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The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic PN scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a keff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%. Keywords: TULIP, STREAM, Fast reactor, MOC core calculation

The authors developed the modified power method (MPM) in previous publications to obtain multiple eigenmodes of an eigenvalue problem at the same time by employing a generalized eigenvalue problem (GEP) of the form of WX = VXK. Special attention has been paid to the Monte Carlo (MC) implementation of the MPM because it always suffers from the inherent statistical noises. In this paper, a preconditioning method for the GEP has been developed for the MC MPM, so that the performance is more stable and robust to the MC statistical noises. This preconditioning method is crucial for MC solving of problems with degenerated eigenmodes, which requires the accumulation of a so-called transfer matrix and the division of the system space into multiple sub-regions, the number of sub-regions being greater than the number of eigenmodes to be solved. The preconditioning method solves the issues arising from the mismatch between the number of sub-regions and the target number of eigenmodes to be calculated. The numerical results for a model cube problem and BEAVRS whole core neutron transport eigenvalue problem successfully demonstrate the validity of the preconditioning method and the extended applicability of the MC MPM for practical problems.

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors. Keywords: Verification and validation, PWR core, Two-step approach, STREAM, RAST-K 2.0

As mental health issues increase worldwide, nature-based experiences are being recognized as alternative treatments for improving health and well-being. Increasing urbanization precludes many people from accessing green spaces owing to time or physical limitations. Therefore, opportunities to connect with nature through virtual technology is being encouraged. We conducted a systematic review of studies on the psychological effects of experiencing nature using virtual technology. We searched the academic databases PubMed, Web of Science, and Scopus for relevant studies and assessed their quality using Cochrane’s RoB 2 and ROBINS-I tools. Twenty-one studies were included and the psychological outcomes were negatively synthesized by the intervention characteristics (duration, observation position of the landscape, interaction, environment description, and sensory type). Psychological outcomes were classified into emotional recovery, cognitive recovery, stress reduction, and other indicators. Emotional recovery was most consistently presented, and virtual natural contact alleviated negative emotions more than it elicited positive emotions. Additionally, virtual nature interventions lasting more than 10 min showed more consistent effects than those of less than 10 min. Moreover, an open field of view led to significant emotional recovery and an in-forest view led to significant cognitive recovery. Despite some limitations, our findings will contribute to the development of virtual forest experiences to improve human well-being.

China Experimental Fast Reactor (CEFR) is a small size sodium-cooled fast reactor (SFR) with a high neutron leakage core fueled by uranium oxide. The CEFR core with 20 MW(e) power reached its first criticality in July 2010, and several start-up tests were conducted from 2010 to 2011. The China Institute of Atomic Energy (CIAE) proposed to release some of the neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activities through the coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”, launched in 2018. This benchmark aims to perform validation and verification of the physical models and the neutronics simulation codes by comparing calculation results against collected experimental data. The six physics start-up tests considered for this CRP include evaluation of the criticality, control rod worth, void reactivity, temperature coefficient, swap reactivity, and foil irradiation. Twenty-nine participating research organizations are performing independent blind calculations during the first phase of the project. As a part of this coordinated research, IAEA performed neutronics calculations using Monte Carlo code SERPENT. Two kinds of 3D core models, homogenous and heterogeneous, were calculated using SERPENT, with ENDF/B-VII.0 continuous energy library. Preliminary results with a reasonably good estimation of criticality, as well as theoretically sound results of other five test cases, are available. The paper will discuss the core modelling assumptions, challenges and key findings of modelling a dense SFR core, preliminary results of the first phase of the CRP, heterogeneity impact analysis between homogenous core models and heterogeneous core models and future work to be performed as a part of this four-year project.

The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models and various engineering features. It is a three-dimensional multi-group nodal diffusion code developed for the steady and transient states using microscopic cross-sections generated by the STREAM code for 37 isotopes. A depletion chain containing 22 actinides and 15 fission products and burnable absorbers was solved using the Chebyshev rational approximation method. A simplified one-dimensional single-channel thermal-hydraulic calculation was performed with various values for the thermal conductivity. Advanced features such as burnup adaptation and CRUD modeling capabilities are implemented for the multi-cycle analysis of commercial reactor power plants. The performance of RAST-K v2 has been validated with the measured data of PWRs operating in Korea. Furthermore, RAST-K v2 has been coupled with a sub-channel code (CTF), fuel performance code (FRAPCON), and water chemistry code for multiphysics analyses. In this paper, the calculation models and engineering features implemented in RAST-K v2 are described, and then the application status of RAST-K v2 is presented.

