generators:
- The chromatographic generator using
0.9% NaCl solution as eluant.
- The chromatographic generator using
organic solvent as eluant Solid-Solventextraction).
- The chromatographic generator using dilute
saline as eluant and 99mTc concentration column.
In the other hand, under the framework
of Forum for Nuclear Cooperation in Asia
(FNCA) program, the PZC based technology
for production of 99mTc- generator has been
studied at DNRI as well as FNCA member
countries in the past several years.
PZC adsorbent of high performance for
Mo adsorption was easy to synthesize from
isopropyl alcohol (iPrOH) and ZrCl4.
99
The procedures and relevant 99mTcgenerator designs for the preparation of PZC
based 99mTc- generators were successfully set
up. The columns of from 1.0 gram to 4.0 gram
weight of PZC and from 100 mCi to 500 mCi
Fig. 11. Schema of 99mTc – Generator
Design of commercial PZC-99mTc generator
52
sulfur. Our glass apparatus for this
production process is shown in Fig. 13. It can
be used for distillation either in the vacuum
or in the N2 gas flow by changing the upper
stopper of the distillation vessel. The
distillation parameters and post-distillation
purification of 32P solution were adopted as
described in literature.
Fig. 12. The semi-automatic loading machine
In conclusion, it is strongly believed that
ZrMo, TiMo and PZC based generator play an
importance role as alternative technology for
production of 99Mo/99mTc generator from
reaction 98Mo(n, γ)99Mo. However these
methods were not very appropriate for the low
power research reactor as DNRR. Because of
those reasons, it is necessary to build a new
research reactor with power at least of 10 MW,
and the neutron flux is high enough to research
and produce radioisotopes.
Fig. 13. The glass apparatus for 32P production
process using nuclear reaction 32S(n, p)32P
The 32P applicators for skin disease
treatment were produced by neutron irradiation
of a soft plate preformed from cloth binder and
a covering mixture of red phosphorus and glue.
After irradiation in the reactor, the radioactive
plate was impregnated with plastic and covered
with Scotch adhesive.
The mechanical
strength of the preformed plate was not lost
under 75-hour irradiation in a thermal neutron
flux of 5x1012 n.cm-2.s-1. Under this irradiation
a plate containing 65 mg P per square
centimeter gives a radioactivity of 15 mCi 32P.
The absorbed dose rate on the surface of the
plate of size 50 x 40 mm2 was measured as 110
Rad.min-1 at the center and 75 Rad.min-1 on the
edge. Medical doctors’ experience over ten
years showed that with repeated treatment of
three or five 15-minute applications the
following diseases will be cured: Eczema, skin
cancer, bump scar, etc. At present more than
75 Ci 32P in applicator form are used annually
in the country.
Phosphorus-32:
32
P isotope was produced according to two
nuclear reactions: 32S(n, p)32P and 31P(n, )32P.
The first reaction was used for the
production of injectable carrier-free 32P
solution, the second for that of 32P –isotope
applicators for skin disease treatment.
First the injectable 32P solution of
radioactivity of ten mCi scale was produced
from irradiated MgSO 4 target using magnesia
as absorbent to separate 32P isotope from
MgSO4 solution. In the case of Ci scale
production, the large amount of waste
produced from this technology caused
storage problems.
Recently, we have
introduced the distillation technique to
separate 32P from reactor irradiated elemental
53
Cr-51 isotope:
chromatography techniques for chemical
purity, and the spectrometry and neutron
activation analysis for chemical purity.
Biodistribution
assay,
biological
tests
(apyrogenity,
sterility,
toxicity)
and
physicochemical tests (pH, turbidity) are also
carried out regularly.
The production of 51Cr isotope was
carried out based on the Szilard-Charmel
reaction using reagent grade K2CrO4 target.
The chemical separation of recoiled 51Cr
nuclide was based on the selective adsorption
of this isotope on an inorganic ion exchanger
Si-ZrP (Silica gel supported zirconium phosphate) synthesized by us.
Other isotopes
were also produced
when requested in small amounts for industrial
and agricultural applications. The methods for
production of these isotopes were selected
from investigation results or different reference
sources.
D. Production of Kits for labelling with 99mTc:
In furthering the application of 99mTc
isotope, the local availability of Kits for
labelling with 99mTc plays an important role.
With IAEA support the basic equipment for the
production of Kits has been installed in our
laboratory. At present many kinds of in-vivo
Kits have been successfully prepared and put
to use in the country, they are Phytate,
Gluconate, Pyrophosphate, Citrate, DMSA,
HIDA, DTPA, Maccroaggregated HSA and
EHDP (1-hydroxy ethylidene-1, 1-disodiumphosphate).
Fig. 14. HPLC system for QC.
III. LOCAL PRODUCTION VOLUME
AND DEMAND
These types of radioisotopes have
regularly been supplied to more than 25
hospitals in Vietnam two times per month. The
131
1 radioisotope labelled radiopharmaceuticals
such as 131I-Hippuran; 131IMIBG have also
been regularly supplied to hospitals.
Radioisotope production rate is shown in Fig.
15 and Table I.
The studies on the preparation of
Radioimmuno-assay Kits and therapeutic
agents and/or radionuclides were also carried
out. The future production of the above
mentioned items is foreseen and planned.
In order to support the application of
Tc, 113mIn, 177Lu and 153Sm radioisotopes in
clinical diagnosis and therapeutics, the
preparation of radiopharmaceuticals in Kit
forms has been carried out. The following Kits
have regularly been manufactured in DNRI:
Phytate, Gluconate, Pyrophosphate, Citrate,
DMSA,
EHIDA,
DTPA,
HSA
macroaggregated, HEDP, HmPAO, MIBI,
MDP.
99m
E. Quality control
Radioisotope and radiopharmaceutical
quality control was carried out for all batches
of our products. The gamma spectrum analysis
using Ge-Li detector coupled with a
multichannel analyser is used for radionuclide
purity control, the TLC, HPLC and gel54
Radioimmunoassay Kits: The RIA Kit
production and distribution programme have
also started. T3 and T4 Kits have been selected
locally by end-users with a share of 50% of
domestic market. Other RIA and IRMA Kits
can be supplied to end-users by dispensing
process based on the contract.
Fig. 15. Total activity of radioisotopes produced
at DNRI
Fig. 16. Radioisotopes and Radiopharmaceuticals
produced at the DNRI
Table I. The supply/demand for radioisotopes and diagnostic Kits in Vietnam.
Product
Supply
131
IDiagnostic
and
therapeutic capsule/solution
99m
Tc-Generator
Demand at present
20-30 Ci/month
40-50 Ci/month
10 generators (200500mCi/each)/month
40 generators (200500mCi/each)/month
50 Ci/month
50 Ci/month
32
-
P-Solution/
Applicator
Kits for 99mTc-Labelling
MDP
DTPA
DMSA
PHYTATE
Orther
400 Kit/ month
200 Kit/ month
200 Kit/ month
200 Kit/ month
200 Kit/ month
IV. THE APPLICATION OF LOCAL
PRODUCTS IN THE COUNTRY
- Number
of
nuclear
departments in Vietnam: 25
500 Kit/ month
300 Kit/ month
300 Kit/ month
300 Kit/ month
300 Kit/ month
- There are six centres of PET-CT and
cyclotron in Hanoi Capital and Ho Chi Minh
City.
medicine
- Radiopharmaceuticals used in these
centres: Na131I solution and capsule, Sodium(99mTc) pertechnetate (99mTc-Generator) 131IHippuran, Sodium-(32P) orthophosphate, 131IMIBG, In-vivo Kits (MDP, DTPA, DMSA,
These departments almost are located in
the big cities of the country (Fig. 17).
- Number of gamma cameras (planar
and SPECT): 22
55
ordination Meeting,
(October 1987).
Phosphon, Glucon, Phytate, MAHSA, EHIDA,
HMPAO, MIBI, MAG-3, etc.).
Bandung,
Indonesia,
[2] Radioisotope production and quality control.
Technical Reports Series No. 128. IAFA,
Vienna, (1971).
- Locally manufactured products take
about 50% of total market. In order to get a
higher market share now we increase the
production by loading generations with
importing raw materials such as 99Mo and 131I
solutions.
[3] Le Van So, Investigation on the silica gel
supported form of micro crystalline zirconiumphosphate ion exchanger and its applications in
chemical separation.
I.- Preparation, ion exchange properties and
stability of Si-ZrP, J. Radioanal. Nucl. Chem.,
(Articles) 9 (1) 17-30 (1986).
[4]
Le
Van
So,
Richard
M.
Lambrecht,
Development of alternative technologies for a
gel-type chromatographic 99mTc generator. J.
Labelled Compd. Radiopharm. 35:270 (1994).
[5] Ngo Quang Huy et al, Reactor physics
experimental studies on Dalat nuclear research
reactor, 50A-01-04 Research Project Final
Report, (1990) (in Vietnamese).
[6] Tran Ha Anh et al, Studies on Dalat Nuclear
Reactor Physics and Technique and on
Measures to ensure the safety and efficiency of
the reactor, KC-09-15 Research Project Final
Report, (1995).
[7] Nguyen Nhi Dien, Dalat nuclear research
reactor - status of operation and utilization,
Dalat Sym. -RR-PI-05, Dalat, (2005).
Fig. 17. Location of Nuclear Medicine Departments
in Vietnam.
[8] Duong Van Dong, Status of Radioisotope
Production and Application in Vietnam, Dalat
Sym. -RR-PI-09, Dalat, (2009).
REFERENCES
[9] Duong Van Dong, Status of the study on PZC
[1] Le Van So, Production of 99mTc isotope from the
based Tc-99m generator and potential of its
commercial production in Vietnam, Nihon
chromatographic generator using zirconiummolydate and titanium-molybdate targets as
column packing materials. Research Co-
Genshiryoku Kenkyu Kaihatsu Kiko JAEAConf, Journal Code: L2150A, page 25-29
(2007).
56
The gamma two-step cascade method
at Dalat Nuclear Research Reactor
Vuong Huu Tan1, Pham Dinh Khang2, Nguyen Nhi Dien3, Nguyen Xuan Hai3,
Tran Tuan Anh3*, Ho Huu Thang3, Pham Ngoc Son3, Mangengo Lumengano4
1)
Vietnam Agency for Radiation and Nuclear Safety, 113 Tran Duy Hung, Hanoi, Vietnam
2)
Vietnam Atomic Energy Institute, 59 Ly Thuong Kiet, Hanoi, Vietnam
3)
Nuclear Research Institute, 01 Nguyen Tu Luc, Dalat, Vietnam
4)
Agostinho Neto University, Av, 4 Fevereiro, 71 Ingombotas, Luanda, Angola
*
Email: ndsdalat@vnn.vn
Abstract: The event-event coincidence spectroscopy system was successfully established and operated
on thermal neutron beam of channel N0. 3 at Dalat Nuclear Research Reactor (DNRR) with resolving
time value of about 10 ns. The studies on level density, gamma strength function and decay scheme of
intermediate-mass and heavy nuclei have been performed on this system. The achieved results are
opening a new research of nuclear structure based on (n, 2) reaction.
Keywords: event-event coincidence, thermal neutron beam, nuclear structure.
I. INTRODUCTION
The nuclear parameters obtained from
intensities of two-step cascades have
considerably higher reliability than those
obtained within known methods due to
unsuccessful relation between the experimental
spectra and desired parameters of the gammadecay process. For excited levels below 2 MeV,
their spectroscopic information in detail were
known very well from investigations of (n, ),
(n, e), (d, p)... reactions. However, for higher
excited levels, the information is not enough
because of low intensity of transitions and bad
resolution of detectors [1].
The traditional gamma spectrometer
allows getting more information about nuclear
data and nuclear structure from their spectra. The
background, however, is high due to Compton
scattering. In order to reduce the background, it
is necessary to develop advanced spectrometers
such as Compton suppression, pair production,
or coincidence systems.
In this work, the gamma two-step
cascade (TSC) method has been developed to
optimize solution and to reduce Compton
scatter and pair-production phenomena in the
gamma spectra of nuclei decay gamma
cascades. This is allowed to determine
precisely gamma cascade intensities and to find
intermediate levels in an energy region near a
binding energy. Since, the transition
probabilities and quantum characteristics of
intermediate levels are split. The characteristics
allow comparing transition probabilities
between theory and empirical results [2].
