The journal of Nuclear Science and Technology - Volume 4, Number 1, March 2014
ISSN 1810-5408
Nuclear Science
and Technology
Volume 4, Number 1, March 2014
Special issue for the 30th anniversary of the inauguration
of the restoration and expansion of Dalat nuclear research reactor
(20 March 1984 - 20 March 2014)
Published by
VIETNAM ATOMIC ENERGY SOCIETY
NUCLEAR SCIENCE AND TECHNOLOGY
Volume 4, Number 1, March 2014
Editorial Board
Editor-in-chief
Tran Huu Phat (VINATOM)
Executive Editors
Vuong Huu Tan (VARANS)
Le Van Hong (VINATOM)
Cao Đình Thanh (VINATOM)
Editors
Ngo Quang Huy (HUI)
Phan Sy An (HMU)
Le Hong Khiem (IOP)
Cao Chi (VINATOM)
Dao Tien Khoa (VINATOM)
Do Ngoc Lien (VINATOM)
Dang Duc Nhan (VINATOM)
Nguyen Mong Sinh (VINATOM)
Le Xuan Tham (DOST of Lamdong)
Tran Duc Thiep (IOP)
Nguyen Nhi Dien (VINATOM)
Bui Dieu (NCI)
Le Ngoc Ha (Tran Hung Dao Hospital)
Duong Ngoc Hai (IOM)
Le Huy Ham (VAAS)
Nguyen Quoc Hien (VINATOM)
Bui Hoc (HUMG)
Le Ba Thuan (VINATOM)
Huynh Van Trung (VINATOM)
Dang Thanh Luong (VARANS)
Nguyen Thi Kim Dung (VINATOM)
Pham Dinh Khang (VINATOM)
Nguyen Phuc (VINATOM)
Nguyen Tuan Khai (VINATOM)
Hoang Anh Tuan (VAEA)
Foreign Editors
Hideki Namba (JAEA, Japan)
Philippe Quentin (CENBG, CNRS, France)
Yang (KAERI, Korea)
Pierre Darriulat (INST)
Myung Chul Lee (WFNM)
IU.E.Ponionzkevich (DUBNA , Russia)
Kato Yasuyoshi (TIT, Japan)
Managing Secretary
Nguyen Trong Trang (VINATOM)
Science Secretary
Hoang Sy Than (VINATOM)
....................................................................................................................................................................................................................
Copyright: ©2008 by the Vietnam Atomic Energy Society (VAES), Vietnam Atomic Energy Institute
(VINATOM).
Pusblished by
Vietnam Atomic Energy Society, 59 Ly Thuong Kiet, Hanoi, Vietnam
Tel: 84-4-39420463 Fax: 84-4-39424133
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Email: [email protected]
....................................................................................................................................................................................................................
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 studyin 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
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 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).
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
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
©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.
II. BRIEF REACTOR DESCRIPTION
AND IT’S OPERATION
The results of operation and utilization
Main specifications of the DNRR are
shown in Table I.
of the DNRR are presented in this paper and
Table I. Specifications of the DNRR.
Reactor type
Swimming pool TRIGA Mark II, modified to Russian
type of IVV-9
Nominal thermal power
Coolant and moderator
Core cooling mechanism
Reflector
500 kW, steady state
Light water
Natural convection
Beryllium and graphite
Fuel types
WWR-M2, dispersed UO2-Al with 19.75% enrichment,
aluminium cladding
Number of control rods
Materials of control rods
7 (2 safety rods, 4 shim rods, 1 regulating rod)
B4C for safety and shim rods, stainless steel for
automatic regulating rod
Neutron measuring channels
Vertical irradiation channels
6 combined in 3 housings with 1 CFC and 1 CIC each
4 (neutron trap, 1 wet channel, 2 dry channels) and 40
holes at the rotary rack
Horizontal beam-ports
Thermal column
4 (1 tangential - No #3 and 3 radial - No #1, #2, #4)
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
Sh
Upper
cylindrical
shell
~ 6840 mm
Extracting
well
RgR
Concrete
shielding
Spent fuel
storage tank
Thermal
column door
Sh
Sh
A
Graphite
Door plug
Core
SR
(ex bulk-shielding
experimental tank)
Fig. 1. Vertical section view of the DNRR reactor.
Fig. 2. Cross-section view of the core with 92
fuel assemblies.
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.
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
RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR
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.
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.
III. MAIN RESULTS OF REACTOR
UTILIZATION
A. Radioisotopes and radiopharmaceuticals
production
In order to support the application of
Research
on
radioisotope
production
and
53
99mTc, 113mIn and Sm radioisotopes in clinical
radiopharmaceutical
serving
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.
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.
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
4
NGUYEN NHI DIEN et al.