The status of nuclear power plants’ conditions must be checked to avoid initial events which may eventually lead to accidents. Axial power and axial offset (AO) are key parameters which are usually used to state 3-dimensional core power peaking, in the form of a practical parameter. Group method of data handling (GMDH) has a wide range of applications. Here, a GMDH is used and modified as an efficient method to predict the axial power and AO of the reactor core as well as fuel assemblies. In this paper, axial power offset of the whole reactor core is reconstructed by using in-core detectors. By using the developed GMDH algorithm, the optimum relationship between the independent in-core detector signals and the dependent variables, the core axial power and AO, is determined. Two separate sets of big data are prepared and analyzed. The first set includes power at each of 24 axial nodes at different core states. The second set is similar to the first set except as it also contains fuel assembly data.

This study presents the first application of the advanced Rossi-alpha method (theoretically introduced by Kong et al., 2014) on the reactivity measurements in a research reactor: detector count signals at the Kyoto University Critical Assembly (KUCA) facility. The detector signals in the KUCA A-type core are analyzed by three subcriticality measurement methods: (1) Feynman-alpha (F-α) method, (2) Rossi-alpha (R-α) method, and (3) advanced Rossi-alpha (advanced R-α) method. Four cases are analyzed for two different subcritical states of the core and two different neutron source locations. Two different negative reactivity ρ values are obtained by the measurements of control rod worth and regarded as the reference reactivity values, comparing the results by the four methods. The F-α shows reactivity errors ranging between 7.1 and 7.3% due to its use of variance-to-mean ratios of detector count signals, which are not very sensitive to neutron background noise. However, the fitting uncertainties associated to the F-α results are large, ranging between 5.4 and 12.8% at one standard deviation. The R-α shows small fitting uncertainties ranging between 2.8 and 3.8%, although reactivity errors are in the range of 3.5–26.5% due to the neutron background noise. Finally, the advanced R-α that explicitly models the neutron background noise contrary to the previous methods shows the reactivity errors in the range of 1.0–11.8%, and provides the lowest uncertainties of the measured ρ in the range of 0.4–0.9%. In conclusion, among the four methods applied to the reactivity measurements at KUCA, the advanced R-α reveals the best accuracy with the lowest uncertainties.

Summary This paper presents a new design for a small modular sodium-cooled fast reactor core with an optimized lifetime and reactivity swing through the analysis of various breed-and-burn strategies and its neutronic analyses in terms of active core movements, isotopic mass balance, kinetic parameters, and inherent safety. The new core design aims at a power level of 260 MW with a long lifetime of 30 years without refueling and a reactivity swing smaller than 1000 pcm. Starting from five initial candidate cores with various breed-and-burn strategies, an optimum core was selected from a combination of the two candidates that shows a proper breeding behavior with the optimized uranium enrichment in the low-enriched uranium region and the optimized size of the blanket region. The depletion analysis of the new core provides various reactor design parameters such as the core multiplication factor, breeding ratio, heavy metal mass change, power distribution, and summary of neutron balance. In addition, the perturbation analysis provides the reactor kinetic parameters and reactivity feedback coefficients for the inherent safety analysis of the core. The integral reactivity parameters of the quasi-static reactivity balance analysis demonstrate that the new core is inherently safe in cases of unprotected loss of flow, unprotected loss of heat sink, and unprotected transient over power. Copyright © 2016 John Wiley & Sons, Ltd.

Summary This paper presents a preliminary design of a Small Modular Pressurized Water Reactor (SMPWR) aimed at boron-free operation which can reduce the size of a nuclear power plant (NPP) and the amount of liquid radioactive waste, and also reduce the corrosion issues caused by boric acid in the coolant. The design parameter limits, such as reactivity swing, axial offset (AO), 3D pin peaking factor (Fq), and the required shutdown margin, have been established for the boron-free SMPWR. Furthermore, a new ring type of burnable absorber (R-BA) with Zr–167Er is adopted as a burnable absorber (BA) and HfB2 is adopted as a control rod material to satisfy those design limits without liquid boron. Optimal fuel assemblies (FAs) and loading patterns have been searched through a sensitivity study, and the excess reactivity control capacity and design requirements have been demonstrated to be satisfactory. Copyright © 2016 John Wiley & Sons, Ltd.

The International Atomic Energy Agency and the China Institute of Atomic Energy proposed a coordinated research project (CRP) to establish a benchmark, based on the China Experimental Fast Reactor (CEFR) start-up tests which include fuel loading and criticality, measurements of control rod worth and reactivity coefficients, and foil activation measurements. As a participant in the CRP, the computational reactor physics and experiment laboratory of the Ulsan National Institute of Science and Technology has been studying and analyzing the neutronic performance of the CEFR. The verification and validation are performed for the simulations in the CEFR analysis. A verification process of the methodology was conducted to compare the predicted results with those obtained by a well-known MCNP6.1 Monte Carlo code. The results obtained by the MCS code were compared against the experimental data. The difference in various wire models at the critical state are evaluated in this study.

