(Luận án tiến sĩ) nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân đà lạt

MINISTRY OF EDUCATION
AND TRAINING

MINISTRY OF SCIENCE AND
TECHNOLOGY

VIETNAM ATOMIC ENERGY INSTITUTE
-----------------------------

NGUYỄN KIÊN CƯỜNG

Study on neutronics, thermal hydraulics and core
management for safe operation and effective utilization of the
Dalat Nuclear Research Reactor

A thesis submitted in fulfillment of the requirements for the
degree of Doctor of Philosophy

HÀ NỘI – 2023
MINISTRY OF EDUCATION

AND TRAINING


TECHNOLOGY

MINISTRY OF SCIENCE
AND

VIETNAM ATOMIC ENERGY INSTITUTE
-----------------------------

NGUYỄN KIÊN CƯỜNG

Study on neutronics, thermal hydraulics and core
management for safe operation and effective utilization of the
Dalat Nuclear Research Reactor

DOCTORAL THESIS
Subject: Atomic and Nuclear Physics
Code number: 9-44-01-06

Supervisor Assoc. Prof. PhD. NGUYỄN NHỊ ĐIỀN
:

Hà Nội – 2023
1


DECLARATION OF AUTHORSHIP

I, Nguyễn Kiên Cường, declare that this thesis titled, ―Study on
neutronics, thermal hydraulics and core management for safe operation and
effective utilization of the Dalat Nuclear Research Reactor‖ is my own work,
conducted under the supervision of Assoc. Prof. PhD. Nguyễn Nhị Điền, and
has not been published in any other works or articles. The results were coauthored with other authors after having permission to use for the thesis.
Author

Nguyễn Kiên Cường

2

ACKNOWLEDGEMENTS
I would like to express my deep gratitude to my supervisor Assoc. Prof. PhD.
Nguyễn Nhị Điền for his valuable assistance and encouragement throughout my
research works.
I would like to thank all my colleagues at the Reactor Center, particularly Mr.
Huỳnh Tôn Nghiêm, Mr. Lê Vĩnh Vinh, Mr. Lương Bá Viên and Mr. Nguyễn Minh
Tuân for their generous support, insightful discussions and contributions to my study.
I appreciate the staff of the Dalat Nuclear Research Institute, Nuclear Training
Center, and Vietnam Atomic Energy Institute for their kind assistance during my
research.
I want to send my appreciation to Assoc. Prof. PhD. Vương Hữu Tấn, Assoc.
Prof. PhD. Phạm Đình Khang, Assoc. Prof. PhD. Nguyễn Xuân Hải, PhD. Trần Chí
Thành, Assoc. Prof. PhD. Nguyễn Tuấn Khải, Assoc. Prof. PhD. Trịnh Anh Đức for
their encouragement for my research works.
I am grateful to Ms. Nguyễn Thúy Hằng for all of her assistance with the
administrative aspects to my research.
Finally, I would like to express my gratitude to my parents, my wife, my son
and my daughter, who always provided me with the confidence and motivation to
complete the thesis.

3


BOC
BWR
CFD
CHF
CITATION

DNBR
DNRI
DNRR
ENDF
FA
FIR
FPD
GA
HANARO
HEU
IAEA
IFA
JEFF
JENDL
LEU
MCNP
MPI
MTR
NPP
ONBR
PVM
PWR
ReR
RERTR
RIA
SA
SaR
ShR
SRAC
TRIGA

VVER
WIMSD
1-D
2-D
3-D

LIST OF ABBREVIATIONS
Beginning of Cycle
Boiling Water Reactor
Computational Fluid Dynamics
Critical Heat Flux
Nuclear Reactor Core Analysis Code
Departure Nucleate Boiling Ratio
Dalat Nuclear Research Institute
Dalat Nuclear Research Reactor
Evaluated Nuclear Data File
Fuel Assembly
Flow Instability Ratio
Full Power Day
Genetic Algorithm
Korean Research Reactor
Highly Enriched Uranium
International Atomic Energy Agency
Instrumental Fuel Assembly
Joint Evaluated Fission and Fusion (European
Evaluated) Nuclear Data Library
Japanese Evaluated Nuclear Data Library
Low Enriched Uranium
Monte Carlo N-Particle Computer Code
Message Passing Interface

Material Testing Reactor
Nuclear Power Plant
Onset Nucleate Boiling Ratio
Parallel Virtual Machine
Pressurized Water Reactor
Automatic Regulating Rod
Reduced Enrichment of Research and Test Reactor
Reactivity Insertion Accident
Simulated Annealing
Safety Rod
Shim Rod
Standard Reactor Analysis Code
Training, Research, Isotope Production of General
Atomics
Water-Water Energetic Reactor
Winfrith Improved Multigroup Scheme
Unidimensional
Bidimensional
Tridimensional
4


