http://www.hindawi.com The latest articles from Hindawi Publishing Corporation © 2013 , Hindawi Publishing Corporation . All rights reserved. Thu, 16 May 2013 11:10:51 +0000 http://www.hindawi.com/journals/stni/2013/945858/ Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing. Michael F. Simpson Copyright © 2013 Michael F. Simpson. All rights reserved. Wed, 08 May 2013 15:22:40 +0000 http://www.hindawi.com/journals/stni/2013/851987/ Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs) leaks assumed) to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS). For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization. Andrej Prošek and Leon Cizelj Copyright © 2013 Andrej Prošek and Leon Cizelj. All rights reserved. Wed, 08 May 2013 10:30:11 +0000 http://www.hindawi.com/journals/stni/2013/127676/ Silicon carbide fiber/silicon carbide matrix (SiC-SiC) composites exhibit remarkable material properties, including high temperature strength and stability under irradiation. These qualities have made SiC-SiC composites extremely desirable for use in advanced nuclear reactor concepts, where higher operating temperatures and longer lives require performance improvements over conventional metal alloys. However, fabrication efficiency advances need to be achieved. SiC composites are typically produced using chemical vapor infiltration (CVI), where gas phase precursors flow into the fiber preform and react to form a solid SiC matrix. Forced flow CVI utilizes a pressure gradient to more effectively transport reactants into the composite, reducing fabrication time. The fabrication parameters must be well understood to ensure that the resulting composite has a high density and good performance. To help optimize this process, a computer model was developed. This model simulates the transport of the SiC precursors, the deposition of SiC matrix on the fiber surfaces, and the effects of byproducts on the process. Critical process parameters, such as the temperature and reactant concentration, were simulated to identify infiltration conditions which maximize composite density while minimizing the fabrication time. Christian P. Deck, H. E. Khalifa, B. Sammuli, and C. A. Back Copyright © 2013 Christian P. Deck et al. All rights reserved. Tue, 09 Apr 2013 10:37:27 +0000 http://www.hindawi.com/journals/stni/2013/618707/ In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2), which is a compact gas-cooled fast reactor (GFR). The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization. Hangbok Choi, Robert W. Schleicher, and Puja Gupta Copyright © 2013 Hangbok Choi et al. All rights reserved. Sun, 31 Mar 2013 15:24:21 +0000 http://www.hindawi.com/journals/stni/2013/787196/ This paper aims at performing a human reliability analysis using THERP (Technique for Human Error Prediction) and ATHEANA (Technique for Human Error Analysis) to develop a qualitative and quantitative analysis of the latent operator error in leaving EFW (emergency feed-water) valves closed in the TMI-2 accident. The accident analysis has revealed a series of unsafe actions that resulted in permanent loss of the unit. The integration between THERP and ATHEANA is developed in a way such as to allow a better understanding of the influence of operational context on human errors. This integration provides also, as a result, an intermediate method with the following features: (1) it allows the analysis of the action arising from the plant operational context upon the operator (as in ATHEANA), (2) it determines, as a consequence from the prior analysis, the aspects that most influence the context, and (3) it allows the change of these aspects into factors that adjust human error probabilities (as in THERP). This integration provides a more realistic and comprehensive modeling of accident sequences by considering preaccidental and postaccidental contexts, which, in turn, can contribute to more realistic PSA (Probabilistic Safety Assessment) evaluations and decision making. Renato A. Fonseca, Antonio Carlos M. Alvim, Paulo Fernando F. Frutuoso e Melo, and Marco Antonio B. Alvarenga Copyright © 2013 Renato A. Fonseca et al. All rights reserved. Sat, 30 Mar 2013 15:56:40 +0000 http://www.hindawi.com/journals/stni/2013/412349/ This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities) was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC) for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering). This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future. S. K. Kim, W. I. Ko, and Yoon Hee Lee Copyright © 2013 S. K. Kim et al. All rights reserved. Wed, 27 Mar 2013 14:30:38 +0000 http://www.hindawi.com/journals/stni/2013/380284/ In the framework of the OECD/NEA project on Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation, and Safety Analysis of LWRs, several approaches and codes are being used to deal with the exercises proposed in Phase I, “Specifications and Support Data for Neutronics Cases.” At UPM, our research group treats these exercises with sensitivity calculations and the “sandwich formula” to propagate cross-section uncertainties. Two different codes are employed to calculate the sensitivity coefficients of to cross sections in criticality calculations: MCNPX-2.7e and SCALE-6.1. The former uses the Differential Operator Technique and the latter uses the Adjoint-Weighted Technique. In this