II. TSC METHOD
The method is based on event-event
coincidence measurements of two γ-rays from
the cascade decay of a compound nucleus
following thermal neutron capture. The total
energies of the γ-rays and their time
differences are measured by two germanium
detectors. Coincidence events are selected
which have a sum energy given by the energy
THE GAMMA TWO STEP CASCADE METHOD AT DALAT RESEARCH REACTOR
difference between the capture state and the
pre-selected low-lying state. The detected
spectrum then contains information on two
types of transitions. The 1st type includes first
transitions populated in the intermediate
region of excited energy. Because of large
number of levels in this region, no
spectrometer is available for data acquisition.
The 2nd one includes transitions that the
intermediate levels dominate low energy
levels [2, 3, 4]. In this case, the event-event
coincidence spectroscopy can be used in
advance for level densities determination.
spectrometer. The detectors were shielded by
lead blocks of 10 cm in thickness. The distance
between the source and the detectors’ surfaces
is 4 cm. In order to decrease the back scattered
gamma rays and filter out X-ray, two lead
plates of 2 mm in thickness were placed in
front of the detectors and sample. The
background count rate was less than 600 counts
per second (cps) in 0.2 ÷ 8 MeV range [5].
Data acquisition system
The electronics configuration used in
those - coincidence experiments is shown in
Figure 1.
III. TSC GAMMA MEASUREMENT
Neutron
arrangement
beam
and
The detector signals are amplified with
572 amplifier (AMP) modules with a shaping
time of 3.0 µs and about 1 keV per channel.
The output signals of the amplifiers are
digitized by 7072 analog-to-digital converter
(ADC) modules. The timing signals of both
detectors are put through 474 timing filter
amplifier (TFA) modules.
detectors
The experiment system has been
installed at the tangential beam port of the
DNRR. The thermal neutron beam was
moderated by Si filter. The neutron flux, the
cadmium ratio and the neutron beam diameter
at the sample position were 2.4105 n.cm-2.s-1,
230 and 1.5 cm respectively.
The shaped and amplified timing
signals by 474 TFA are plugged into 584 CFD
modules, which are used in slow rise time
rejection option (SRT) mode. The CFD output
signal of the first channel is used as 556 timeto-amplitude converter (TAC) start signal.
Two horizontal GMX35 detectors
manufactured by ORTEC with the energy
resolutions of 1.9 keV at 1332 keV (60Co)
have been used in the - coincidence
Fig. 1. The - coincidence electronics.
58
NGUYEN XUAN HAI et al.
appearing in the corresponding coincidence
data file. The coincidence spectrum of one
detector with the chosen peak in another
detector can be created by the same procedure.
They are coincidence spectra between highenergy primary and low-energy secondary
transitions or among the low-energy secondary
transitions as obtained in the work [3, 4].
Besides, the summation spectrum of amplitudes
of coincidence pulses can be created by
summation of pairs of coincidence data. Every
full-peak in the summation spectrum is
corresponding to the - cascade decays from
the capture state to the determined low-lying
excited level. The TSC spectrum of one
detector associated with the defined energy (E)
summation peak will be taken by choosing
pairs of coincidence data having summation in
the range of E ± E (with E/E ≤ 0.005) (see
Figure 3). The TSC spectrum gives information
on levels in the region between the capture state
and the defined E low-lying level. From all
obtained TSC spectra we can build up the decay
scheme of the investigated nucleus on the base
of methods and the criteria given in Ref. [5].
The measured values of gamma two-step
cascade energies and intensities of 35Cl(nth,
2γ)36Cl reaction were shown in Table 1.
1000
100
7413.95 keV
200
7790.32 keV
10ns
2000
6627.75 keV
300
6977.85 keV
3000
E1+E2 = 8579 keV
788.43 keV
400
Counts
Counts
4000
1164.87 keV
500
5000
5517.2 keV
5715.19 keV
5902.7 keV
In the experiment, the data, which
contains all pairs of - coincidence data from
two HPGe-detectors, were stored in the
memory of computer. Indeed, that is pairs of
channel numbers associated with energies of
- coincidence pairs. The coincidence
spectrum of each detector can be created from
the corresponding data file by the procedure
that the count number of each channel of the
spectrum is equal to times of that channel
2676.30 keV
2863.82 keV
3061.86 keV
Coincidence Data Processing
1959.36 keV
The full scale of TAC is set at 100 ns,
and output signal is digitized in 8713 ADC
with selection of 1024 channels for a 10 V
input pulse. The TAC “Valid Convert” signal
is used to gate 7072 ADCs, and the delay or
synchronizing with AMP output signal is
implemented by interface software. Recorded
coincident events have three values, including
coincidence gamma-ray energies from detector
1, detector 2 and time interval between two γrays in a pair event [5]. The resolving time for
this configuration is about 10 ns with 60Co
source measurement (see Figure 2).
1601.08 keV
The CFD output signal of the second
channel is delayed 100 ns and served as a TAC
stop signal.
0
0
0
10
20
30
40
2000
Resolving time (ns)
4000
6000
8000
Energy keV
Fig. 3. The TSC spectrum of 36Cl belongs to final
level from 8579 keV.
Fig. 2. The resolving timing spectrum
59
THE GAMMA TWO STEP CASCADE METHOD AT DALAT RESEARCH REACTOR
Table 1. The gamma two-step cascade energies and intensities of 35Cl(nth, 2γ)36Cl reaction.
Eγ
(keV)
787.03
1164.01
1370.00
1958.98
1164.01
3723.00
517.05
1950.98
789.03
1164.60
1601.49
1958.48
2864.28
7413.06
6979.37
3062.98
5518.16
6621.31
5716.18
788.23
1950.17
6629.20
7792.32
Measured values
Up level
Low level
(keV)
(keV)
1952.98
1164.01
1952.98
787.03
3331.99
1958.98
1958.98
0.00
1164.01
0.00
4886.09
1164.01
517.05
0.00
2465.97
517.05
789.03
0.00
1164.60
0.00
1601.49
0.00
1958.48
0.00
2864.28
0.00
8579.71
1165.01
8579.71
1602.99
8579.71
5518.16
5518.16
0.00
8579.71
1957.98
8579.71
2863.98
788.23
0.00
1950.17
0.00
8579.71
1950.17
8579.71
788.23
IV. RESULTS
Within the framework of this research
project, the obtained results are as follows:
- Setting up successfully the eventevent coincidence spectrometer with for
measuring nuclear structure data on thermal
neutron beam.
- Measuring and analyzing the
gamma cascade transition data for nuclei of
239
U, 182Ta, 153Sm, 172Yb, 59Ni, 55Fe and 49Ti.
The experimental data are to evaluate
excited states in the intermediate energy
below the neutron binding energy.
- Evaluating nuclear structure for
those nuclei based on analyzed data and
theoretical models.
Eγ
(keV)
786.30
1162.78
1372.86
1959.36
1164.87
3723.00
517.08
1951.14
788.43
1164.87
1601.08
1959.36
2863.82
7413.95
6977.85
3061.86
5517.2
6619.64
5715.19
788.43
1951.14
6627.75
7790.32
XCI 6/18/013
Up level
Low level
(keV)
(keV)
1951.20
1164.89
1951.20
788.44
3332.32
1959.41
1959.41
0.00
1164.89
0.00
N/A
2468.28
1951.20
1951.20
0.00
788.44
0.00
1164.89
0.00
1601.12
0.00
1959.41
0.00
2863.96
0.00
8579.70
1164.89
8579.70
1601.12
8579.70
5517.76
5517.76
0.00
8579.70
1959.41
8579.70
2863.96
788.44
0.00
1951.20
0.00
8579.70
1951.20
8579.70
788.44
I-
10.520
2.290
0.384
12.560
27.20
24.300
19.390
16.320
27.20
3.484
12.560
5.770
10.520
2.290
3.521
1.689
7.830
5.310
16.32
19.39
4.690
8.310
- Determining the lifetime level, width
level and gamma transition strength from the
experimental data of gamma intensity and
electromagnetic transfer selection.
- Providing methods and experimental
facilities for basic researches, education and
training.
V. CONCLUSION
The γ-γ coincidence spectrometer is a
useful tool in research on nuclear spectroscopy in
DNRR. Besides, the spectrometer can also be
used in research on the lifetime of some excited
states and γ-γ angular correlations that are
completely new research fields. For some
elements in the deformed nuclei region with high
possibility
of
cascade
transitions,
this
60
NGUYEN XUAN HAI et al.
spectrometer can be used for the neutron
activation analysis because of very low
gamma backgrounds.
The research method and facilities
for TSC measurements will play a
significant role in carrying out R&D
programs of nuclear technique applications
so far, as well as in preparing human
resources for the nuclear data program in
Vietnam in the near future.
REFERENCES
[1] A. A. Vankov et al. In Proc. Conf. on Nuclear
Data for Reactors. Helsinki 1970, IAEA, Vienna,
Vol.1, p.559 (1970).
[2] H.H. Bolotin. Thermal-neutron capture gamma-
gamma coincidence studies and techniques,
Proceedings of the 1981 International Symposium
on Neutron Capture Gamma Ray Spectroscopy
and Related Topics, Grenoble, France, p.15-34
(1981).
[3] S.T. Boneva et al. Two-step cascades of neutron
radiative capture: 1. The spectroscopy of excited
states of complex nuclei in the range of the
neutron binding energy, Physics of Elementary
ACKNOWLEDGMENTS
The authors would like to express
their sincere thanks to the researchers of
DNRR for their cooperation concerning to
neutron irradiations. This research is funded
by Ministry of Science and Technology,
Vietnam Atomic Energy Institute and
Nuclear Research Institute.
Particles and Atomic Nuclei, Vol.22, Part.2,
p.479-511 (1991).
[4] S.T. Boneva et al. Two-step cascades of neutron
radiative capture: 2. Main parameters and
peculiarities complex nuclei compound-states decay, Physics of Elementary Particles and
Atomic Nuclei, Vol.22, Part.6, p.1431-1475
(1991).
[5] Vuong Huu Tan et al. Investigation of gamma
cascade transition of 153Sm, 182Ta, 59Ni and 239U
using the gamma two step cascade method, Final
report of the research project, Ministry of
Sciences and Technology, Code BO/05/01/05,
(2005-2006).
61
Progress of Filtered Neutron Beams Development and Applications
at the Horizontal Channels No.2 and No.4 of Dalat Nuclear
Research Reactor
Vuong Huu Tan1, Pham Ngoc Son2*, Nguyen Nhi Dien2, Tran Tuan Anh2, Nguyen Xuan Hai2
1
Vietnam Agency for Radiation and Nuclear Safety, 113-Tran Duy Hung, Hanoi
2
Nuclear Research Institute, 01 Nguyen Tu Luc, Dalat
*
E-mail: pnson.nri@gmail.com
Abstract: The neutron filter technique has been applied to create mono-energetic neutron beams with
high intensity, at the horizontal channels No.2 and No.4 of the Dalat nuclear research reactor. The
mono-energetic neutron beams that have been developed for researches and applications are thermal
(0.025eV), 24keV, 54keV, 59keV, 133keV and 148keV. The relative intensities of main peak in
filtered neutron energy spectra and the collimated neutron fluxes at the sample irradiation positions
are 90  96% and 2.8×105  7.8×106 n/cm2.s, respectively. Monte Carlo simulations and transmission
calculations were performed to each neutron energy beam for optimal design of geometrical structure
and neutron filter materials. These filtered neutron beams have been applied efficiently for
experimental researches on neutron total and capture cross sections measurements, and elemental
analysis in various kinds of samples based on the prompt gamma neutron activation analysis method.
This paper reviews the progress of filtered neutron beams development and its applications for past
many years at the Dalat nuclear research reactor.
Keywords: Filtered neutron beam, nuclear data measurement, Dalat nuclear research reactor
I. INTRODUCTION
The Dalat nuclear research reactor
(DNRR), located in campus of the Nuclear
Research Institute, VINATOM, was originally
a TRIGA MARK II reactor with a nominal
power of 250kW completed construction and
reached critical state in 1963. The reactor then
has been upgraded to nominal power of 500
kW since 1984. There are three radial and one
tangential beam ports at DNRR, each of which
penetrates the concrete shield structure and the
reactor water to provide external beams of
neutron originated from reactor core [1]. The
cross section view of horizontal channels of
DNRR is shown in Fig.1. The radial beam port
No.4 has been used to develop mono-energetic
neutron beams of thermal, 54keV and 148keV
(previous reported as 55 and 144keV) by the
neutron filter technique for basic research on
neutron induces nuclear reaction data
measurements since 1991 [2]. For efficient and
extensive uses of the neutron channel, the
neutron filter technique has been also applied
to create high intensity neutron beams with
quasi-monoenergies of 24keV, 59keV and
133keV at the channel No.4 in 2008 [3].