CONTENT
ACKNOWLEDGEMENTS........................................................................................ 3
LIST OF ABBREVIATIONS..................................................................................... 4
CONTENT ................................................................................................................... 5
LIST OF TABLES .................................................................................................... 10
INTRODUCTION..................................................................................................... 11
CHAPTER 1. GENERAL OVERWIEW................................................................ 21
1.1. Motivation of the thesis .................................................................................. 21

1.2. General introduction about the DNRR ........................................................ 23
1.2.1. History, structure and reactor core arrangement ...................................... 23
1.2.2. Fuel of VVR-M2 HEU and LEU ................................................................. 31
1.2.3. Neutronics characteristics of the DNRR .................................................... 34
1.2.4. Thermal hydraulics characteristics of the DNRR ...................................... 36
1.3. The development of computer codes for reactor calculation in the world 38
1.4. The research situation about research reactor in Vietnam ........................ 41
1.5. Reactor kinetics in three dimensions ............................................................ 44
1.6. Burn-up calculation for core and fuel management.................................... 46
CHAPTER 2. CALCULATION MODELS FOR THE DALAT NUCLEAR
RESEARCH REACTOR USING LEU FUEL....................................................... 48
2.1. Neutronics calculation models ....................................................................... 48
2.1.1. Deterministic code ...................................................................................... 48
2.1.2.1. Lattice cell model .................................................................................. 48
2.1.2.2. Whole core model ................................................................................. 54
2.1.2. Calculation model for computer codes using Monte Carlo method .......... 57
2.2. Thermal hydraulics calculation for the DNRR ............................................ 59
2.3. Reactor kinetics application for the DNRR ................................................. 62
2.3.1. Preparation group constants for the PARCS code ..................................... 62
2.3.2. Calculation model for the DNRR using the PARCS code ......................... 63
2.4. Burn-up calculation for the DNRR ............................................................... 64
2.4.1. Development MCDL computer code ......................................................... 64
2.4.2. Application of MCDL for burn-up and refueling calculation for LEU core
.............................................................................................................................. 73
2.5. Summary of Chapter 2 ................................................................................... 74
CHAPTER 3. RESULTS AND DISSCUSSIONS .................................................. 76
3.1. Neutronics and thermal hydraulics for LEU core ....................................... 76
5

3.1.1. Neutronics calculation results....................................................................76
3.1.1.1. Neutronics characteristics of the HEU and LEU VVR-M2 fuel types. . 76

3.1.1.2. Criticality, reactivity, control rod worths.............................................80
3.1.1.3. Excess reactivity, control rod worth, beryllium rods...........................85
3.1.1.4. Neutron flux distribution......................................................................88
3.1.1.5. Power peaking factor............................................................................92
3.1.1.6. Feedback reactivity temperature coefficients and void factor.............95
3.1.1.7. Kinetics parameters..............................................................................96
3.1.2. Thermal hydraulics calculation results.......................................................97
3.1.2.1. Validation of the PLTEMP code...........................................................97
3.1.2.2. Steady state of PLTEMP code without hot channel factors...............101
3.1.2.3. Steady state of PLTEMP code with hot channel factors....................103
3.2. Kinetics calculation results for LEU core...................................................105
3.2.1. Calculation results from the Serpent and PARCS codes at steady state
condition.............................................................................................................105
3.2.2. Calculation and experiment results during increasing power of the DNRR
from 80% to 100%.............................................................................................107
3.2.3. Simulation of the accident when uncontrolled withdrawal of one control
rod at nominal power.........................................................................................110
3.2.4. Simulation of the changing of power when inserting positive reactivity
smaller than 10 cents..........................................................................................113
3.3. Burn-up calculation results..........................................................................115
3.3.1. Validation of the MCDL code..................................................................115
3.3.2. Calculation results of the HEU core.........................................................121
3.3.3. Calculation results of the LEU core.........................................................125
3.4. Summary of Chapter 3.................................................................................133
CONCLUSIONS.....................................................................................................135
NEW CONTRIBUTIONS OF THE THESIS.......................................................139
LIST OF PUBLICATIONS....................................................................................140