paper, the main results for exercise I-2 “Lattice Physics” are presented for the criticality calculations of PWR. These criticality calculations are done for a TMI fuel assembly at four different states: HZP-Unrodded, HZP-Rodded, HFP-Unrodded, and HFP-Rodded. The results of the two different codes above are presented and compared. The comparison proves a good agreement between SCALE-6.1 and MCNPX-2.7e in uncertainty that comes from the sensitivity coefficients calculated by both codes. Differences are found when the sensitivity profiles are analysed, but they do not lead to differences in the uncertainty. C. J. Díez, J. J. Herrero, O. Cabellos, and J. S. Martínez Copyright © 2013 C. J. Díez et al. All rights reserved. Sun, 24 Mar 2013 08:52:07 +0000 http://www.hindawi.com/journals/stni/2013/154706/ The in situ breeding and burning reactor (ISBBR), which makes use of the outstanding breeding capability of metallic pellet and the excellent irradiation-resistant performance of SiCf/SiC ceramic composites cladding, can approach the design purpose of ultralong cycle and ultrahigh burnup and maintain stable radial power distribution during the cycle life without refueling and shuffling. Since the characteristics of the fuel pellet and cladding are different from the traditional fuel rod of ceramic pellet and metallic cladding, the multiphysics behaviors in ISBBR are also quite different. A computer code, named TANG, to model the specific multiphysics behaviors in ISBBR has been developed. The primary calculation results provided by TANG demonstrate that ISBBR has an excellent comprehensive performance of GEN-IV and a great development potential. Shengyi Si Copyright © 2013 Shengyi Si. All rights reserved. Sat, 23 Mar 2013 15:40:40 +0000 http://www.hindawi.com/journals/stni/2013/946173/ The Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC) benchmark based on the Nuclear Power Engineering Corporation (NUPEC) pressurized water reactor (PWR) subchannel and bundle tests (PSBTs). The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA) of OECD and the Japan Nuclear Energy Safety Organization (JNES), Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs) codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB), under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases. M. Avramova, A. Rubin, and H. Utsuno Copyright © 2013 M. Avramova et al. All rights reserved. Thu, 21 Mar 2013 11:21:43 +0000 http://www.hindawi.com/journals/stni/2013/172518/ The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting SFRs is described. The drawback on the feedback effects due to the introduction of minor actinides into SFR fuel is analyzed. The possibility of compensation of the effect of the minor actinides on the feedback effects by the use of fine distributed moderating material is demonstrated. The consequences of the introduction of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation are analyzed, and the positive effects on the transmutation efficiency are shown. Finally, the possible increase of the Americium content to improve the transmutation efficiency is discussed, the limit value of Americium is determined, and the possibilities given by an increase of the hydrogen content are analyzed. Bruno Merk Copyright © 2013 Bruno Merk. All rights reserved. Wed, 20 Mar 2013 11:41:53 +0000 http://www.hindawi.com/journals/stni/2013/961360/ Nuclear material accounting (NMA) is the only safeguards system whose benefits are routinely quantified. Process monitoring (PM) is another safeguards system that is increasingly used, and one challenge is how to quantify its benefit. This paper considers PM in the role of enabling frequent NMA, which is referred to as near-real-time accounting (NRTA). We quantify NRTA benefits using period-driven and data-driven testing. Period-driven testing makes a decision to alarm or not at fixed periods. Data-driven testing decides as the data arrives whether to alarm or continue testing. The difference between period-driven and datad-riven viewpoints is illustrated by using one-year and two-year periods. For both one-year and two-year periods, period-driven NMA using once-per-year cumulative material unaccounted for (CUMUF) testing is compared to more frequent Shewhart and joint sequential cusum testing using either MUF or standardized, independently transformed MUF (SITMUF) data. We show that the data-driven viewpoint is appropriate for NRTA and that it can be used to compare safeguards effectiveness. In addition to providing period-driven and data-driven viewpoints, new features include assessing the impact of uncertainty in the estimated covariance matrix of the MUF sequence and the impact of both random and systematic measurement errors. T. Burr and M. S. Hamada Copyright © 2013 T. Burr and M. S. Hamada. All rights reserved. Mon, 18 Mar 2013 15:06:51 +0000 http://www.hindawi.com/journals/stni/2013/610598/ The subject of the present paper is the uncertainty and sensitivity studies for steady state BFBT results including pressure drop and void fraction measurements. The investigations are performed with TRACE (version 5.0 patch 2), thermal hydraulic modeling, and SUSA and DAKOTA; both tools are for the evaluation of uncertainties and sensitivities. For this purpose, the NUPEC BFBT experimental data base is used. The advantage of applying two different