Development of REFT Code System for the Analysis of Highly Heterogeneous Research Reactors

2014 ◽  
Author(s):  
Tengfei Zhang ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Yunzhao Li
Keyword(s):  
Least Square Method ◽  
Least Square ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
User Friendly ◽  
Control Rods

As an effort to enhance the accuracy in simulating the operations of research reactors, a fuel management code system REFT was developed. Because of the possible complex assembly geometry and the core configuration of research reactors, the code system employed HELIOS in the lattice calculation to describe arbitrary 2D geometry, and used the 3D triangular nodal SN method transport solver, DNTR, to model unstructured geometry in the core analysis. Flux reconstruction with the least square method and micro depletion model for specific isotopes were incorporated in the code. At the same time, to make it more user friendly, a graphical user interface was also developed for REFT. In the analysis of the research reactors, the calculations involving the control rod movement are encountered frequently. The modeling of the control rods differential worth behavior is important in that the movement of the control rod may introduce variations on the reactivity. To handle the problem two effective ways of alleviating the control rod cusping effect are recently proposed, based on the established code system. The methodologies along with their application and validation will be discussed.

A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System

2018 ◽  
Author(s):  
Heng Yu ◽  
Guan-bo Wang ◽  
Da-zhi Qian ◽  
Yu-chuan Guo ◽  
Bo Hu
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
Core Damage ◽  
The World ◽  
Safety Features ◽  
Control Rods ◽  
Pool Type

An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

scholarly journals DESIGN OF NEUTRONIC PARAMETERS OF MTR REACTOR USING WIMSD-5B/BATAN-FUEL CODES

2020 ◽  
Vol 5 (3) ◽  
pp. 239-248
Author(s):  
Tukiran Surbakti ◽  
Surian Pinem ◽  
Lily Suparlina
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Cycle Length ◽  
Thermal Neutron Flux ◽  
Diffusion Method ◽  
Irradiation Position ◽  
Research Reactors ◽  
The Core ◽  
Neutronic Parameters ◽  
Control Rods

BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.

scholarly journals Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Energies ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 7377
Author(s):  
Michał Górkiewicz ◽  
Jerzy Cetnar
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Reactor Performance ◽  
Fission Rate ◽  
Control Rod ◽  
Structured Design ◽  
The Core ◽  
Stable Core ◽  
Control Rods ◽  
Burnup Calculation

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

scholarly journals Calculation of excore detector weighting functions for a sodium-cooled TRU burner mockup using MCNP5

2015 ◽  
Vol 5 (2) ◽  
pp. 15-25
Author(s):  
Viet Ha Pham Nhu ◽  
Min Jae Lee ◽  
Sunghwan Yun ◽  
Sang Ji Kim
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Detector Response ◽  
Weighting Functions ◽  
Control Rod ◽  
The Core ◽  
Power Regulation ◽  
Fuel Assemblies ◽  
Control Rods ◽  
Horizontal Layers

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A.

scholarly journals SMALL MODULAR PWR DESIGN FOR TRU RECYCLING WITH McCARD-MASTER TWO-STEP PROCEDURE

2021 ◽  
Vol 247 ◽  
pp. 01003
Author(s):  
Dae Hee Hwang ◽  
Ser Gi Hong
Keyword(s):  
Mass Flow ◽  
Cross Sections ◽  
Treatment Method ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Group Constant ◽  
Step Procedure ◽  
Neutronic Characteristics

In our previous study, a small modular PWR core was designed for TRU (Transuranics) recycling with multi-recycling scheme with a typical two-step procedure using DeCART2D/MASTER code system in which the lattice analysis for producing homogenized group constant was performed by DeCART2D while whole core analysis was conducted by MASTER code. However, the neutron spectrum hardening of the LWR core loaded with TRU requires validating the multi-group cross section library and resonance self-shielding treatment method in lattice calculation. In this study, a new procedure using McCARD/MASTER was used to analyze the SMR core, in which the lattice calculation was performed by a Monte Carlo code called McCARD with a continuous energy library to generate homogenized two-group assembly cross sections. The SMR core analysis was performed to show neutronic characteristics and TRU mass flow in the SMR core with TRU multi-recycling. The result shows that the analyses on the neutronic characteristics and TRU mass flow using the McCARD/MASTER code system showed good agreement with the previous ones using the DeCART2D/MASTER code system. The neutronic characteristics of each cycle of the core satisfied the typical limit of a commercial PWR core and the SMR core consumes effectively TRU with net TRU consumption rates of 8.46~14.33 %.