REFERENCES........................................................................................................142

6


LIST OF FIGURES
Fig. 1.1. Cross sections of the DNRR in axial and radial directions..........................24
Fig. 1.2. Annual operation time of the DNRR from 1984 to 2011 (the core
configuration using HEU fuel and mixed cores)........................................................30
Fig. 1.3. Annual operation time of the DNRR from 2012 to 2022 (using LEU fuel) 31
Fig. 1.4. Specific dimensions and geometry of the VVR-M2 fuel type ....................... 33
Fig. 2.1. Calculation model for VVR-M2 FA of the DNRR ...................................... 50
Fig. 2.2. Calculation model for group constants of the neutron trap .......................... 52
Fig. 2.3. Calculation model for lattice cells in side the DNRR core .......................... 53
Fig. 2.4. Calculation model for lattice cells outside the DNRR core
.........................
54
Fig. 2.5. Calculation model for the DNRR using the REBUS-PC and CITATION
codes (a- model in REBUS-PC and b- model in CITATION) ................................... 55
Fig. 2.6. Calculation model in axial direction using the CITATION code ................ 56
Fig. 2.7. Model cross section of VVR-M2 FA in MCNP code for normal calculation
and power peaking factor calculation (a- FA; b- FA model in the MCNP code and cexample of detailed power peaking factor calculation inside FA) ............................. 57
Fig. 2.8. Calculation model for a) SaRs or ShRs, b) ReR, c) beryllium rod, d)
aluminum chock rod, e) wet or dry irradiation channels, f) the neutron trap ............. 58
Fig. 2.9. Calculation model for full core of the DNRR using the MCNP code .......... 59
Fig. 2.10. The DNRR model calculation for the PLTEMP/ANL code and LEU core
with 92 FAs ................................................................................................................. 60
Fig. 2.11. Super-cell model in the Serpent code to create group constants of ShR.... 63
Fig. 2.12. Calculation model of the DNRR using the PARCS code .......................... 64
Fig. 2.13. General structure of the MCDL code ......................................................... 69

Fig. 2.14. Burn-up chain model of actinide and fission products isotopes in the
MCDL code ................................................................................................................. 71
Fig. 3.1. Neutron spectrum of HEU and LEU VVR-M2 fuels with 108 neutron
energy groups in average power (89 HEU FAs core and 92 LEU FAs core) ............. 78
Fig. 3.2. Neutron spectrum of LEU VVR-M2 fuel type with different calculation
libraries ........................................................................................................................ 79
Fig. 3.3. Critical core configuration with a) 72 FAs and b) working core with 92 FAs
..................................................................................................................................... 84
Fig. 3.4. Thermal neutron flux distribution in the radial direction (unit ×1012 n/cm2.s)
..................................................................................................................................... 92
Fig. 3.5. Relative power distribution in axial direction and depending on control rod
positions of working core of 92 LEU FAs .................................................................. 94
7


Fig. 3.6. Calculation results of relative power distribution in the working core using
92 LEU FAs (upper value from MCNP and below value from REBUS) ................... 95
Fig. 3.7. Comparison of the measured cladding and coolant temperatures of the HEU
core to validate the PLTEMP/ANL code .................................................................... 98
Fig. 3.8. The HEU VVR-M2 IFA of the DNRR ......................................................... 99
Fig. 3.9. Axial power distribution of the hottest FA (at cell 10-5) calculated by the
MCNP code for 25-cm insertion of 4 ShRs .............................................................. 100
Fig. 3.10. Calculation results at nominal power without errors and uncertainties in the
hottest FA .................................................................................................................. 102
Fig. 3.11. Comparison of the fuel cladding and coolant temperatures at different
reactor power levels................................................................................................... 103
Fig. 3.12. Calculation results at nominal power with systematic errors ................... 104
Fig. 3.13. Calculation results at nominal power with systematic and random errors
................................................................................................................................... 105
Fig. 3.14. Calculation results of the increasing power of the DNRR from 80 to 100%

................................................................................................................................... 108
Fig. 3.15. Experimental data of the changing position of ReR (TD) when increasing
reactor power from 80% to 100% ............................................................................. 109
Fig. 3.16. Position of ReR (TD) and power when increasing reactor power from 80% to
100% ...................................................................................................................... 109
Fig. 3.17. Reactivity and reactor power when increasing reactor power from 80 to
100%.......................................................................................................................... 110

Fig. 3.18. Calculation results the changing position of ReR (TD) when increasing
power from 80 to 100% ............................................................................................. 110
Fig. 3.19. Power and reactivity in the accident of uncontrolled withdrawal of the ShR
number 1 with and without feedback reactivity temperature coefficients of water and
fuel ............................................................................................................................. 112
Fig. 3.20. Reactor power transient of one ShR withdrawal from operating power
100%.......................................................................................................................... 113