Fig. 1. Structure of horizontal neutron channels of
the Dalat nuclear research reactor
VUONG HUU TAN, PHAM NGOC SON, NGUYEN NHI DIEN, TRAN TUAN ANH, NGUYEN XUAN HAI
In order to enhance the utilizations of
DNRR for neutron capture experiments and
prompt gamma-rays neutron activation
analysis (PGNAA) applications, the beam port
No.2 of the reactor have been opened in
advance since 2010 [4] for development of a
modern
prompt
gamma-ray
Compton
suppression facility used a high efficient
HPGe-BGO detectors system. In the works of
neutron filters development, the Monte-Carlo
simulation used MCNP5 code, and
transmission calculation by CFNB code [5]
have been performed for each neutron beam
for optimal design of geometrical structure,
neutron filter materials and radiation
shielding.
II. FILTERED NEUTRON BEAMS
The optimal design structure for
insertion of filters into the horizontal channel,
beam collimators and radiation shielding
chamber at the Dalat nuclear research reactor is
shown in Fig.2 [4]. The neutron energy spectra
after filtered through a suitable composition
materials used as filters can be calculated as the
following expression:
 ( E )  0 ( E )*exp( k dk t ,k ( E )) , (1)
k
where
0 ( E ) ,
 ( E)
are
energy
distributions of the neutron spectra before and
after transmitted through the filters; k, dk and
t,k(E) are the mass density, length of filter and
total cross section of kth filter material,
respectively. The filter information and
physical parameter of each energy beam is
presented in the following sub-sections.
The applications of these filtered
neutron beams were mainly focus on nuclear
data measurements and PGNAA elemental
analysis, although these beam lines have
possibility for many other researches and
applications such as nuclear level density and
isomer ratio determination, Boron neutron
capture therapy (BNCT) research, neutron
dosimeter calibrations,... On the nuclear data
measurements respects, the channels provide
essential neutron beams for precise
experimental reaction data of neutron total and
capture reaction cross sections. On the
PGNAA application subject, the assessment of
analytical sensitivity for elements of B, H, Hg,
Si, Ca, C, S, Al, Fe, Cl, Ti,... has been carried
out, and shown that the new PGNAA
spectroscopy installed at the channel No.2 is a
good facility supplemented to the neutron
activation analysis (NAA) method at the Dalat
nuclear research reactor. The detail
characteristics of filtered neutron beams
development and results of its applications are
presented in the next sections.
Fig. 2. The design structure of filtered neutron beam
facility at the channel No. 2 of DNRR
The thermal neutron beams:
The thermal neutron beam at the channel
No.4 was developed in 1991 [2]. The material
compositions of filters are 98cm Si, 1cm Ti and
35g/cm2 S. The measured thermal neutron flux
is 1.7106 n/cm2.s, and Cadmium ratio Rcd(Au)
= 112 [3]. In order to enhance the utilizations
of the Dalat research reactor for researches and
applications based on the neutron capture
63
PROGRESS OF FILTERED NEUTRON BEAMS DEVELOPMENT AND APPLICATIONS AT …
reactions, the well thermal neutron beam at the
channel No. 2 has been developed and serviced
since 2011 [4]. The neutron filters for this
0.0253eV neutron beam line are single crystals
of 80cm Si and 6cm Bi. The measured thermal
neutron flux at outer position of the beam line
nuclear research reactor from 1990s [2, 3].
Firstly, the two neutron beams with monoenergies of 54keV and 184keV were created at
the channel No.4, and provided a good
experimental station for basis researches on
reactions of neutron with material in keV
energy region. The filter information and
physical parameters of these neutron beam
lines are introduced in Table I [3], and the
corresponding neutron spectra are shown in
Figs. 3-4.
is 1.6106 n/cm2.s, and the value of Cadmium
ratio Rcd(Au) is 420.
The neutron beams of 54keV and 148keV:
The neutron filter technique has been
applied at the horizontal channels of Dalat
Table I. Physical parameters of the 54keV and 148keV neutron beams at the channel No.4
Parameters
54keV
148keV
Neutron flux (n/cm2.s)
6.7x105
3.9x106
1.5
14.8
Peak relative intensity (%)
78.05
95.78
Beam collimated diameter
3 cm
3 cm
B 0.2g/cm2
B 0.2g/cm2
Si 98cm
Si 98cm
S 35g/cm2
Ti 1cm
Energy resolution (keV)
Filter compositions
200
Transport calculation
12000
140
4000
3000
10000
Intensity (a.u)
120
100
80
8000
2000
6000
148keV
Counts
Relative intensity
160
Exp. data
Fitted line
Intensity
14000
54keV
180
Unf olding spectrum
60
4000
40
20
0
8.0E+04
1000
2000
1.2E+05
1.6E+05
2.0E+05
Neutron energy (eV)
0
0
200
400
600
800
1000
0
1200
Channel
Fig. 3. Energy spectrum of the 148keV neutron beam at
the channel No.4 of DNRR
64
Fig. 4. Measured energy spectrum of the 54keV neutron
beam at the channel No. 4 of DNRR
The neutron beams of 24keV, 59keV and 133keV:
[3] based on the neutron source from the
channel No. 4 of DRR. The characteristics of
these neutron beam lines are introduced in
the references [6, 7], and summarized in
Table II. The calculated and measured energy
spectra of these neutron beam lines are
shown in Figs. 5-7.
As progressive necessary of reactor
based mono-energetic neutron beam lines for
experimental
researches
on
neutron
interaction with mater, the new three filtered
neutron beam of 24keV, 59keV and 133keV
have been developed and applied from 2008
Fig. 5. Measured neutron spectrum for 24keV beam by
proton recoil proportional counter
Fig. 6. Calculated neutron spectrum for the 59keV
filtered neutron beam
Table II. Characteristics of the 24keV, 59keV and 133keV neutron beams at the channel No.4
Parameters
24keV
2
Neutron flux (n/cm .s)
6.1x10
Energy resolution (keV)
Peak relative intensity (%)
Beam collimated diameter (cm)
Composition of Filters
59keV
5
5.3x10
133keV
5
3.2x105
1.8
2.7
3.0
96.72
92.28
92.89
3
B 0.2g/cm2
Fe 20cm
Al 30cm
S 35g/cm2
3
B 0.2g/cm2
Ni 10cm
V 15cm
Al 5cm
S 35g/cm2
3
B 0.2g/cm2
Cr 50g/cm2
Ni 10cm
Si 60cm
3
1.0x10
CFNB
MCNP
2
Intensity (a.u)
8.0x10
2
6.0x10
2
4.0x10
2
2.0x10
0.0
-2
5.0x10
-1
-1
1.0x10
1.5x10
-1
2.0x10
En (MeV)
Fig. 7. Calculated neutron spectrum for the 133keV filtered neutron beam
65
193
Ir [9]; 185Re and 187Re [10, 11]. In addition,
the horizontal thermal Column of DNRR, a
well thermalized neutron channel, has been
also used for measurement of thermal neutron
capture cross section and resonance integral of
69
Ga and 71Ga [12]. A typical result of our
measurements in comparison with data from
other laboratories is shown in Fig. 7 [10].
III. NUCLEAR DATA MESUREMENTS
Neutron capture cross section measurements:
The measurements of neutron capture
cross sections for a number of nuclides have
been performed on the filtered neutron beams
with mono-energies of 24, 54, 59, 133 and 148
keV, at the Dalat nuclear research reactor. The
measured neutron capture cross sections data
were obtained relative to the standard capture
cross sections of the 197Au(n,)198Au reaction
by the activation method. An abridged
description of data analysis procedure is
presented as follows:
10
ENDF/BVII
185
YU.N.TROFIMOV
M.LINDNER
Cross section (barn)
S.J.FRIESENHAHN
The average capture cross sections,
<a> , for nuclide x at average neutron
spectrum <> can be determined relative to
that of 197Au standard by the following
relations:
x
  a 
Re(n,γ)186Re
A.K.CHAUBEY
R.P.ANAND
A.A.BERGMAN
This w ork
1
0.1
1.E+04
6.E+04
1.E+05
2.E+05
Ne utron e ne rgy (e V)
C x f ( , t ) x fcx I Au Au N Au   a  Au (2)
;
C Au f ( , t ) Au fcAuI x x N x
Fig. 7. Neutron capture cross section of 185Re [10]
Measurements of neutron total cross sections:
f ( , t ) 

(1  e
t1
)e
t2
(1  e
t3
)
,
The total neutron cross section
measurements are being carried out by the
transmission method for natural elements of U,
C, Fe and Al, at the filtered neutron energies of
24keV, 54keV, 59keV, 133keV and 148keV.
The experimental value of neutron total cross
section, t, can be exactly determined from the
following expression:
(3)
where the superscript „x‟ denotes sample
nucleus, and „Au‟ denotes the reference
nucleus 197Au. „C‟ stands for net counts of
the corresponding gamma peak. „t1‟, „t2‟ and
„t3‟ are irradiating, cooling and measuring
times, respectively. „λ‟ is decay constant of the
product nucleus; „εγ‟ is the detection efficiency
of detector; „Iγ‟ is the intensity of interesting γray, and „fc‟ is the correction factor for selfshielding multiple scattering effects that can be
exactly calculated by the Monte Carlo method.
t 
1
d
 1  1  0  ,
ln   
ln 

 T  d   
(4)
where „T‟ is transmission coefficient of
the collimated neutron beam that transmitted
through a purity sample with thickness d
(cm); „‟ denotes density of the sample
(Atom/cm3). „0‟ and „‟ are measured
neutron fluxes at before and after positions of
the irradiating sample, respectively. A
measurement of the transmission spectrum for
In recent years, we have conducted a
series of cross section measurements for
neutron capture (n, ) reactions in different
nuclides, and reported in scientific papers such
as: 109Ag, 186W, 158Gd [8]; 139La, 152Sm, 191Ir,
66
54keV neutron beam at the channel No.4 of
DNRR is shown in Fig.8.
of BGO-HPGe detectors was completely
developed and installed at the experimental
space of this beam port, from 2012. A
preliminary study on calibration and analytical
sensitivity for several domination elements has
been conducted. A prompt gamma-ray
spectrum of geometrical sample measured in
single and Compton-suppression modes is
shown in Fig.9. The results in this study allows
us to estimate that the new PGNAA facility
installed at the channel No.2 of DNRR is
qualified to participate in analytical services at
the Institute. A calibration curve for Boron
analysis is presented in Fig.10, and the results
of comparison analysis used standard soil
sample (NIST-2711a) is given in Table III.
Fig. 8. Measured neutron spectrum of 54keV
neutron beam transmitted through different
thickness of C sample.
IV. PGNAA APLICATIONS
From 1998, the 148keV filtered neutron
beam at the channel No.4 has been applied for
possibility studies the method of in-vivo
prompt gamma neutron activation analysis
(IVPGNAA) that involves the exposure of the
living human organs to a small dose of
neutrons. At that time, IVPGNAA is a new
technique for directly determination of toxic
elements accommodated in a specific living
human organ such as concentrations of Hg in
kidney and Cd in liver. The research was
carried out on a physical phantom installed at
the channel No.4 [13]. The results given from
this investigation introduced a high effective
new experiment with 148keV neutron beam
instead of thermal neutrons [13].
Fig. 9. Prompt gamma-rays spectrum measured at
the thermal neutron beam No.2 for soil sample, in
single and Compton-suppression modes.
25
Model
Line
Equation
y = A + B*x
Reduced
Chi-Sqr
0
Adj. R-Square
1
Exp data
Linear fitting
Value Standard Error
20
cps
A
-2.799
0.158620
cps
B
0.520
0.00544162
cps
15
The low background and well thermal
filtered neutron beam from the channel No.2
[4] of Dalat nuclear research reactor is an
advantage neutron source for prompt gammaray neutron activation analysis (PGNAA).