Other applications of radioisotopes
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.
produced at the DNRR are radiotracer
technique in sediment studies, oil exploitation,
chemical industry, biology, agriculture and
46
hydrology. Some main products are Sc, 192Ir,
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.
198Au, 131I, 140La, etc. In addition, some small
sources of 192Ir and 60Co with low radioactivity
have also been produced for industry
applications.
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
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).
prospecting,
agriculture,
biology,
environmental studies, etc.
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
RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR
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
Column
door
Beam port
# 2
Beam port
# 3
Thermal
Column
Stainless steel
Aluminum
Graphite reflector
Pool tank wall
Core
Bellow s
assembly
Beam port
#
1
Beam port # 4
No. 4: Nuclear data
measurement
Thermalizing column (closed)
This port is closed
Concrete
shielding
Spent fuel storage 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,
151Eu, 153Eu on the 55 keV and 144 keV
neutron beams;
6
NGUYEN NHI DIEN et al.
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 mono-
crystals 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
RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR
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
(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.
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.
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 2005-
2007 to replace neutron measurement and
signal processing parts of the existing C&I
system by the digital system named ASUZ-
14R. 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 24th, 2011 until January 13th,
2012 according to a planned program that
was approved by Vietnam Atomic Energy
Institute (VINATOM). At 15:35 on
November, 30th, 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, 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 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
NGUYEN NHI DIEN et al.
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 in-
core 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.
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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
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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
Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 10-25
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
(Received 5 March 2014, accepted 10 March 2014)
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
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.
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.
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.
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
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.
©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
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 and Thermal Hydraulics
Calculation
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
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.
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, Np-
237, 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 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
for PLTEMP code. A fuel assembly was
modelled as three concentric cylindrical tubes.
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].
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] thermal-
hydraulics code for plate and concentric-tube
geometries with capability of calculating
natural circulation flow was used for thermal-
hydraulics 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
Safety system settings
550 kW (110% FP)
20s
60 cm
40 m3/h
Secondary coolant flow rate
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
LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG
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 anticipated
occurrences:
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;
2)
Fig. 2b. DNRR model for PLTEMP
(1-fuel assembly cross-section; 2-FA model
for PLTEMP; 3-reactor coolant system model).
(3) Fuel cladding integrity shall be
assured.
- For accident conditions:
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.
(1) Core covering shall be maintained;
(2) Core shall not be remarkably
damaged;
(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
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
DESIGN ANALYSES FOR FULL CORE CONVERSION OF …
Table II. Core parameters used for safety analysis.
Parameters
Values
500
Power, kW
Coolant inlet temperature, oC
Peaking factor (shim rods at 300 mm)
- Axial peaking factor
- Radial peaking factor
- Local peaking factor
Reactor kinetics
32
1.363
1.376
1.411
8.92510-5
7.55110-3
- Prompt neutron life, s
- Delayed neutron fraction (1$)
Temperature reactivity coefficients
- Moderator, %/K; (293-400oK)
- Fuel, %/oC;
- 1.26410-2
(293-400oK)
- 1.8610-3
- 1.9210-3
- 1.5610-3
(400-500oK)
(500-600oK)
- Void, %/% of void
(0-5%)
(5-10%)
(10-20%)
-0.2432
-0.2731
-0.3097
Reactivity control
- Shutdown worth, % (2 safety rods)
- Maximum withdrawal speed of one shim
rod, mm/s
3.7
3.4
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
20
0.16
9.1
6.7
Working range
- Drop time of control rods, s
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
3He 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
3
6
3
buildup of He, Li and H concentrations
14
LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG
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
16.34
-3.89 0.4
-4.65 0.0038
2 Beryllium Rods at
the end 2002
12.66
-6.28 0.4
-7.19 0.0039
The working core characteristics
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.
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
Table IV. Calculation results of working core compared with current mixed core.
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
6.63
3.79
-2.92
-6.62
-4.56
1.431
1.398
1.434
Control Rods In
Thermal Neutron Flux at Neutron Trap Center (n/cm2)
Control Rods Out
2.22E+13
2.14E+13
2.22E+13
3.15E+12
Control Rods In
Fast Neutron Flux at Neutron Trap Center (n/cm2)
Control Rods Out
1.95E+12
1.92E+12
Control Rods In
15
DESIGN ANALYSES FOR FULL CORE CONVERSION OF …
Table V. Control Rods worths (%k/k).