Fig. 3.21. Reactor power and reactivity transient of ShR number 1 uncontrolled
withdrawal from operating power 100% ................................................................... 113
Fig. 3.22. Power and reactivity changing when inserting 10 cents reactivity .......... 114
Fig. 3.23. The changing of the ReR (TD) when inserting 10 cents .......................... 115
Fig. 3.24. Experimental data of power (D1) and ReR (TD) position when inserting 10
cents ........................................................................................................................... 115
Fig. 3.25. a) Infinite multiplication factor of HEU fuel depending on burn-up steps
and b) atom density of actinide isotopes at the end of burn-up step (~ 36% burn up of
117
U-235)........................................................................................................................
8


Fig. 3.26. a) Infinite multiplication factor of LEU fuel depending on burn-up steps

and b) atom density of actinide isotopes at the end of burn-up step (~ 29% burn-up of
U-235).......................................................................................................................117
Fig. 3.27. Burn-up (% U-235) distribution of fresh HEU core after 538 FPDs
operation (REBUS-MCNP system code at upper values and MCDL code at lower
values).......................................................................................................................119
Fig. 3.28. Burn-up (%U-235) distribution of fresh LEU core after 600 FPDs
operation (MCNP_REBUS code at upper values and MCDL code at lower values)
120
Fig. 3.29. The changing of heavy isotopes of a) HEU and b) LEU fuels of the DNRR
121
Fig. 3.30. Difference (%) of calculation results and experimental data (using Cs-137
isotope) for 106 burnt HEU FAs...............................................................................123
Fig. 3.31. Atomic number density of Li-6 and He-3 for 240 nodes in calculation
model for LEU core..................................................................................................125
Fig. 3.32. Core configuration of the LEU core with 92 fresh FAs including 12 burnt
LEU FAs (red and blue color numbers are BU% U-235 of LEU FAs slightly burn-up,
black values are identification number of fresh LEU FAs)......................................126
Fig. 3.33. Fuel burn-up distribution of the LEU core with 92 FAs in March, 2021
(upper values are order number, under values are burn-up percent of U-235).........127
Fig. 3.34. The procedure to carrying out refueling 6 FAs of the LEU working core
with 92 FAs (upper values are order number of FAs, lower numbers are BU%).....128
Fig. 3.35. Fuel burn-up distribution of LEU core with 98 FAs (under values)........130
Fig. 3.36. Core configuration of 98 FAs loaded 4 fresh FAs and discharged 4 burnt
FAs having burn-up about 27% (under values)........................................................131
Fig. 3.37. Fuel burn-up distribution of the last cycle using 10 fresh LEU FAs........132

9


LIST OF TABLES

Table 1.1. Material in structure of the DNRR............................................................26
Table 1.2. The parameters of VVR-M2 HEU and LEU fuel.....................................31
Table 2.1. Length and material of LEU FA in axial direction...................................50
Table 3.1. Infinite multiplication factor of the VVR-M2 HEU and LEU fuels.........76
Table 3.2. Calculation results of infinite multiplication factors with different
calculation libraries.....................................................................................................77
Table 3.3. The critical core configurations established during physical start-up.......82
Table 3.4. Multiplication factor of the critical cores using LEU fuel........................84
Table 3.5. The effective control rods worth of the working core using 92 LEU FAs 85

Table 3.6. The effective reactivity of LEU FAs.........................................................86
Table 3.7. Effective reactivity of beryllium rods.......................................................87
Table 3.8. The calculation results and experimental data of relative thermal neutron
flux in radial direction ................................................................................................. 88
Table 3.9. The calculation results and experimental data of relative thermal neutron
flux in axial direction .................................................................................................. 89
Table 3.10. Calculation results and experimental data of thermal neutron flux at the
neutron trap of LEU working core .............................................................................. 90
Table 3.11. Neutron flux distribution at irradiation positions of LEU working core 92
Table 3.12. Power peaking factor of working core using 92 LEU FAs ..................... 92
Table 3.13. The calculation results of feedback reactivity coefficients of LEU core 95
Table 3.14. Calculation results and experimental data of kinetics parameters of LEU
core .............................................................................................................................. 96
Table 3.15. Hot channel factors in thermal hydraulic analysis of the DNRR .......... 101
Table 3.16. Calculation results and experimental data for decay constant .............. 105
Table 3.17. Calculation results and experimental data of delayed neutron fraction. 105
Table 3.18. Calculation results of multiplication factors from the Serpent and PARCS
codes .......................................................................................................................... 106