Accordingly, a modern Compton suppression
PGNAA spectroscopy used a compact system
10
5
0
10
20
30
40
50
g B
Fig. 10. The calibration curve for Boron analysis by
using the PGNAA facility at the channel No.2
67
Table III. The results of comparison analysis used the standard soil sample (NIST-2711a), by using the
PGNAA facility at the channel No.2
NIST-2711a
(Standard sample: Montana soil)
Sample
Elements
Measured values
Reference values
B (g/g)
50.5 ± 2.9
50
Gd (g/g)
7.59 ± 3.34
5
Sm (g/g)
6.96 ± 1.07
5.93 ± 0.28
Ca (%)
2.43 ± 0.59
2.42 ± 0.06
Al (%)
7.1 ± 0.3
6.72 ± 0.06
Si (%)
31.66 ± 3.93
31.4 ± 0.7
K (%)
2.39 ± 0.28
2.53 ± 0.10
Ti (%)
0.29 ± 0.06
0.32 ± 0.01
Na (%)
2.02 ± 0.48
1.20 ± 0.01
Fe (%)
3.01 ± 0.35
2.82 ± 0.04
V. CONCLUSIONS
supplementation to the neutron activation
analysis (NAA) method at the Dalat research
reactor.
The accomplishment of research
activities on the topics of filtered neutron
beams development and it‟s applications based
on the neutron sources from the horizontal
channel No.2 and No.4 of Dalat nuclear
research reactor is reviewed in this report. The
neutron filter technique has been effectively
applied to provide mono-energetic neutron
beam lines with qualified characteristics for
related applications at the Nuclear Research
Institute, VINATOM. The basis researches on
experimental neutron induce nuclear reaction
cross sections conducted by using these
neutron beams have been performed with
interesting results, and this research activity is
proposed to be continued, in order to
participate in providing of precise experimental
nuclear reaction data and educational
experiments. The new PGNAA facility
installed coupling with the well thermal
neutron beam at the channel No.2 plays as an
important application of this channel for
studies on neutron capture experiments and
elemental analysis. This will be an important
The new development of neutron beam
with possible mono-energy of 2keV, and
extension of application studies such as Boron
neutron capture therapy (BNCT) and neutron
dosimeter calibration are proposed.
ACKNOWLEDGEMENTS
This research is partly funded by
Vietnam National Foundation for Science
and
Technology
Development
(NAFOSTED) under grant number “103.042012.59”. The authors are immensely
grateful to Mr. Luong Ba Vien, Deputy
Director of the Nuclear Research Institute,
VINATOM, for his great encouragement
and critical reading of the manuscript.
68
[8] Vuong Huu Tan, Pham Ngoc Son, et al.
REFERENCES
Neutron Capture Cross Section Measurements
of 109Ag, 186W and 158Gd on Filtered Neutron
Beams of 55keV and 144keV. Nuclear Science
[1] General Atomic, Triga Mark II Reactor –
General Specifications and Description. GA2627 (1961).
and Technology, Vol.3 No.1, pp.1-7; IAEA,
Nuclear data section, INDC-VN-011, (2004).
[2] Vuong Huu Tan. Application study on the
reactions induced by neutron, gamma and
charge particles based on the available nuclear
facilities in Vietnam. State scientific project
Measurement of Neutron Capture Cross
Section of 139La, 152Sm and 191,193Ir at 55keV
and 144keV. Proc. of Symposium on Nuclear
report, code: KC-09-08A (1994).
Data, Tokai, Ibaraki, Japan, SND2006-V.02-1
(2007).
[3] Vuong Huu Tan. Study on development of
nuclear spectroscopy to be used at the neutron
beams for cascade gamma transitions and
nuclear data measurements.
Ministry
Capture Cross Section Measurements of 185,
187
Re with Filtered Neutron Beams at the Dalat
Research Reactor. Journal of the Korean
scientific project report, code: BT12-07-09NLNT, (2009) (in Vietnamese).
Physical Society, Vol.59, No.2, pp. 1757-1760
(2011).
[4] Pham Ngoc Son. Development of filtered
neutron beam based on the horizontal channel
No.2 of the Datal nuclear research reactor.
[11] Pham Ngoc Son, Vuong Huu Tan. Filtered
Ministry scientific project report, code:
ĐT.08/09/NLNT, (2012) (in Vietnamese).
Neutron
Capture
Cross
Section
of
186
W(n,γ)187W reaction at 24 keV. Proceedings
of the 4th Asian Nuclear Reaction Database
Development Workshop, al-Farabi Kazakh
National University, Almaty, Kazakhstan, 23 –
25 October 2013, IAEA Nuclear Data Section,
INDC(KAS)-001, (2014).
Development of filtered neutron beams of 24,
59, and 133 keV at Dalat research reactor.
Nuclear Science and Technology, ISSN: 18105408, No.3, pp.8-15 (2009).
[12] Pham Ngoc Son, Vuong Huu Tan, et al
[6] Pham Ngoc Son, Vuong Huu Tan, Phu Chi
Hoa, Tran Tuan Anh. Development of Filtered
Measurement of Thermal Neutron Crosssection and Resonance Integrals of the
69
Ga(n,)70Ga and 71Ga(n,)72Ga Reactions at
Dalat Research Reactor. Journal of the Korean
Neutron Beams of 24keV and 59keV at Dalat
Research Reactor. Accepted to be published in
World Journal of Nuclear Science
Technology, Vol.4 (2014).
and
Physical Society, Vol.59, No.2, pp. 1761-1764,
ISSN: 0374-4884 (2011).
[7] Tran Tuan Anh, Pham Ngoc Son, Vuong Huu
Tan, Pham Dinh Khang, Phu Chi Hoa.
[13] V. H. Tan, et al. Development of In-Vivo
Prompt Gamma Activation Analysis Using
The Filtered Neutron Beam at The Dalat
Reactor. Proceeding of 11th Pacific Basin Nucl.
Characteristics of Filtered Neutron Beam
Energy Spectra at Dalat Reactor. Accepted to
be published in World Journal of Nuclear
Science and Technology (2014).
Conf., Canada, (May 1998).
69
Characterization of neutron spectrum parameters at irradiation
channels for neutron activation analysis after full conversion of the
Dalat nuclear research reactor to low enriched uranium fuel
C.D. Vu1*, T.Q. Thien1, H.V. Doanh1, P.D. Quyet2, T.T.T. Anh3, and N.N. Dien1
1
Nuclear Research Institute, 01 Nguyen Tu Luc St., Dalat, Lamdong
2
Chu Van Anhigh school, Ductrong, Lamdong
3
The University of Dalat, 01, Phu Dong Thien Vuong St., Dalat, Lamdong
*Email: caodongvu@yahoo.com
(Received 12 March 2014, accepted 9 May 2014)
Abstract: In the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and
the program on Reduced Enrichment for Research and Test Reactor (RERTR), the full core
conversion of the Dalat Nuclear Research Reactor (DNRR) to low enriched uranium (LEU, 19.75%
235
U) fuel was performed from November 24, 2011 to January 13, 2012. The reactor is now operated
with a working core consisting of 92 WWR-M2 LEU. After the full core conversion, the neutron
spectrum parameters which are used in k0-NAA such as thermal neutron flux (th), fast neutron flux
(fast), f factor, alpha factor (), and neutron temperature (Tn) have been re-characterized at four
different irradiated channels in the core. Based on the experimental results, it can be seen that the
thermal neutron flux decreases by 6÷9% whereas fast neutron flux increases by 2÷6%. The neutron
spectrum becomes‘harder’ at most of irradiated positions. The obtained neutron spectrum parameters
from this research are used to re-establish the procedures for Neutron Activation Analysis (NAA)
according to ISO/IEC 17025:2005 standard at NuclearResearch Institute.
Keywords: Neutron Activation Analysis (NAA), k-zero method, neutron flux, HEU, LEU.
I. INTRODUCTION
Dalat nuclear research reactor was
upgraded from the TRIGA Mark-II designed
and constructed by the United States. The
project of reconstruction and upgrade of the
reactor was started in March 1982. The
criticality was reached at 19:50 on November
01, 1983 and its regular operation at nominal
power of 500 kW was started from March 1984
with the core loaded with 88 WWR-M2 fuel
assemblies enriched to 36% (HEU- Highly
Enriched Uranium) [1].
Through the full core conversion project
performed from November 24, 2011 to January
13, 2012, the DNRR now is operated with a
core configuration consisting of 92 WWR-M2
LEU fuel assemblies [1, 2]. Since March 2012,
the reactor has been continuously operated
about 100÷130 hours per month at nominal
power of 500 kW for radioisotopes production,
activation analysis and other researches.
At the DNRR, there are four irradiated
channels used for NAA (Fig. 1): (1) the fast
pneumatic transfer system for short irradiation
at the channel 13-2 and thermal column (Ti<45
sec); (2) another pneumatic transfer system for
short and medium irradiation at the 7-1 channel
(Ti: 45÷1200 sec); (3) the rotary rack with 40
irradiated holes placed inside the graphite
reflector for long irradiation (Ti>20 min).
CAO DONG VU et al.
II. EXPERIMENTAL
To standardize the irradiation channel, it
is necessary to identify three basic parameters
such as thermal neutron flux (th), the ratio of
thermal to epi-thermal neutron flux (f), and the
coefficient describing the deviation of neutron
spectrum distribution from the 1/E shape (α).
In addition, two other parameters, fast neutron
flux (fast), and neutron temperature (Tn) are
also considered to be the characteristic
parameters of the neutron spectra in the
irradiation channel [3].
Fig. 1. Dalat research reactor cross-section.
The neutron spectrum parameters used
in k0-NAA including thermal neutron flux, fast
neutron flux, and the factors of f, alpha, and
neutron temperature have been re-characterized
at four irradiated channels after full core
conversion to LEU fuel. The obtained neutron
spectrum parameters from this research are
used to re-establish the procedures for Neutron
Activation Analysis using k0-IAEA software.
In this study, bare multi-monitor method
[3, 4, and 5] using set of four monitors (Al0.1%Au, Al-0.1%Lu, 99.98%Ni and 99.8%Zr)
was applied to determine the parameters of the
neutron spectra at four irradiated positions of
the reactor. The experimental conditions are
described in Table I.
Table I. The irradiation, decay and counting times for the monitors with Au, Lu, Ni, and Zr.
(Ti)/position + Monitors
Irradiation time
Decay
time
(weight)
(Td)
- 15 m/channel 7-1 and 132
- 3 h/Thermal column
- 1h/Rotary rack
4÷6 h
- 1200s for Ni and Lu monitors
- 1800s for combination
~1 d
- 7200s for Zr monitor and
combination
~3 d
- 900s for Au monitor
- 7200s for Zr monitor
- 10800s for Ni and
Lumonitors, and combination
+Al-0,1%Au wire (~5mg)
+Al-0,1%Lu wire (~5mg)
+ 99.8%Zr foil (~10mg)
+99.98%Ni foil (~30mg)
Products
[T1/2, E (keV)]
Counting time (Tc)
After an appropriate decay time for each
isotope, the samples were measured with the
gamma spectrometry using HPGe detector
(FWHM ~ 2.2 keV at 1332 keV). The samples
were placed at 14 cm from the detector surface.
In order to determine the neutron spectrum
65
Ni [2.5h, 366.3, 1115.5,
1481.8], 176mLu [3.6h, 88.4]
97
97
Zr [16.7h, 743.4]
Nb [16.7h, 657.9]
198
Au [2.7d, 411,8]; 177Lu
[6.7d, 112.9, 208.4]; 95Zr
[64d, 756.7]; 95Nb [64d,
765.8]; 58Co [70.8d, 810.8]
parameters simultaneously, monitors were
combined and measured at 0.5 hours, 1 hour,
and 3 hours with the decay time of 6 hours, 1
day and 3 days, respectively. The k0-IAEA
software was employed for the treatment of
experimental data. For the purpose of quality
71
CHARACTERIZATION OF NEUTRON SPECTRUM PARAMETERS AT …
assigned value; and (5) 3.28<|Uscore|, the
laboratory result is significantly different from
the assigned value [4].
control of the analytical procedure, 30 mg, 70
mg and 100 mg samples of the standard
reference material named NIST-679 (Brick
Clay) were irradiated at 45 sec, 1 hour, and 10
hours, respectively. The U-score is calculated
according to the following equation: 𝑈𝑠𝑐𝑜𝑟𝑒 =
2
𝜎𝐴𝑛𝑎
III. RESULTS AND DISCUSSION
A. Neutron spectrum parameters at the
irradiated channels of the DNRR after full core
conversion to LEU fuel
2
𝜎𝐶𝑒𝑟𝑡
,
(𝑋𝐴𝑛𝑎 − 𝑋𝐶𝑒𝑟𝑡 )/
+
where:
XAna, and XCert are the analytical results, and
certificated values, Ana,and Cert are the
uncertainty of XAna, and XCert. The results of
Neutron spectrum parameters at the
channel 13-2, thermal column, channel 7-1,
and rotary rack of the DNRR after full core
conversion to LEU fuel are given in Table II,
III, IV and V. In order to study the stability of
the neutron field at irradiated channels, the
experiments at channels 13-2, 7-1, and
thermal column (Table II, III and IV) were
repeated three times in three different
operation cycles of the reactor. However, at
the rotary rack (Table V), the parameters were
obtained only from two experiments (in
March, and April 2012).
the laboratory are interpreted according to the
5 possible evaluation classes as follows: (1)
|Uscore|1.64, the laboratory result does not
differ significantly from the assigned value; (2)
1.64<|Uscore|<1.96, the laboratory result
probably does not differ significantly from the
assigned value; (3) 1.96<|Uscore|<2.58, it is not
clear whether the laboratory result differs
significantly from the assigned value; (4)
2.58<|Uscore|<3.28, the laboratory result is
probably significantly different from the
Table II. Neutron spectrum parameters at the channel 13-2 after core coversion of the DNRR.
