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In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ± 4%, and the pin-wise results are within ± 4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

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

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.

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.