Table 3.19. Infinite multiplication factors of HEU and LEU FAs depending on

burn-up (% mass of U-235) ....................................................................................... 116
Table 3.20. Operation time and excess reactivity of the HEU cores and mixed-core
................................................................................................................................... 122
Table 3.21. The calculation results and experimental data of negative effective
reactivity of beryllium rods ....................................................................................... 124
10


INTRODUCTION
Research reactors are essential to the implementation of a nation's nuclear
program, and can be used for research and training, material testing, neutron
activation analysis, production of radioisotopes for medicine or industry, and other
purposes. More than 220 research reactors with different power, fuel type, and
neutron energy are currently operating in 53 countries [11]. The structure and power
of all research reactors are quite simple with low power, temperature, and pressure
when compared to nuclear power plants. Under Reduced Enrichment Research and
Test Reactor (RERTR) program [12], almost all operational research or test reactors
were converted from highly enriched uranium (HEU) to low enriched uranium (LEU)
fuel but they still kept the purpose in utilizations and applications. Three main factors
related to the existence of the research reactor are management, operation, and
utilization. From a management point of view, the reactor must be in good condition
and operating staff or managers can know clearly or deeply about the reactor in
practice and parameters as well. In terms of safety operation, the reactor must meet or
exceed the design requirements for safety in physics, thermal hydraulics, and
adequate operation. The reactor also has a design to meet for safe operation even in
abnormal, transients, and accident conditions. Depending on the characteristics of the
reactor, the utilization can be exploited as much as possible. The purposes of
application of the reactor must be explicitly defined before building and operating.
In general, reactor physics can be divided into two problems: statics and
dynamics, along with reactor kinetics and burn-up. In statics calculations, the time

variable in transport or diffusion neutron equations is ignored. Multiplication factor
or reactivity and neutron flux distributions or power distribution are the most
important characteristics derived from static neutronics calculations. For thermal
hydraulics, the safety parameters need to be evaluated including fuel, fuel cladding,
and coolant temperatures, other safety parameters (ONBR or DNBR) under
maximum nominal power, and inlet coolant temperature condition. Reactor kinetics
11


describe the behavior of a reactor based on the insertion or withdrawal of reactivity in
reactor core at time step intervals. Three-dimensional (3-D) reactor kinetics is crucial
and must be considered for any reactor in normal and transient/accident conditions.
During the simulation's subsequent time steps, the power distribution of the hottest
fuel assembly (FA) in radial and axial directions within the reactor core can be
determined by using 3-D reactor kinetics computer code. Fuel burn-up is also a very
important process that directly influences the properties and safety of a reactor.
Changing fuel compositions such as the production of actinide isotopes and fission
products, and reducing the reactor‘s excess reactivity or core lifetime are the two
most significant factors affecting reactor characteristics. The burn-up process of a
reactor occurred according to operation time in day, month, or even yearly timescale.
The burn-up distribution of FAs, excess reactivity, and other parameters are
important for core and fuel management in addition to enhancing safe operation and
effective utilization as well.
In order to conduct experiments on a reactor, it is possible to obtain accurate,
reliable data, but careful preparation in terms of equipment and other resources is
required, whereas the modern calculating method is more straightforward,
economical, and practical due to advanced capabilities of computer nowadays.
During the full core conversion of the Dalat Nuclear Research Reactor (DNRR) [1] at
the end of 2011, numerous experiments were conducted to determine characteristics
of neutronics parameters at the start-up and working LEU core with varying LEU

FAs loadings from 72 to 92 FAs. These experimental data are extremely valuable for
validating computer codes used into design and management of the DNRR. In
addition, the burn-up distribution of 106 burnt HEU FAs was also measured using the
gamma scanning method to estimate the burn-up percentage of U-235 [2], and these
data can be used to validate a self-developed burn-up computer code MCDL (Monte
Carlo Depletion Light water reactor).
Neutronics computer codes for core and fuel management of the DNRR use
both deterministic and Monte Carlo methods for theoretical calculations, especially
after the complete full core conversion in 2012. Popular deterministic codes include
12