Parameters
Experimental period
Average ± SD
Aug. 2012
Feb. 2013
Mar. 2013
th(× 1012 n/cm2/s)
4.21 0.17
4.34 0.17
4.070.09
4.21 0.14
fast(× 10 n/cm /s)
6.220.39
7.610.75
6.010.39
6.610.87

f
-0.073  0.009
-0.068  0.019
-0.067  0.004
-0.069  0.003
13.1  0.3
10.8  0.2
8.3 0.7
10.7  2.4
Tn (K)
317 5
307  9
312  11
312  5
12
2
Table III. Neutron spectrum parameters at the thermal column after core coversion of the DNRR.
Parameters
Average ± SD
Experimental period
Jul. 2012
Mar. 2013
Apr. 2013
th(× 10 n/cm /s)
1.26  0.54
1.24 0.03
1.27  0.09
1.21  0.27
8.99 0.06
8.440.49
8.03 0.06
8.29 0.11
11
8
2
2

f
-
-0.117  0.032
-0.094 0.167
-0.140  0.015
190  8
195  4
198  2
197  4
Tn (K)
306  6
298  7
291  8
297  3
72
CAO DONG VU et al.
Table IV. Neutron spectrum parameters at the channel 7-1 after core coversion of the DNRR.
Parameters
Experimental period
Average ± SD
Mar. 2012
Apr. 2012
May2012
th(× 1012 n/cm2/s)
4.30 0.14
4.12 0.18
4.24 0.12
4.22 0.04
3.860.35
3.690.10
4.140.23
3.900.23

-0.022  0.032
-0.041  0.025
-0.031  0.028
-0.031  0.009
f
9.6  0.9
10.2  0.4
9.3  0.7
9.7  0.5
Tn (K)
300  5
300  5
301  5
300  0.6
12
2
Table V. Neutron spectrum parameters at the rotary rack after core coversion of the DNRR.
Parameters
Experimental period
Average ± SD
Mar. 2012
Apr. 2012
th(× 1012 n/cm2/s)
3.68 0.04
3.84 0.15
3.760.11
fast(× 1012 n/cm2/s)
0.31 0.05
0.32 0.04
0.32 0.01

0.099  0.010
0.104  0.010
0.102  0.003
f
30.1  2.5
30.0  1.0
30.1  0.4
Tn (K)
294  6
297  6
295  2
results in Table VI the absolute value of α at
channel 7-1 increases by approximately 1.7
times at negative side after full core
conversion. This means that the neutron
spectrum at the channel 7-1 becomes 'harder'
rather than that of before conversion. At the
rotary rack, the α factor significantly increases
by 2.5 times at positive sign. This means that
epi-thermal neutron spectrum at this position
tends to deviate below the 1/E distribution [5].
B. Comparison of the neutron spectrum
parameters before and after full core conversion to
LEU fuel
Table VI shows the thermal (th) and
fast (fast) neutron fluxes,  coefficient, and f
at the channel 7-1, and rotary rack measured
before [3] and after the full core conversion.
The obtained results in Table VI show that
after full core conversion, thermal neutron flux
reduces 8% at channel 7-1, and 6% at rotary
rack whereas the fast neutron flux at channel
7-1 and rotary rack increases by 2% and 6%,
respectively. This means that epi-thermal
neutron flux also increases (f decreases)
leading to the occurrence of the interference
reactions in k0-NAA such as (n, p), (n, n')
etc.[4]. On the other hand, also from the
As the old channel 13-2 was removed
from the core in November 2006, and a new
pneumatic transfer system together with the
channel 13-2 was reinstalled in June 2012,
therefore, there are no data for neutron
spectrum
at
channel
13-2
during
2006÷2011 period.
73
CHARACTERIZATION OF NEUTRON SPECTRUM PARAMETERS AT …
Table VI. The thermal and fast neutron flux at channel 13-2 and Rotary rack before
and after full core conversion of the DNRR.
th (n/cm2/s)
fast(n/cm2/s)
α
f
[3], measured in 2010 with HEU-LEU fuel
Channel 7-1
4.59 × 1012
3.81 × 1012
-0.019
11.09
Rotary rack
4.01 × 1012
0.30 × 1012
0.040
42.28
This work, measured in 2012 with LEU fuel
Channel 7-1
4.22 × 1012
3.90 × 1012
-0.031
9.70
Rotary rack
12
12
0.102
30.10
3.76 × 10
0.32 × 10
This work/[3], 7-1
0.92
1.02
1.67
0.87
This work/[3], Rotary rack
0.94
1.06
2.54
0.71
Table VII presents the thermal neutron
flux values at the channel 13-2 and thermal
column measured in 2003 [5] and after full
core conversion. The results from Table VII
show that after the core conversion, the thermal
neutron fluxes at channel 13-2 reduce 9% and
increase approximately 21 times at thermal
column. This unusual change at the thermal
column does not result from the core
conversion, but mainly relates to the
modification of structure of the thermal column
which was installed together with a new
pneumatic transfer system in 2012. The new
facility for thermal column was put close to the
graphite reflector in which the sample was
placed 10.8 cm deeper in contrast to the old
irradiated position.
Table VII. Themal neutron flux before and after full core conversion.
Thermal column
Channel 13-2
HEU (2003) [5]
5.80 × 109
4.62 × 1012
LEU (2012)
1.24 × 1011
4.21 × 1012
LEU/HEU
21.38
0.91
C. Analysing of SRM NIST-679 (Brick clay)
using obtained neutron parameters
Tables VIIIa and VIIIb show that the
|Uscore| for all analytical values are less than
1.64, which means that all results are
acceptable. This analysis also shows that it is
necessary to re-characterize the neutron
spectrum parameters after the core conversion.
Nevertheless, the data obtained from this study
are reliable and can be used to calibrate the
irradiated channels for k0-NAA at the DNRR.
To assess the quality of the neutron
spectrum data set obtained through this study,
the SRM named NIST-679 was analyzed by k0NAA. The analytical results obtained before
and after the core conversion are given in
Table 8a and Table 8b, respectively.
74
CAO DONG VU et al.
Table VIIIa. Analytical results of SRM NIST-679 before the core conversion.
No.
1
2
3
4
5
6
7
8
Element
Al
Dy
Mn
As
La
Fe
Sc
Th
Analyzed value
Conc.
Unc.
103500
5208
6.95
1.98
1764
436
8.9
3.1
50.0
12.4
92133
6168
22.1
2.3
13.47
1.63
Certified value
Conc.
Unc.
110100
3400
7.15
0.27
1852
45
9.5
0.2
49.9
0.5
90500
2100
22.8
0.2
13.46
0.12
Uscore
Position
-1.06
-0.10
-0.20
-0.19
0.01
0.25
-0.30
0.01
7-1
7-1
7-1
RR
RR
RR
RR
RR
Table VIIIb. Analytical results of SRM NIST-679 after the core conversion.
No.
1
2
3
4
5
6
7
8
Element
Al
Dy
Mn
As
La
Fe
Sc
Th
Analyzed value
Conc.
Unc.
106500
8758
6.4
1.3
1742
116
8.3
1.42
45.5
2.77
92880
3001
21.9
2.5
13.2
0.2
Certified value
Conc.
Unc.
110100
3400
7.15
0.27
1852
45
9.5
0.2
49.9
0.5
90500
2100
22.8
0.2
13.46
0.12
Uscore
Position
-0.38
-0.56
-0.88
-0.84
-1.56
0.65
-0.36
-1.11
7-1
7-1
7-1
RR
RR
RR
RR
RR
IV. CONCLUSION
REFERENCES
Re-establishment of the neutron spectrum
parameters including th, fast, , f, and Tn at
four irradiated channels for NAA at the DNRR
after full core conversion to LEU fuel was
carried out.
[1] N.N. Dien, Project of fuel conversion at Dalat
research reactor, Dalat Nuclear Research
Institute (2011).
[2] N.N. Dien, Report on the physics start-up for
conversion to LEU fuel at Dalat research reactor,
Dalat Nuclear Research Institute, (2012).
[3] C.D. Vu, Project report (code CS/09/01-01)
After replacement of the core with LEU
fuel assemblies, the thermal neutron flux in
most of irradiated channels decreases by 6÷9%
while the epi-thermal neutron flux and fast
neutron increase by 2÷6%; neutron spectrum
becomes‘harder’ in most of the investigated
positions.
Study on application of k0-IAEA at Dalat
research reactor, Vietnam Atomic Energy
Institute (2010).
[4] H.M. Dung*, M.C. Freitas, J.P. Santos, J.G.
Marques, Re-characterization of irradiation
facilities for k0-NAA at RPI after conversion to
LEU fuel and re-arrangement of core
configuration, Nuclear Instruments and Methods
New neutron spectrum parameters
obtained through this study will be useful for
characterization of the irradiation channels in
k0-NAA analytical procedure at the DNRR
after full core conversion to LEU fuel.
in Physics Research A 622, 438–442 (2010).
[5] H.M. Dung, Study for development of k-zero
Neutron Activation Analysis for multi-element
characterization, PhD thesis, the Natural
Science University, Hochiminh city (2003).
75
Some results of NAA collaborative study in white rice
performed at Dalat Nuclear Research Institute
T.Q. Thien*, C.D. Vu, H.V. Doanh, N.T. Sy
Dalat Nuclear Research Institute
01 Nguyen Tu Luc St., Dalat, Lam Dong
* Email: tqthien0613104@yahoo.com
Abstract: White rice is a main food for Asian people. In the framework of Forum for Nuclear
Cooperation in Asia (FNCA), therefore, the eight Asian countries: China, Indonesia, Japan, Korea,
Malaysia, the Philippines, Thailand and Vietnam selected white rice as a common target sample for a
collaboration study since 2008. Accordingly, rice samples were purchased and prepared by following
a protocol that had been proposed for this study. The groups of elements that were analyzed by using
neutron activation analysis in the white rice samples were toxic elements and nutrient elements,
including: Al, As, Br, Ca, Cl, Co, Cr, Cs, Fe, K, Mg, Mn, Na, Rb and Zn. The analytical results were
compared between the different countries and evaluated by using the Tolerable Intake Level of World
Health Organization (WHO) and Recommended Dietary Allowance or Adequate Intake (AI) of the
U.S. Institute of Medicine (IOM) guideline values. These data will be very useful in the monitoring of
the levels of food contamination and in the evaluation of the nutritional status for people living in
Vietnam and other Asian countries.
Keywords:White rice, neutron activation analysis, FNCA,tolerable intake level, dietary reference
intakes, adequate intake.
I. INTRODUCTION
FNCA (Forum for Nuclear Cooperation
in Asia) was formally established in March
1999 at the 10th session of the International
Conference on Nuclear Cooperation in Asia
region ICNCA (International Conference for
Nuclear Cooperation in Asia) initiated and
funded by the Japanese government. FNCA is
supposed to enhance mutual understanding,
exchange of information and experience to social
and economic development in Asia through
research, collaboration, technology applications
initiatives for peaceful purposes. Up to 2012,
FNCA has 12 member countries, including:
Australia, Bangladesh, China, Indonesia,
Malaysia, Japan, Kazakhstan, Korea, Mongolia,
the Philippines, Thailand and Vietnam.
NAA (Neutron Activation Analysis) is
one of the projects under the ResearchReactor
Utilization in the framework of the forum
FNCA. Vietnam has participated in the FNCA
since 2000.