Experience of Xenon Oscillation During an Initial Physics Test of UCN Unit 5 in Korea

2010 ◽  
Author(s):  
Yong Rae Kim ◽  
Tae Young Choi ◽  
Sun Ho Shin ◽  
Ki Bong Seong
Keyword(s):  
Steady State ◽  
Annual Cycle ◽  
Stability Index ◽  
Control Rod ◽  
The Core ◽  
Axial Stability ◽  
Xenon Oscillation ◽  
Burnable Absorber ◽  
Control Rods ◽  
Initial Cycle

Initial core of Ulchin Nuclear Unit 3 (UCN3), which is one of earlier OPR1000 model, was 4 batches and designed as annual cycle after second cycle. The utility requested that UCN Unit 5 (UCN5), which is another of OPR1000 model, had capability of a longer cycle operation from second cycle. Therefore, KNF modified the number of batches from 4 to 3 for OPR1000 initial core, as well as, the number of burnable absorber, and the cutback length of the absorber. However, due to these changes, Xenon oscillation was slightly increased at 100% power during the physics test of UCN5, while that oscillation at 100% power in UCN3 had been gone down without any control rod motion. The xenon oscillation direction is related to axial stability index. The index of UCN3 increased from a slightly negative at BOC to positive at EOC, the index of UCN5 was positive even at BOC, which meant that the core does not go to be stable without the control rod motion. The core of UCN5 became the steady state by the insertion of control rods into the core. To meet the physics test condition, the oscillation was controlled by control rods immediately. After the happening, KNF optimized the cutback length of burnable absorber rods and applied to APR1400, which will keep being stable in xenon oscillation during physics test at the initial cycle.

Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Yuki Honda ◽  
Nozomu Fujimoto ◽  
Hiroaki Sawahata ◽  
Shoji Takada ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Flux Distribution ◽  
Cell Model ◽  
Control Rod ◽  
The Core ◽  
Macroscopic Cross Section ◽  
Conventional Cell ◽  
Test Reactor ◽  
Control Rods

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross-section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

Alternative Absorber Materials for Control Rods in ALFRED

2021 ◽  
Author(s):  
Hui Guo ◽  
Xin Jin ◽  
Kuaiyuan Feng ◽  
Hanyang Gu
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Operating Life ◽  
Insertion Depth ◽  
Control Rod ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reference Design ◽  
Core Characteristics ◽  
Control Rods

Abstract The next-generation reactors require improved safety performance and longer cycle length, which initiate the research on alternative absorber materials. In this context, potential absorber materials including borides (B4C, HfB2, and ZrB2), rare earth oxides (Eu2O3, Gd2O3, Sm2O3, and Dy2TiO5), metals/alloys (Hf and AIC), and metal hydride (HfHx) were compared in a large sodium fast reactor. The design of control rods for Generation-IV fast reactors strongly depends on the core characteristics. In this paper, some alternative absorbers are assessed in a lead fast reactor ALFRED using depletion capability in the Monte-Carlo particle transport code OpenMC. Results show that the ALFRED reference control rod design with B4C largely satisfies the shutdown and operation requirements. 60% 10B enriched HfB2 and HfH1.18 can replace the operation part of the reference design. In the future, the safe operating life of B4C and HfB2 should be assessed taking into account the irradiation-induced swelling, temperature margin, and gas release. HfH1.18 has a limited and local influence on the core power distribution. Eu2O3 has little loss on the absorption ability after 5 cycle irradiation. This oxide absorber satisfies the shutdown function even with only half control rod insertion, while its critical insertion depth at beginning of the cycle should be increased to realize reactivity compensation function.

Econometric Analysis of Brazilian Scientific Production and Comparison with BRICS

2018 ◽  
Vol 23 (1) ◽  
pp. 25-46 ◽  
Author(s):  
L.A. Hayne ◽  
A.T.S. Wyse
Keyword(s):  
Large Scale ◽  
Least Square Method ◽  
Scientific Production ◽  
Least Square ◽  
Production Model ◽  
Capitalist Production ◽  
The Core ◽  
Starting Point ◽  
Key Factor ◽  
Near Future

This article proposes a study that placed Brazilian scientific production at the core of the debate, and Brazil as a member of BRICS framework emphasises its importance. It starts from the argument that scientific production is a key factor considering that it is through knowledge that societies evolve. This issue becomes even more relevant when pointing to the capitalist production model that transforms knowledge into wealth. That was the starting point for leading an econometric study to formulate a model to explain Brazilian scientific production from 1994 to 2014. The least square method was used to estimate the parameters of a production function to identify the variables that most influenced, statistically, the behaviour of Brazilian scientific production. The main variables that strongly explained the number of published articles, proxy of the scientific production, are number of post-graduate programmes, number of masters and doctors, Brazilian population and expenditure on research and development (R&D). The present study showed that postgraduate programmes are the ones that most influenced the behaviour of Brazilian published articles. Therefore, it is necessary to equip Brazilian postgraduate system to create the basis of a structure, which, in the near future, will produce knowledge in large scale and with high competitiveness.

Sign in / Sign up

Export Citation Format

Share Document