cell code, whole core code, and burn-up code; for example, the SRAC2006 (PIJ,
CITATION, COREBN) system code [17, 18] or the WIMS-ANL [19] and the
REBUS-PC [20] codes or the REBUS-MCNP linkage code [21]. For the LEU core,
the MCNP5 [22] or the MCNP6 [23] codes are primarily used for design and
neutronics calculations, as well as burn-up. Because of complicated geometry, the
DNRR is suitable with neutronics codes employing the Monte Carlo method, such as
the MCNP, MVP [24], and Serpent 2 [25] codes. The PLTEMP4.2 [26] thermal
hydraulics code agrees very well with fuel, core models, fuel correlation, and using
natural convection to remove produced heat from the reactor core of the DNRR. All
computer codes used in design of the LEU core were validated by comparing their
obtained results with experimental data or other results from different computer
codes. The REBUS-MCNP linking codes were applied to calculate burn-up and burnup distribution of the HEU or LEU cores. The MCNP code in the system code has a
role in determining neutron flux, reaction rates while the REBUS code calculates
burn-up and updates atomic number density of depletion regions in each FAs. In
reactor kinetics calculation, the PARCS code [27, 28] was used to evaluate the
power, reactivity insertion, and power distribution in radial and axial directions.
The DNRR was reconstructed from the former TRIGA Mark II, which was
built in the early 1960s, operated at 250 kW from 1963 to 1968, and extended

shutdown until March 1975. All TRIGA fuels were unloaded at the end of March
1975 and then shipped back to the United States of America. The reactor
reconstruction project began in 1982 and criticality was reached on November 1 st,
1983. In February 1984, a nominal power of 500 kW was obtained. The initial fresh
working core was loaded with 88 VVR-M2 fuel assemblies that were 36% HEU. The
DNRR was granted permission to partial conversion from HEU core to a mixed
HEU-LEU core beginning in 2006, and the first six LEU FAs (19.75% enrichment)
were installed in September 2007. In 2009, the DNRR established a mixed core with
92 HEU and 12 LEU FAs. In December 2011, the DNRR completed full core
conversion and established the initial critical core loaded 72 LEU FAs with a neutron
trap. The working core was created with 12 slightly burnt LEU FAs and 80 fresh
13


LEU FAs and 12 beryllium rods located around the neutron trap. In September 2019,
two irradiation channels 5-6 and 9-6 were installed in the reactor core by replacing
two beryllium rods to increase the amount of I-131 radioisotope production. In April
2021, the reactor implemented refueling by replacing two beryllium rods with two
new LEU fuel assemblies, and in May 2022, the reactor continued refueling by
replacing two other beryllium rods with two new LEU fuel assemblies. The reactor
was refueled in May 2023 to attain 98 FAs in the working core [7, 8].
Even operating with low power and neutron flux, the DNRR has contributed
significantly to the social-economic development of Vietnam. Various nuclear
engineering and radioisotope applications for medical, agricultural, industrial,
geological, hydrological, and environmental purposes have been implemented on the
DNRR in order to promote economic growth and protect public health. In addition,
the fundamental research on reactor engineering, nuclear physics, and other applied
research conducted on the DNRR has resulted in the development of highly skilled,
experienced staff as well as national expertise in this particular field.
The purpose of the dissertation is to calculate neutronics and steady-state

thermal hydraulics for the DNRR using LEU fuel including characteristics
parameters of the reactor core such as excess reactivity, neutron flux distribution,
power peaking factor, safety parameters, kinetics parameters, and thermal hydraulics
safety parameters. The second objective is to apply the PARCS code having 3-D
reactor kinetics for the DNRR with 92 LEU FAs loaded in mainly reactivity insertion
transients or accidents; the obtained results were compared with experimental data or
calculation results from the RELAP5 code. The third objective is to develop the
MCDL burn-up code by integrating the MCNP code with the burn-up module and
taking into account beryllium poisoning. The MCDL code is capable to estimate fuel
burn-up distribution in 3-D, to support safety analysis, and to calculate in predicting
fuel burn-up. The MCDL code was used for analyzing the burn-up of the HEU and
LEU cores and recommending the fuel loading patterns for using 10 fresh LEU FAs.
The obtained results of 106 HEU FAs burn-up distributions were validated by
comparing them with experimental data using the gamma scanning technique [2].
14