In the FNCA workshop held in Dalat,
Vietnam, in 2008, the eight among twelve
member countries of the FNCA which are
China, Indonesia, Malaysia, Japan, Korea, the
Philippines, Thailand and Vietnam, agreed to
participate in a collaborative study on the
analysis of food samples as a sub-project
thematic in NAA. White rice has been selected
as research subjects for this work because of its
importance as the basic staple food for people
wholives in Asia. Specifically, the major rice
producing countries in Asia are China, India,
Indonesia, Malaysia, Bangladesh, Thailand,
Vietnam, etc. These countries accounts for over
80% of production and consumption of rice in
the world. This highlights the importance of the
TRAN QUANG THIEN et al.
information gained from the study because rice
is the staple food as well as providing a large
portion of the calories in the Asian diet [1].
II. EXPERIMENTS
A. Sample collection and preparation
Eighteen samples were collected from the
Department of Agriculture and Rural
Development Centre Tiengiang province
agricultural seed wherein rice is the most
common type on the market which are
presented in Table I.
The objective of this study was to
determine the inorganic elements in the white
rice of Vietnam and compared it with seven
Asian countries by NAA method, these results
are preliminary by the level of nutrients and
toxic elements in rice for safety.
Table I.The information sampling of Vietnam’s rice samples at Department of Agriculture and Rural
Development Centre Tiengiang province
No.
Type
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
Ham Chau Rice
IR 50404 Rice
Japan 504 Rice
Jasmine 85 Rice
Jasmine Rice
OM 4218 Rice
OM 4900 Rice
OM 5451 Rice
OM 5472 Rice
OM 5976 Rice
OM 6162 Rice
OM 6377 Rice
OM 6976 Rice
Otim Rice
Seri Rice
Tai Nguyen Rice
Taiwan Fragrant Rice
Thom Lai Rice
The collected rice samples were brought
to the lab and washed with distilled water and
then dried in a drying oven at a temperature of
60 0C for 4 hours, then ground into fine
particles using an agate mortar in order to
prevent contamination. Rice samples were
repeatedly ground until a particle size of 60
meshes. Finally, the samples were subdivided
into subsamples weighing from 100-300 mg
prior to analysis by INAA.[1]
B. Analysis
The rice samples were analyzed by
INAA in Dalat Nuclear Research Institute. The
analytical procedures are followed with
ISO/IEC 17025 [2]. A concurrent analysis of
reference standard samples for quality control
was made for each batch of analysis. Analyses
were made using a combination of both short
and long irradiations. The HPGe detector with
77
SOME RESULTS OF NAA COLLABORATIVE STUDY IN WHITE RICE ...
multichannel analysis system was used to
The result of standard reference material
was shown in Table II and the result of fifteen
elements concentrations in eighteen samples of
white rice are in Table III.
measure the gamma rays from the sample after
irradiation. The concentration of the elements
was calculated using the relative method and/or
k-zero method.
In Table II, The average result
caculated through 3 times analysis, it's
much different to the value of certificate.
Z-score of all elements is lower than 2,
mean this results are satisfactory.
III. RESULTS AND DISCUSSION
A. The content of elements in white rice
samples
are
not
The
that
Table II. Result of standard reference material IAEA-V-10
No.
Ele.
Aver.
1
Mg
1579
2
Ca
20865
3
Cl
7360
4
Mn
46
5
Na
507
6
K
20119
7
Br
7.3
8
Sc
0.016
9
Cr
6.6
10
Fe
196
11
Co
0.15
12
Zn
25.5
13
Rb
7.7
Aver: Average result;
Sd: Standart deviation
Cert: Certificate
Sd.
Cert.
Z-score
Ana/Cert
121
2892
100
5
9
3530
0.5
0.002
0.5
21
0.03
2.2
0.5
1360
21600
47
500
21000
8
0.014
6.5
186
0.13
24
7.6
1.81
-0.25
-0.20
0.78
-0.25
-1.40
1.00
0.20
0.48
0.67
0.68
0.20
1.161
0.966
0.979
1.014
0.958
0.913
1.143
1.015
1.054
1.154
1.063
1.013
In Table III, the concentration of Mg
element are not analyzed in all samples, elements
concentration of Al, Ca and Fe are not obtained
and reported limit of detection, the result of other
elements are included concentration and
uncertainty. The highest concentration are K
element, the lowest come from Co and Cs. The
other elements have no significant differences
between all samples except Rb.
Zn in rice samples determined by eight
participating countries are summarized in Table
4. The results of quality control analysis for
fifteen elements are summarized as a relative
error (%) with absolute value and are shown in
Fig. 1. The relative error of most of the
elements evaluated in Fig. 1 were less than
15%, except for some few elements such as Al
of Malaysia; Co of Vietnam; Mg of China,
Korea and Vietnam, Mn and Na of Korea rice
samples.
B. Comparing the elements concentration in
white rice of 8 countries
Results of fifteen elements: Al, As, Br,
Ca, Cl, Co, Cr, Cs, Fe, K, Mg, Mn, Na, Rb and
78
Table III.The analytical results of eighteenwhite rice samples in Vietnam
Al
As
Br
Ca
Cl
Co
Cr
Cs
Fe
K
Mg
Mn
Na
Rb
Zn
No.
Type
1
Ham Chau
Rice
<3
0.12 0.02 0.43 0.05 <165
294 11 0.031 0.006 <0.4
0.026 0.007 <14
518 10 NA.
8.0 0.2 22.6 0.3 1.5 0.3 23.0 0.5
2
IR 50404 Rice
<8
0.16 0.04 0.17 0.05 <160
242 11 0.026 0.008 <0.5
0.061 0.010 <10
1649 17 NA.
15.9 0.2 11.5 0.2 11.4 0.6 21.3 0.6
3
Japan 504
Rice
<4
0.06 0.02 0.78 0.07 <120
407 14 0.022 0.008 <0.4
0.019 0.008 <14
527 11 NA.
5.1 0.1 48.5 0.3 1.0 0.3 21.7 0.8
4
Jasmine 85
Rice
<5
0.13 0.03 0.31 0.07 <185
235 30 0.028 0.008 <0.5
0.056 0.012 <18
1543 17 NA.
19.1 0.1 18.2 0.3 10.3 0.7 22.8 0.8
5
Jasmine Rice
<4
0.12 0.02 0.22 0.04 <100
192 10 0.030 0.008 <0.5
0.052 0.007 <20
453 9 NA.
3.5 0.1 9.0 0.2 1.9 0.4 19.2 0.6
6
OM 4218 Rice
<5
0.15 0.04 0.22 0.06 <130
282 25 0.031 0.008 <0.6
0.045 0.009 <17
1548 16 NA.
12.8 0.1 17.1 0.3 8.3 0.6 22.2 0.7
7
OM 4900 Rice
<7
0.10 0.03 0.34 0.07 <140
346 31 0.043 0.008 <0.5
0.044 0.010 <19
1780 18 NA.
16.0 0.1 14.8 0.3 5.9 0.5 26.2 0.7
8
OM 5451 Rice
<7
0.15 0.03 0.29 0.06 <120
199 23 0.036 0.007 <0.6
0.046 0.008 <16
1254 16 NA.
11.8 0.1 18.8 0.3 11.4 0.7 22.6 0.7
C. U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
C.
U.
9
OM 5472 Rice
<4
0.11 0.03 0.26 0.06 <100
293 22 0.034 0.007 <0.5
0.047 0.011 <22
1414 15 NA.
11.9 0.1 14.9 0.2 9.7 0.6 24.8 0.7
10
OM 5976 Rice
<5
0.11 0.03 0.18 0.05 <110
186 11 0.031 0.009 <0.6
0.043 0.011 <17
1227 14 NA.
12.7 0.1 11.3 0.2 10.7 0.6 24.0 0.7
11
OM 6162 Rice
<1
0
0.08 0.03 0.17 0.05 <120
307 13 0.043 0.009 <0.4
0.039 0.011 <16
1509 16 NA.
13.9 0.2 12.6 0.2 8.3 0.6 23.7 0.7
12
OM 6377 Rice
<7
0.14 0.03 0.23 0.06 <120
266 27 0.058 0.009 <0.6
0.063 0.011 <13
1636 16 NA.
16.8 0.1 13.1 0.2 15.7 0.7 23.4 0.8
13
OM 6976 Rice
<6
0.11 0.03 0.34 0.05 <100
382 14 0.027 0.007 <0.5
0.038 0.009 <19
1487 16 NA.
6.7 0.3 13.5 0.2 4.8 0.5 25.0 0.7
14
Otim Rice
<4
0.17 0.03 0.47 0.04 <100
225 11 0.044 0.009 <0.3
0.051 0.013 <28
568 11 NA.
7.5 0.1 14.4 0.2 2.8 0.5 21.4 0.7
15
Seri Rice
<4
0.22 0.02 0.18 0.04 <135
236 11 0.041 0.009 <0.4
0.074 0.011 <21
501 9 NA.
5.4 0.1 13.5 0.2 8.2 0.7 21.1 0.7
16
Tai Nguyen
Rice
<3
0.06 0.02 0.65 0.07 <100
378 12 0.039 0.009 <0.5
0.016 0.007 <16
470 9 NA.
5.2 0.1 40.9 0.3 1.0 0.3 17.5 0.7
17
Taiwan Rice
<5
0.11 0.03 0.23 0.04 <110
330 12 0.024 0.010 HL
0.066 0.010 <22
842 13 NA.
5.9 0.1 10.5 0.2 7.5 0.6 24.8 0.9
Thom Lai Rice <3
0.08 0.02 0.40 0.04 <110
208
0.029 0.007 <14
463 8 NA.
3.9 0.1 7.8 0.2 1.2 0.4 19.8 0.6
18
Average
<5
0.12
0.33
<124
9
0.031 0.008 <0.5
278
0.034
<0.5
Unit: mg/kg; C. : Concentration; U. : Uncertainty; NA. : Not Applicable
79
0.045
<18
1077
NA.
10.1
17.4
6.8
22.5
Table IV. The analytical results of white rice (unit: mg/kg) [1]
e
Philippine
Thailand
Vietnam
(This work)
<2
<2.82
<2.33
<5
0.13
0.11
0.07
0.09
0.12
0.5
0.19
13.6
5.35
0.43
0.33
<4.53
49.5
53.9
<10
39.1
<15
<124
264
210
239
193
225
236
239
278
Co
<0.3
0.77
N.A
0.005
0.026
N.A
0.022
0.034
Cr
0.25
0.38
N.A
<0.01
<0.08
N.A
<0.4
<0.5
Cs
<0.07
0.09
N.A
0.009
0.016
N.A
N.A
0.045
Fe
N.A
4.65
N.A
1.58
<5
N.A
<16
<18
K
977
739
611
660
573
637
620
1077
Mg
379
131
149
241
<150
90
59
N.A
Mn
9.25
9.95
7.66
9.06
6.19
7.89
9.23
10.1
Na
10.3
7.7
5.69
4.1
13.7
5.17
4.58
17.4
Rb
<3.35
7.64
N.A
1.39
2.1
3.24
1.34
6.8
Zn
15.3
24.2
18.5
15.3
10.1
15.4
21.4
22.5
b
Japan
Korea
<20.52
<1.66
<1.38
0.55
0.08
0.1
Br
0.35
0.45
Ca
N.A
Cl
Ele.
a
China
Indonesia
Al
<4.46
As
c
Malaysia
d
N.A: not applicable; (a) Mean values are derived from four different samples; (b) Mean values from a sample
of known origin and two samples of unknown origin; (c) Mean values are derived from two different
samples; (d) Mean values from four samples of unknown origin; (e) This work, average value of eighteen
samples from known origins.
As can be seen from Table IV, the Al
concentrations in rice saples from all
participating countries were below the
detection limit, hence only the limit of
detection (LOD) were reported. Indonesia had
an LOD value of 20.52 for Al, highest
compared to other countries. Korea and Japan
had the lowest LOD values in the eight
countries. K concentration range is from 553 to
1077 mg/kg. K in Vietnam rice samples had
the highest value which is 1077 mg/kg. Cl and
Mg have similarly eminent values. Seven
elements of As, Br, Cl, K, Mn, Na and Zn were
determined by all participating countries, but
LODs were not reported. As content of China
had the highest value, 0.55 mg/kg and the other
countries have equivalent levels of As, 0.1
mg/kg. Br concentrations of Malaysia,
Indonesia and the Philippines were more than a
dozen times higher than those of other
countries. Five elements Cl, K, Mn, Na and Zn
did not differ significantly and the average
content of the standard deviation were 236±27,
737±178, 8.67±1.32, 8.58±4.84 and 17.8±4.7
mg/kg respectively. Concentrations of Mg
were reported by six countries excluding
Malaysia and Vietnam. Thailand showed the
lowest levels of Mg, 59 mg/kg, while the Mg
content of China was the highest at 379 mg/kg.