A design study of ultra-long cycle fast reactor with a rated power of 1000 MWe (UCFR-1000) was performed to flatten its radial power distribution throughout the cycle depletion of the core. It is expected that the flattening of the radial power distribution lowers the power peak factor at the center and consequently the maximum peak temperature of the fuel and clad. Thorium fuel has lower reproduction factor than uranium; thus, thorium-loaded fuel zone has relatively lower power. A lot of test cases were introduced that have thorium-loaded fuel in the inner fuel region to lower the center power peak. Each test case had a different zoning strategy; different fuel forms were loaded for both the upper blanket region and the lower low enrichment uranium (LEU) region. The final evaluation models were designed by loading the thorium-loaded fuel only in the inner blanket for the 1-zone case, in both blanket and LEU regions for the 2-zone case, and in an additional zone for the other cases. It was demonstrated that the fuel zoning has a significant impact on the power flattening, and a gradual fuel zoning can achieve a flattened radial power distribution, and can consequently reduce the maximum power peak.

This paper presents a new high performance burnable absorber (BA) design for advanced Pressurized Water Reactors (PWRs) aiming for a long-cycle operation with a low soluble boron concentration. The new BA consists of a UO 2 – 157 Gd 2 O 3 rod covered with a thin layer of Zr 167 Er 2 . A key feature of this new BA is that enriched isotopes, 157 Gd and 167 Er, are used as absorber materials. Since the high absorption cross section of 157 Gd can reduce the mass fraction of Gd 2 O 3 in UO 2 –Gd 2 O 3 , the thermal margin of fuel rods will increase with higher heat conductivity. Also, the 157 Gd transmutes into 158 Gd by neutron absorption and therefore the residual penalty at the end of cycle (EOC) will decrease. Since 167 Er has a resonance near the thermal neutron energy region, the moderator temperature coefficient (MTC) will become more negative and the control rod worth will increase. These advantages of the new BA are demonstrated with three verification cases: a 17 × 17 Westinghouse (WH) type fuel assembly, a 16 × 16 Combustion Engineering (CE) type fuel assembly, and an OPR-1000 equilibrium core.

Summary This paper presents detailed analyses of a pressurized water reactor with a new reflector design using zirconium metal. The optimization of the reflector design has been performed using a two-dimensional fuel assembly reflector model. The three-dimensional core calculation results with the optimized reflector were compared against those with the existing water reflector and iron reflector. The high scattering cross section of zirconium enhances neutron reflections from the reflector to the core, increasing the peripheral assembly powers. From the analysis based on the equilibrium core, it was noted that the cycle length can be extended, and the pin peaks can be decreased when using zirconium reflector. The analysis has been performed for the optimized power reactor 1000 core with combustion engineering type fuel assemblies using the CASMO-4E/SIMULATE-3 (Studsvik Scandpower, Inc., Waltham, MA, USA) code system and SERPENT (VTT Technical Research Centre of Finland, Vuorimiehentie 3, 02150 Espoo, Finland) code, with ENDF/B-VI data. Copyright © 2015 John Wiley & Sons, Ltd.

This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research formore » a complete reactor model with the other reactor components.« less

As one of participants of recent coordinated research project proposed by International Atomic Energy Agency and China Institute of Atomic Energy, the computational reactor physics and experiment laboratory of Ulsan National Institute of Science and Technology, Republic of Korea, is currently focused on a study involving the neutronic simulation of the China Experimental Fast Reactor (CEFR) start-up tests. The CEFR start-up tests include obtaining measurements of the criticality, control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. The fast reactor neutronic analysis code SARAX is selected for performing such simulations for additional validation. The obtained numerical results are compared with measurement data. In addition, the differences in the expected results of the heterogeneous and homogeneous models are quantified in this study. According to the simulation, the calculated results agreed well with the measured values in the case of the majority of the neutronic characteristics.

This paper proposes a new control rod material for a boron-free small modular pressurized water reactor (SMPWR) to compensate the loss of the soluble boron. To design control rod for SMPWR is difficult due to the large amount of reactivity control necessity and the multiple roles of control rod, i.e., the excess reactivity control, the coarse/fine power control, and the axial shape control and shutdown. This paper presents a study of various control rod materials: B4C, Ag-In-Cd, and HfB2. The rod worth and the excess reactivity control capability will be assessed for the SMPWR using Studsvik’s reactor core design code system, CASMO-4E/SIMULATE-3.

To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolant material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.

A soluble boron-free small modular pressurized water reactor (SMPWR) uses burnable absorbers (BA) instead of soluble boron to reduce excess reactivity. As a consequence, the fuel cycle length can be shortened by the residual penalty of BA. This paper performs cutback sensitivity tests to extend the cycle length. The influence of the height of the cutback, of the 235U enrichment rate, and of the BA material on the power peaking factor (Fq), the axial offset (AO) and the fuel cycle length is analyzed with the reactor core design system, CASMO-4E/SIMULATE-3 code system.