Objectives, research methods and research object of the dissertation
The purpose of the thesis is to achieve the following three objectives in the
field of reactor engineering as following:
(1) Physics and thermal hydraulics calculations appropriate for the operating
conditions of the DNRR using LEU fuel, including specific calculations related to
physics such as excess reactivity, neutron flux distribution, power peaking factor,
safety parameters, kinetic parameters, and thermal hydraulics parameters in steady
state at power level of 500 kW.
(2) Calculating and analyzing the safety of the DNRR's transient and RIA using the
3-D reactor kinetics PARCS code. Determining the parameters of the reactor's power,
reactivity, and thermal hydraulics safety and comparing to experimental data or
calculation results using the RELAP5 code.
(3) Developing a 3-D fuel burn-up calculation code in order to provide a set of

calculation tools for fuel and core management and to support experimental research
pertaining to fuel burn-up. The program has the capacity to compute and anticipate
fuel burn-up for reactors, as well as determine the 3-D burn-up distribution to enable
safety evaluations. The fuel burn-up computation program is used to calculate for the
DNRR's core and fuel management, as well as the core's refueling using LEU fuel.
Research methods: Using deterministic calculation codes such as SRAC2006,
WIMS-ANL, and REBUS as well as Monte Carlo methods such as MCNP, MVP, and
Serpent 2 to develop models and determine detailed physical characteristics of the core
using HEU and LEU fuels. The PLEMP4.2 code is used to determine the thermal
hydraulics safety parameters. The majority of the calculations results will be compared
to the experimental data collected during the DNRR's start-up HEU and LEU cores.
Utilizing the PARCS code, which is capable of computing the 3-D kinetics of several
neutron groups, satisfies the requirements for generating a hexagonal model for the
DNRR. The calculation results were compared with the results acquired from the
RELAP5 code in the computation of the design of the core, as well as the experimental
data obtained at the DNRR's startup. It combines the assessment of beryllium

poisoning for the berrylium blocks and rods in the core of the DNRR with the
15


development of the MCDL fuel burn-up code by coupling the MCNP code and the
fuel burn-up calculation module. Experimental data monitoring the burn-up of 106
burnt HEU FAs using the gamma scanning method were used to validate the MCDL
program's computation results. The calculation of refueling for the DNRR was also
performed and compared with the results of the REBUS-MCNP linkage code and the
experimental data of the excess reactivity through the ontrol rod worths.
Computer codes were used in the thesis to apply for the DNRR:
The MCNP code: Solving integral difference neutron/particles transport
equation form using Monte Carlo method with real geometry and continuous

calculation library of neutron, photon, electron, and other particles. The code has the
ability to apply to reactor calculation in criticality as the original feature. All physical
parameters and characteristics of any research or power reactor can be determined by
the code, especially multiplication factor and neutron flux or power distribution.
Criticality calculation is one of the original features of the code to apply for fission
source or fixed neutron source. After integrating the MCNPX code, a new version of
the MCNP code (MCNP6) can be used for burn-up calculation for core and fuel
management purposes. To determine kinetics parameters, a point kinetics model with
six delayed neutron groups was included in the code.
The SRAC2006 code: It is a system code with integrating many codes inside
for cross section calculation as PIJ code using probability collison method to solve
transport equation for 16 geometry types with 107 neutron energy groups. Especially,
the PIJ code can be used for cell burn-up and reaction rate tallies with PEACO option
for resonance treatment. The ANISN code and TWOTRAN code are also cell code
but available applying to calculate in 1-D and 2-D for whole core models,
respectively. The TUD, CITATION (12 geometries application) are whole core codes
to use for 1-D, 3-D solving diffusion equation by difference method to receive
multiplication factor and neutron flux or power distribution in space. The COREBN
code is applied for burn-up calculation by multi-dimensions diffusion and
interpolation of macro cross section.
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The WIMS-ANL code: It was developed primarily from WIMS-D4 code for
lattice cell calculation. The code is mainly used to generate micro or macro cross
section for the REBUS code in ISOTXS format. The code contains 69 or 172 neutron
energy groups calculation libraries for more than 100 isotopes in order to solve
neutron transport equation with 4 steps. Important output of the code is the change of
atomic number density in one or a few neutron energy groups of heavy and fission
products isotopes (including one aggregated fission product isotope) following burnup. The third-order spline curve is used to fit micro cross sections of heavy and