Only three countries namely Japan, South
Korea and the Philippines reported Ca data
which were 49.5, 53.9 and 39.1 mg/kg
respectively. In addition, the levels of Cr, Cs
and Fe in Indonesian rice were higher
compared to those of other countries.
80
C. Dietary intake level of the toxic elements
rice consumption varies in different countries,
and therefore a consensus value of 300
grams/day was set, to be able to compare the
intake of As, Cl, K, Mn, Na and Zn from rice
consumption in all participating countries. This
is to assess whether or not, the ingested levels
of the elements can be considered as harmful
or beneficial to human health. Data are shown
in Table V.
and nutrition elements of 8 countries
To estimate the dietary intake level of
inorganic constituents on consumption of white
rice, it was necessary to conduct a survey of
daily consumption of rice. For example, the
amount of the average daily consumption of
rice in Korea in 2000 was 256 grams, or in
Vietnam in 2010 is 360 gram [3, 4]. However,
Error, %
China
Indonesia
Japan/Philippines
25
20
15
10
5
0
Al As Br Ca Cl Co Cr Cs Fe
K Mg Mn Na Rb Zn
Fig. 1. The absolute value of the relative error (%) of the value analysis to value certification/reference.
Table V. The RDA value of 6 elements each day through white rice, assuming consumption
of 300 grams /day for adults[1]
Ele.
China Indonesia
Japan
Korea Malaysia Philippine Thailand
Vietnam
(This work)
As (µg)
165
24
30
39
33
21
27
36
Cl (mg)
79.2
63
71.7
57.9
67.5
70.8
71.7
83.4
K (mg)
293
222
183
198
172
191
186
323
Mn
(mg)
2.78
2.99
2.30
2.72
1.86
2.37
2.77
3.03
Na (mg)
3.09
2.31
1.71
1.23
4.11
1.55
1.37
5.22
Zn (mg)
4.59
7.26
5.55
4.59
3.03
4.62
6.42
6.75
The WHO has established a Tolerable
Intake Level for weekly consumption, which is
15 mg/kg of body weight for As [5]. Assuming
a body weight of 70 kg of an adult, the
Tolerable Intake Level for As daily
consumption will be 150 microgram As. In
addition, the Institute of Medicine (IOM) in the
United States has established the value of the
Recommended Dietary Allowance (RDA) or
adequate intake (AI) for the necessary elements
[6, 7]. Zn has the highest RDA of 11 mg/day
for men. AI highest values for Cl, Mn defined
by the IOM is 2.3 g/day for all adults, for Na
and K, the highest AI values are respectively
1.5 and 4.7 g/day.
81
Calculations for the RDA or AI for the
elements As, Cl, K, Mn, Na, Zn are shown in
Figure 2.Tolerable Intake Level of As in China
is higher than Tolerable Intake Level of WHO
which was about 10%, for the other countries.
The level of Mn is almost equal to the value of
the RDA of IOM. This shows just rice
consumption of 300 g/day may provide
sufficient Mn necessary for the human body.
The intake level for the remaining elements
(Cl, K, Na and Zn) were below the RDA or AI.
In the case of Zn, the range of daily
consumption
from
21.6%
(Malaysia,
Indonesia) to 51.9% (Indonesia) can only
supply approximately 21.6% to 51.9% Zn
necessary for the human body. Similarly,
consumption of Cl at 2.5% to 3.6%, K at 3.7%
to 6.2% and 0.3% Na were below the
recommended values. These essential elements
can be obtained anyway, from other foods such
as meat, fish, vegetables, eggs, milk, etc. which
are eaten together with the rice.
countries namely China, Indonesia, Japan,
Korea, Malaysia, the Philippines, Thailand and
Vietnam. A total of fifteen elements in thirty
five samples of white rice collected from eight
countries were determined by INAA method.
Within the framework of project participants
FNCA/NAA, NAA laboratory of Vietnam has
collected and analyzed fifteen elements in
eighteen samples of white rice types. Results of
Vietnam’s rice has been compared with the
results of the seven countries participating
members.
The analytical data were compared
between the participating countries and
assessed according to the daily intake using the
guideline values set by the WHO and IOM.
The results showed an elevated amount of As
in Chinese rice which exceeded by
approximately 10%, the RDA recommended
by WHO. In addition the research gave an
overview of the levels of nutritional elements
Na, Mn, Cl, K and Zn in rice consumed in the
eight countries. Information on the intakes of
Mn (of approximately 100%), Zn, Na, Cl
(21.6÷51.9) % and K (lower than 10%) in
comparison to the requirements of IOM was
obtained from the study.
IV. CONCLUSIONS
A
collaborative
study
on
the
determination of elemental abundance in rice
using NAA was participated in by eight
China
Malaysia
Indonesia
Philippines
Japan
Thailand
Korea
Vietnam
%RDA
1000.0
100.0
10.0
1.0
0.1
As
Cl
K
Mn
Na
Zn
Fig. 2. Assess daily nutrient consumption (%) for the six elements through white rice.
82
[3]
In future, FNCA will carry on to expand
the scope of research in elemental abundance in
food samples to strengthen the collaboration
between Asian countries for the continued
application of NAA in the assessment for
contamination and mineral potentiality in the
basic foodstuffs.
Ministry
of
Agriculture
and
Forestry,
Agricultural and forestry statistical yearbook
2003. Ministry of Agriculture and Forestry,
Seoul, (2003).
[4] National Institute of Nutrition, A review of the
nutrition situation in Vietnam 2009-2010,
Medical Publishing House, Hanoi, (2011).
[5] World Health Organization, Evaluation of
certain food additives and contaminants,
(Thirty-third report of the Joint FAO/WHO
Expert Committee on Food Additives). WHO
Technical Report Series, No. 776, (1989).
ACKNOWLEDGEMENTS
We would like to thank the MEXT of
Japan for support of this research.
[6] Institute of Medicine, Food and Nutrition
Board, Dietary reference intakes for vitamin A,
REFERENCES
[1] J. H. Moon et. al, A NAA collaborative study
in white rice performed in seven Asian
countries, Journal of Radio- analytical
vitamin K, arsenic, boron, chromium, copper,
iodine, iron, manganese, molybdenum, nickel,
silicon, vanadium, and zinc, National Academy
Chemistry, Volume 291, Issue 1, pp 217-221
(January 2012).
of Sciences, Washington DC, (2001).
[7] Institute of Medicine, Food and Nutrition
Board, Dietary reference intakes for water,
potassium, sodium, chloride and sulfate.
National Academy of Sciences, Washington
DC, (2004).
[2] Center for Analytical Techniques (CATech),
Dalat Nuclear Research Institute (NRI),
“TCCS-MSH from 01 to 03”, Dalat, (2011).
83
A new rapid neutron activation analysis system
at Dalat nuclear research reactor
H.V. Doanh*, C.D. Vu, T.Q. Thien, P.N. Son, N.T. Sy, N. Giang and N.N. Dien
Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat, Vietnam
*
E-mail: hovandoanh@gmail.com
Abstract: An auto-pneumatic transfer system has been installed at the Dalat research reactor for rapid
instrument neutron activation analysis based on very short-lived nuclides. This system can be used to
perform short irradiations in seconds either in the vertical channel 13-2 or in the horizontal thermal
column of the reactor. The transferring time of sample from irradiation to measurement position is
approximately 3.2 seconds. A loss-free counting system using HPGE detector has been also setup in
compacting with the pneumatic transfer system for measurement of sample’s activity, automatically
starting for data acquisition at irradiated sample’s arrival. This new facility was tested and shown to
have high potential for the determination of short-lived nuclides with half-lives from 10  100
seconds. This work presents the results of timing parameter measurements, characterization of
irradiation facilities, and application of this system to determining Selenium concentration in several
biological reference materials.
Keywords: Auto-pneumatic transfer system, neutron activation analysis, short-lived nuclides.
I. INTRODUCTION
Instrumental neutron activation analysis
(INAA) has been developed and applied at the
500 kW Dalat research reactor (DNRR) since
1984. Until now, it is capable of analyzing
more than 40 elements based on radionuclides
with short, medium and long-lived time. For
short-lived nuclides with half-lives from 2
minutes to 2.6 hours, samples are often
irradiated at the neutron channel No.7-1 of
Dalat research reactor through a semi-auto
pneumatic transfer system (PTS) with valid
irradiation time from 45 seconds to 20
minutes. Measurements are often performed
using a gamma spectrometer coupled with a
HPGe (GMX-30190), but with manual
manipulation between loading and counting
procedures. Therefore, the shortest-lived
nuclides that could be detected are 28Al (T1/2 =
2.24 min), 52V (T1/2 = 3.75 min), and 51Ti (T1/2
= 5.76 min).
In the recent years, through the IAEA
TC Project RER/4/028, a new automatic PTS
for rapid neutron activation analysis based on
short-lived nuclides has been developed. This
facility consists of three main parts introduced
in reference [1]. The first part, consisting of
two aluminum irradiation tubes, which are
inserted into the vertical channel No.13-2 and
the horizontal thermal column (TC) of the
reactor. The second part is a digital signal
processing spectrometer connected to a 40%
relative efficiency HPGe detector coupled with
a transistor reset preamplifier. The third part is
composed of pneumatic chambers, loading and
sliding devices in Cabin-1 which facilitates the
fully
automatic
irradiation-counting
procedures. It has also a sample automatic
loader for the sequential routing of the samples
in multi-samples operation mode: when the
measurement for one sample is finished, the
next sample is loaded and sent to the
irradiation and then counting positions. In this
HO VAN DOANH et al.
77m
system, there is also has optical sensors for
controlling the transport of the capsule as a
sample carrier, and for accurately measuring
the capsule flight time from irradiation position
to detector. The installation diagram is shown
in Fig. 1.
Se,
179m
Hf,
46m
Sc, and
110
Ag can be used for
INAA at Dalat reactor, which the former
system can not detect.
The main purpose of this work is to test
the system for both mechanical and analytical
reliability. A systematic study has been carried
out including measurements for timing
parameters of the system and neutron flux at
irradiation positions, and the application of this
system to determining of Selenium in a number
of biological reference materials for validation
purpose.
This PTS system can be used to perform
short irradiations in seconds. The return time of
sample from irradiation position to counting
position is about 3.2 s. Timing information for
both irradiation and counting will be instantly
delivered to the activation analysis workstation
computer. The digital gamma spectrometer is
selected and tuned for accurate measurement at
high and varying counting rates, using loss-free
counting technology. Accordingly, shorterlived nuclides (half-life < 1 min) such as 20F,
Fig. 1. Diagram of the auto-pneumatic transfer system installed at DNRR.
85
A NEW RAPID NEUTRON ACTIVATION ANALYSIS SYSTEM AT …
II. EXPERIMENT
recorded at the computer via a signal from the
fast solenoid providing the return gas (START
of return time) and from the optical sensor at
the detector station recording the arrival of
capsule (END of return time).
A. Timing measurements
The accuracy of irradiation time of an
irradiation facility should be checked and
calibrated as a type of analytical qualify
control [2]. In this experiment, the absolute
irradiation time, the sample transferring time
from entrance to irradiation position of
aluminum tube (Tin) and the reverse movement
(Tout) were determined. The experiments were
done outside the reactor before installation of
irradiation facilities inside the reactor. The
arrangement of timing measurements is shown
in Fig. 2. The setup includes two NaI(Tl)
detectors placed at the top and bottom sides of
the irradiation tube. The detectors detect
gamma radiation pulses from a 131I source
inserted in a capsule while moving inside. The
counters are set for running in Multi-Channel
Scaler (MCS) mode; MCS mode records the
counting rate of events as a function of time.
B. Neutron spectrum parameters of irradiation
position
The neutron spectrum parameters
including thermal neutron flux th, fast neutron
flux f, the thermal flux to epithermal neutron
flux (epi) ratio were measured at sample
irradiation positions in channel No.13-2 and
thermal column using Au, Zr, and Ni monitors.