Core1
Fresh
MCNP
error
Core1
Burnt
MCNP
error
Control Rods
Shim rod 1
2.5896
2.6100
2.7784
2.4687
0.4363
2.1955
2.2356
0.000091
0.000111
0.000118
0.000122
0.000126
0.000106
0.000119
2.3539
2.4033
2.5381
2.2604
0.3629
2.3084
2.3579
0.000091
0.000124
0.000122
0.000117
0.000119
0.000115
0.000105
Shim rod 2
Shim rod 3
Shim rod 4
Regulating rod
Safety rod 1
Safety rod 2
Table VI. Neutron flux performance.
Thermal, <0.625eV
(n/cm2.s)
Epithermal,
<0.821MeV
(n/cm2.s)
Fast, <10MeV
(n/cm2.s)
Irradiation positions
Fresh
Maximum 2.07E+13 2.20E+13 6.79E+12 7.12E+12 1.83E+12 1.92E+12
Average 1.45E+13 1.49E+13 6.00E+12 6.04E+12 1.62E+12 1.63E+12
Maximum 9.45E+12 9.86E+12 8.19E+12 8.42E+12 2.98E+12 3.02E+12
Average 7.00E+12 7.12E+12 6.53E+12 6.51E+12 2.46E+12 2.44E+12
Maximum 5.41E+12 5.66E+12 9.63E+12 9.76E+12 4.22E+12 4.26E+12
Average 4.11E+12 4.18E+12 7.23E+12 7.15E+12 3.19E+12 3.15E+12
Maximum 9.24E+12 9.71E+12 8.02E+12 8.22E+12 2.92E+12 2.99E+12
Burnt
Fresh
Burnt
Fresh
Burnt
Neutron
Trap
Channel
13-2
Channel
7-1
Channel
1-4
Average
6.85E+12 7.01E+12 6.41E+12 6.40E+12 2.42E+12 2.40E+12
Rotary
Specimen
Average
3.55E+12 3.56E+12 7.58E+11 7.56E+11 1.93E+11 1.93E+11
Power Distribution and Power Peaking
Factors
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
Peaking Factor
F.A. Radial Core Radial Axial
Position
(mm)
Total
2.498
2.589
2.615
2.648
2.646
2.605
2.537
0
1.378
1.378
1.375
1.377
1.376
1.378
1.378
1.398
1.399
1.403
1.409
1.411
1.415
1.434
1.296
1.343
1.356
1.365
1.363
1.336
1.284
150
200
250
300
350
600
16
LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG
0.980
0.957
1.165
1.126
1.164
1.129
0.973
0.947
0.849 0.858 0.979 1.107 1.006 0.885
0.851 0.858 0.964 1.085 1.001 0.895
0.913
0.901
0.963
0.958
1.004
1.008
0.835
0.889
0.877 0.762
0.875 0.817
1.021
1.023
0.775 0.908
0.833 0.913
0.825
0.876
SR
1.082
1.124
1.406
1.368
ShR
ShR 1.090 1.410
0.790
0.841
0.787 0.921
0.843 0.915
0.917
0.906
1.126 1.370
0.962 0.872 0.906 1.281
0.939 0.873 0.953 1.296
1.259 0.882 0.842
1.283 0.937 0.850
0.919
0.903
1.138
1.090
0.983
0.975
0.959
0.933
0.998
0.974
1.313
1.289
1.353
1.321
0.865
0.855
0.930
0.929
1.139
1.122
0.911
0.900
RgR
1.056 1.220
1.027 1.192
1.307
1.284
1.167 1.018 0.996 1.364
1.117 0.991 0.987 1.332
0.918 0.843 0.911
0.919 0.837 0.887
0.868
0.919
1.252
1.273
0.845
0.835
1.020 0.903
0.992 0.904
0.929 1.296
0.975 1.312
0.808
0.818
0.996 0.830
0.977 0.881
1.079
1.114
ShR
ShR
1.421 1.408
1.381 1.368
1.106
1.145
0.755 0.841
0.808 0.836
0.968 0.810 0.860 1.038
0.959 0.863 0.910 1.038
1.005 0.816 0.745 0.831
1.006 0.866 0.803 0.836
SR
1.016 0.918 1.031 1.124 0.986 0.858 0.843 0.895
0.994 0.917 1.019 1.097 0.970 0.857 0.846 0.889
1.180 0.985
1.137 0.960
1.198
1.156
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
Moderator Temperature Reactivity Coefficient (%/oC)
293 oK to 400 oK
-0.01317
0.00005
Fuel Temperature (Doppler) Reactivity Coefficient
(%/oC)
293 oK to 400 oK
400 oK to 500 oK
500 oK to 600 oK
-0.00192
-0.00182
-0.00154
0.00005
0.00003
0.00002
Moderator Density (Void) Reactivity Coefficient (%/%
of void)
0 to 5 %
-0.2514
-0.2784
-0.3255
0.0011
0.0012
0.0006
5% to 10 %
10 % to 20 %
17
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