fission product isotopes in burn-up steps. The code has ability to create multigroup
library for any isotope in burn-up chain with ACE format for the MCNP code.
The REBUS-PC code: The code has the ability for whole core calculation for
triangle and hexagonal lattices using finite difference and nodal methods with two
fundamental analyses for balanced or unbalanced problems in core management.
Four searchings can be conducted inside the code including adjustment burn-up, fuel
enrichment, control absorb density like boron, burn-up time to determine maximum
burn-up, time for a fuel cycle, multiplication factor, and multiplication value at each
burn-up step. When performing burn-up calculations, the main output parameters of
the code are neutron fluxes distribution, power distribution, effective multiplication
factor, reaction rate, and changing of nuclear concentration of fuel regions. In
addition, the code has the ability to manage fuel when carrying out refueling
according to a specified fuel cycle program supplied by user.
The REBUS-MCNP linkage code: The code is built by coupling two computer
codes including the REBUS and MCNP codes together with the WIMS-ANL code for
preparing one energy group cross section of heavy and fission product isotopes (19 to 21
isotopes). Linkage code has two way calculations: 1) using the MCNP code to obtain
reaction rates, one group neutron energy and REBUS for burn-up calculation then update
input for the MCNP code for next burn-up step; 2) using REBUS code to compute
reaction rates, neutron flux and burn-up then update to the MCNP code to calculate
detail for burnt core. Depending on requirements for calculation purposes,
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user can choose suitable scheme for solving specific problems. The most application
of the linkage code is in burn-up calculation for core and fuel management.
The PARCS code: The code has features for solving problems related to
eigenvalue (multiplication factor), xenon transient, residual heat, power distribution,
burn-up, adjoint flux, and especially 3-D kinetics coupling with thermal hydraulics
calculation for heavy water, high-temperature gas cool, and light water reactors. The

code can be applied to nuclear power plants with square or hexagonal lattice
geometries. The thermal hydraulics module of the code is mainly for nuclear power
plants with fuel pin rod type operating at high power, temperature, and pressure
conditions. For transient and steady state calculations, the nodal diffusion methods
with two neutron energy groups are used. The code can also be coupled with
RELAP5 code to create a system code with 3-D kinetics to carry out safety analysis
for research or power reactors.
The PLTEMP code: The code is mainly used for steady-state thermal
hydraulics calculation with determining a few safety parameters to research reactor
using Russian fuel types and others. Numerous heat transfer correlations were added
to the code for many kinds of fuel types and the code can be applied to one channel,
one fuel assembly, or entire core with a maximum of five fuel types. The six hot
channel factors used in this version are also applicable to natural convection. In
general, the code has three main solutions for temperature profiles, which include the
Broyden method and the analytical method for three-layer plates (fuel meat and
claddings), and the analytical method for five-layer plates (fuel meat, gas gaps, and
claddings). The calculation model of the DNRR having extracting well, heat removal
via natural convection, and Russian fuel type VVR-M2 is quite suitable of algorithms
and capability of the code.
The nuclear data library is played a crucial role and affect directly to obtains
results. The data of a library contains characteristics parameters of nuclei such as
cross sections of numerous reactions with different particles, isotope lifetime, decay
chain, energy in gamma decay, fission yield, etc. The nuclear calculation libraries
ENDF/BVII.0 và ENDF/BVII.1, which include approximately 400 isotopes and 20
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isotopes treated in thermal energy with countinuous energy from 10 -11 eV to 20 MeV
of neutron, were mainly used in calculation for Monte Carlo codes or deterministic
code. These libraries were processed from evaluation data libraries from ENDF/B by

using the NJOY2016 code. The S(, β) in the thermal energy range of material, such) in the thermal energy range of material, such
as beryllium, graphite, hydro in light water, etc. was also prepared. The updating of
calculation libraries for the MCNP code in ACE format or the WIMS-D code in
multigroup (69 or 172 groups) are always implemented to be suitable for using in
reactor calculation.
Research object: The thesis focuses on the validation of computer codes (the
MCNP, PLTEMP codes) used for core and fuel management for the DNRR. Detailed
calculation results in neutronics, thermal hydraulics, fuel burn-up, and 3-D kinetics of
the PARCS code for LEU core with 92 FAs of the DNRR were compared to
experimental data or other calculation results for validation purposes. For burn-up
calculation, the MCDL code was applied to determine burn-up distribution in 3-D of
the DNRR. These codes can be used for calculating characteristics parameters of
neutronics and thermal hydraulics before and after refueling. Coupling the PARCS 3D kinetics code and the RELAP5 system code is a good tool for safety analysis in
research or power reactors. This system code for neutronic, thermal hydraulics
calculation, and safety analysis is also suitable for applying to high-power 10 to 15
MW, multi-purpose research reactor using VVR-KN or IR-4M Russian or MTR fuel
types in the future.
Scientific significance and practical contribution
The thesis has the following scientific significance and practical contributions:
-

Neutronics and thermal hydraulics computer codes were used to the core and

fuel management for the DNRR in order to ensure safe operation and efficient
utilization. In particular, the evaluation and determination of specific physical, thermal
hydraulics parameters of the core following long-term operations, as well as the
preparation for the refueling. When comparing obtained results with experimental data,
neutronics, and thermal hydraulics calculation codes were also validated. These codes
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