Monitors were inserted into a high purity
polyethylene vial and loaded into rabbit
(capsule) for irradiation. The activity
measurements were carried out by a calibrated
gamma-ray spectrometer combined with HPGe
detector (GMX-30190). The measured
spectrum were analyzed by using the k0-IAEA
program. The irradiation, decay and counting
times for each monitor are shown in Table I.
Typically monitors with masses of 4 mg for
Al-0.1%Au foil (IRMM-530R), 30 mg for pure
Ni (wire), 10 mg for Zr (foil) were irradiated
for 10 min at 13-2 channel (2 h at thermal
column), and the decay time is 1 day for 97Zr
and 3 days for 198Au, 95Zr and 58Co.
The return time from the irradiation to
the measurement positions was determined by
a series of irradiation (50 replicates) for a total
weight (capsule, vial and sample) of about 4.4
gram and air pressure of 3.1 bars over a
distance of 40 m for 13-2 channel and 36 m for
thermal column. The return times were
Fig 2. Arrangement for experiment of timing
measurements.
86
HO VAN DOANH et al.
Table I. The irradiation, decay and counting times for the monitors.
Time/position
irradiation (monitor,
mass)
10 min/ 13-2 channel
Decay
time
Counting time
(combination)
~1d
12h
97m
Nb (60 s, 743.4)*;
97
2 hours/ thermal column
(Al-0.1% Au, ~ 4 mg)
Nb (16.7 h, 657.9)
~3d
0.5  3 h (5 h)
198
Au (2.7 d, 411.8);
95
(99.8% Zr, ~ 10 mg)
Zr (64 d, 765.8);
58
(99.98% Ni, ~ 30 mg)
* Nuclide
Measured
radionuclides
(T1/2, -rays in keV)
Co (70.8 d, 810.8)
97m
97
Nb is decayed from nuclide Zr with half-life of 16.7h.
C. Determination of Selenium
were irradiated for 25 s, allowed 20 s delay
time to eliminate interference of 116mIn with a
half-life of 2.18 s [3, 4]) and counted for 25 s
at a distance of 10 cm from the detector
(GMX40-76-PL). The concentrations of
Selenium were determined by both k-zero and
relative methods.
A variety of reference materials (Tuna
Fish IAEA-436, Oyster tissue NIST 1566b,
Bovine Liver NIST 1577, Bovine Liver NIST
1577b) were selected to assess reliability of
this system on the short-time activation
application. All of the samples were irradiated
at a neutron flux of 4.21012 n.cm-2.s-1 in the
III. RESULTS AND DISSCUSION
13-2 channel and counted on the calibrated
HPGe gamma-ray spectrometer (GMX40-76PL).
A. Timing measurements
The results for average transferring time
of sample from the top to bottom of the
In order to evaluate the limit of detection
of Se in biological samples, two 200mg
replicates of each material (IAEA 436 and
NIST 1566b) were weighed and packed in high
purity polyethylene bags. The samples were
irradiated for 5, 10, 15, 20, 25, 30, 35 and 40 s.
After a delay of 3.2 s (including both
transferring time of sample from irradiation
position to detector and the time required to
start the detector). Each sample were counted
for 20 s at a distance of 10 cm from detector.
aluminum irradiation tube (Tin) is (0.628 
0.021) s for the channel No.13-2 irradiation
tube (a length of 6 m) and Tout is (0.323 
0.030) s (averaged for 90 runs over the three
days). For thermal column irradiation tube (a
length of 2.8 m), Tin is (0.248  0.019) s and
Tout is (0.146  0.004) s, as shown in Table II.
The result obtained for measuring the return
time from the irradiation position to the
measurement position was found to be (3.165 ±
0.002) s for channel No.13-2. That for thermal
To test accuracy for the analysis of the
Se concentration in biological reference
materials, four 200 mg replicates of each
material (IAEA 436, NIST 1566b, NIST 1577
and NIST 1577b) were weighed. The samples
column was (3.025  0.013) s. It should be
noted that this timing parameters are included
in the time required to start the detector after
receiving the start signal.
87
A NEW RAPID NEUTRON ACTIVATION ANALYSIS SYSTEM AT …
Table II. The result of time measurements.
Irradiation
position
The transferring time throughout
aluminum irradiation tube (second)
The return time from irradiation
position to detector position (second)
Tin
Tout
This word
Manufacturer*
13-2 channel
0.628  0.021
0.323  0.030
3.165 ± 0.002
3.301  0.013
Thermal column
0.248  0.019
0.146  0.004
3.025  0.013
3.261  0.022
* Sample weight:  8 g for thermal column tube and  6 g for 13-2 channel tube,
operation air pressure:  3.1 bars, distance: 30 meters.
There are significant differences
between this work and that of the manufacturer
in capsule sample weight and distance from
irradiation position to measurement position.
Hence, there are differences ( 7%) in the
result of the return time from irradiation
position to detector position. However, it is not
a problem for analytical measurements.
channel No.13-2 and 4.91% for thermal
column. The relative error is less than 1% at
irradiation time of 5 s for channel No.13-2, and
10 s for thermal column. The large error for the
first second is due to delay of the system in
starting the irradiation timer and in ejecting the
capsule once the “end of irradiation” signal has
been received.
Results for absolute irradiation time at
channel No.13-2 and thermal column were
determined by a series of irradiations ranging
from 1 to 30 s (3 replicates), as shown in Fig. 3
and Fig. 4. The relative error of irradiation time
in the first second is 16.02% for channel
No.13-2 and 26.43% for thermal column, and
those for irradiation time of 2 s is 1.5% for
This timing delay problem can be
adjusted through the control unit and the
software package for managing optimal
operation and the analytical procedures.
However, it is not a problem for INAA because
the time parameters remain unchanged for all
samples, standards, and control material
18
28
16
24
Relative error, %
Relative error, %
14
12
10
8
6
20
16
12
8
4
4
2
0
0
0
2
4
6
8
10 12 14 16 18 20 22 24 26 28 30
0
2
4
6
8
10 12
14 16
18 20 22
24 26
28 30
Irradiation time, second
Fig. 3. The relative error of irradiation time
for 13-2 channel.
Fig. 4. The relative error of irradiation time for
thermal column.
88
HO VAN DOANH et al.
energy [5], measured using the 58Ni(n,p)58Co
nuclear reaction. The thermal neutron flux at
the irradiation position in the thermal column
is 1.25E+11 n.cm2.s-1, associated with much
lower fast and epithermal neutron flux.
Hence, thermal column is a useful irradiation
channel for eliminating interference reactions
induced by fast neutron, in which sample is
irradiated in an extremely well thermalized
neutron field [6].
B. Neutron spectrum parameters of irradiation positions
The results of the determination of
neutron spectra parameters are shown in Table
III. This table includes data obtained for the
thermal, fast neutron flux, the ratio of thermal
to epithermal neutron flux (th/epi). The
thermal neutron flux at the irradiation position
in the channel No. 13-2 is 4.2E+12 n.cm2.s-1,
and associated with 0.5 times of epithermal.
The integral fast neutron flux is 6.61E+12
n.cm-2.s-1 for all neutrons above 2.9MeV in
Table III. The results of neutron spectra parameters at irradiation positions in the channel No.13-2
and thermal column of DNRR.
Irradiation position
 t h (n/cm2/s)
 F (n/cm2/s)
th / epi
13-2 channel
(4.2  0.1) x 1012
(6.6  0.9) x 1012
10.7  2.4
Thermal column
(1.24  0.03) x 1011
(8.4  0.5) x 108
195  4
C. Determination of Selenium
sensitivities for Se rapid determination in a
variety of biological matrices.
Finally, measurements of detection
limits of Se in IAEA 436 and NIST 1566b
samples were performed. The results for these
measurements are presented in Fig 4. The
obtained results confirm that in irradiation
from 15 s to 25 s at irradiation position of the
channel No.13-2 coupled with counting for
roughly 20 s at 10 cm distance from detector,
the detection limits for Se is within the range
0.5  0.7 ppm, depending on the sample
composition. It provides adequate analytical
The accuracy for determination of
Selenium using the short-lived nuclide 77mSe
was evaluated by analyzing a number of
certified reference materials with different
levels of Se (IAEA 436, NIST 1566b, NIST
1577 and NIST 1577b). The agreement
between measured and certified values was
generally very good with u-score < 1.64, as
shown in Table IV.
1.8
NIST 1566b
detection limit, ppm
1.6
IAEA 436
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0
5
10
15
20
25
30
35
40
45
Fig. 4. The detection limits of Se in IEAE 436 and 1566b.
89
A NEW RAPID NEUTRON ACTIVATION ANALYSIS SYSTEM AT
For the determination of the Selenium
by the instrumental neutron activation analysis,
the long-lived nuclide 75Se or the short-lived
nuclide 77mSe can be used [7]. With the short-
lived nuclide, not only completion times are a
distinct advantage but analytical sensitivities
are also improved. The data for procedures are
listed in Table V.
Table IV. The results of concentration analysis for Se in biological reference materials.
Reference
material
Certificated
value
(in ppm)
IAEA 463
k-zero method
The relative method
This work
(in ppm)
u-score
This work
(in ppm)
u-score
4.63  0.48
4.55  0.50
0.12
4.19  0.46
0.66
NIST 1566b
2.06  0.15
2.48  0.57
0.71
2.18  0.42
0.27
NIST 1577
1.10  0.10
1.24  0.31
0.43
1.17  0.22
0.29
NIST 1577b
0.73  0.06
0.70  0.11
0.24
0.80  0.17
0.39
Table V. Parameters were used for INAA analysis of Selenium in biological sample by
using 77mSe and 75Se isotopes.
75
Radionuclide
Se
77m
Se
Half-life
120 d
17.4 s
Activation
20 h at 3.5 x
1012 (n/cm2/s)
1525 s at 4.2
x 1012
(n/cm2/s)
Decay time
20 d
20 s
Counting time
23 h
25 s
Detection limit
1.4 ppm
0.6 ppm
Sample: IAEA
436
4.63 ± 0.48
ppm
4.63 ± 0.48
ppm
The results
4.35 ± 1.1 ppm
4.19 ± 0.46
ppm
IV. CONCLUSION
A fast pneumatic sample transfer system
for analyzing of extremely short-lived nuclides
by neutron activation analysis has been
installed and operated at Dalat nuclear research
reactor. In this study, time parameters of the
system were calibrated, thereby reducing
irradiation time to seconds with precision.
Neutron spectra parameters of the thermal
90
column and channel No.13-2 were also
determined in order to establish analytical
procedures using the k0-NAA method. The
system was applied to determine the
concentration of Se in the biological sample by
using the short-lived nuclide 77mSe. The results
obtained through this research have opened a
new possibility on using INAA technique for
measurement of extremely short-lived nuclides
at Nuclear Research Institute.
HO VAN DOANH et al.
[4] L.S. McDowell, et al., Determination of
Selenium in individual food items using the
short-lived nuclide 77mSe, Journal of
Radioanalytical and Nuclear Chemistry, Vol.
110, No. 2, p. 519 (1987).
ACKNOWLEDGEMENTS
This project was carried out under the
nuclear research and development program of
the Ministry of Science and Technology,
Vietnam.
[5] A. D. Becker, Characterization and use of the
new NIST rapid pneumatic tube irradiation
facility, Journal of Radioanalytical Chemistry,
Vol. 233, No. 1-2, p. 155 (1998).
REFERENCES
[1]
S.S. Ismail, A new automated sample transfer
system for instrumental neutron activation
analysis, journal of Automated Methods and
Management in Chemistry, Vol. 2010, (2010).
[6] R. Gwozdz, F. Grass, J. Dorner, Fluorine
analysis of standard materials by short-time
activation analysis using 20F, Journal of
Radioanalytical and Nuclear Chemistry, Vol.
169, No.1, p. 57 (1993).
[2] Yong-Sam Chung, et al., Characteristics of a
new pneumatic transfer system for a neutron
activation analysis at the HANARO research
reactor, Nuclear Engineering and Technology,
Vol. 41, No. 6, p. 813 (2009)
[7] D. Behni, et al., Combination of Neutron
Activation Analysis, Tracer Techniques, and
Biochemical Methods in the Investigation of
Selenium
Metabolism,
Journal
of
Radioanalytical and Nuclear Chemistry, p.439
(1989).
[3] U.M. El-Ghawi, et al., Determination of
Selenium in Libyan Food Items Using
Pseudocyclic Instrumental Neutron Activation
Analysis, Biological Trace Element Research,
Vol. 107, p. 61 (2004).
91

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