uncertainty and sensitivity tools in combination with TRACE is that the user effect can be excluded. Since in both cases the TRACE model of the BFBT bundle is identical the differences in the results are related to the capabilities of the uncertainty and sensitivity tools. The reference results with TRACE show that the code is very well able to represent both single- and two-phase flows even though it is a 1D coarse mesh system code. For selected cases, an uncertainty study was performed. Even though a reduced number of uncertain parameters are considered in the DAKOTA investigation, compared to the one with SUSA, similar results are obtained. The results indicate also that even small parameter variations can yield to rather large variations of the selected output parameters. Wadim Jaeger, Victor Hugo Sánchez Espinoza, Francisco Javier Montero Mayorga, and Cesar Queral Copyright © 2013 Wadim Jaeger et al. All rights reserved. Wed, 13 Mar 2013 15:41:27 +0000 http://www.hindawi.com/journals/stni/2013/437409/ The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ) due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3) of fuel material compositions were analyzed. A remarkable increase in uncertainty in was observed for the case of MOX fuel. The increase in uncertainty of in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′), contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel. C. Arenas, R. Bratton, F. Reventos, and K. Ivanov Copyright © 2013 C. Arenas et al. All rights reserved. Mon, 11 Mar 2013 13:15:09 +0000 http://www.hindawi.com/journals/stni/2013/343492/ Pyroprocessing technology has been actively developed at Korea Atomic Energy Research Institute (KAERI) to meet the necessity of addressing spent fuel management issue. This technology has advantages over aqueous process such as less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, and compact equipments. This paper describes the pyroprocessing technology development at KAERI from head-end process to waste treatment. The unit process with various scales has been tested to produce the design data associated with scale-up. Pyroprocess integrated inactive demonstration facility (PRIDE) was constructed at KAERI and it began test operation in 2012. The purpose of PRIDE is to test the process regarding unit process performance, remote operation of equipments, integration of unit processes, scale-up of process, process monitoring, argon environment system operation, and safeguards-related activities. The test of PRIDE will be promising for further pyroprocessing technology development. Hansoo Lee, Geun-IL Park, Jae-Won Lee, Kweon-Ho Kang, Jin-Mok Hur, Jeong-Guk Kim, Seungwoo Paek, In-Tae Kim, and IL-Je Cho Copyright © 2013 Hansoo Lee et al. All rights reserved. Wed, 06 Mar 2013 09:46:22 +0000 http://www.hindawi.com/journals/stni/2013/879634/ An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety. Toshio Wakabayashi Copyright © 2013 Toshio Wakabayashi. All rights reserved. Sun, 03 Mar 2013 15:14:11 +0000 http://www.hindawi.com/journals/stni/2013/352757/ The Canadian Supercritical-Water-Cooled Reactor (SCWR) is a vertical pressure tube reactor cooled with supercritical light water and moderated with heavy water. For normal operation, the local conditions of the coolant (density and temperature) and fuel (temperature) vary substantially along the channel. This means that to simulate adequately the behavior of the core under operating conditions or for anticipated accident scenario, expensive 3D transport calculations for a complete fuel channel are required. Here, we propose a simulation strategy that takes into account axial variations of the local conditions and avoids 3D transport calculations. This strategy consists in replacing the 3D simulation by a series of isolated 2D calculations followed by a single 1D simulation. It is shown that this strategy is efficient because the axial coupling along the fuel channel is relatively weak. In addition, the neutronic properties of a channel with axial reflector can be modeled using a simplified 3D transport calculation. G. Harrisson and G. Marleau Copyright © 2013 G. Harrisson and G. Marleau. All rights reserved. Thu, 28 Feb 2013 13:49:59 +0000 http://www.hindawi.com/journals/stni/2013/565246/ In the present paper, an uncertainty and sensitivity study is performed for transient void fraction and pressure drop measurements. Two transients have been selected from the NUPEC BFBT database. The first one is a turbine trip without bypass and the second one is a trip of a recirculation pump. TRACE (version 5.0 patch 2) is used for the thermohydraulic study and SUSA and DAKOTA are used for the quantification of the model uncertainties and the evaluation of the sensitivities. As uncertain parameters geometrical values, hydraulic diameter, and wall roughness are considered while mass flow rate, power, pressure, and inlet subcooling (inlet temperature) are chosen as boundary and input conditions. Since these parameters change with time, it is expected that the importance of them on pressure drop and void fraction will change, too. The results show that the pressure drop is mostly sensitive to geometrical variations like the hydraulic diameter and the form loss coefficient of the spacer grid. For low void fractions, the parameter of the highest importance is the inlet temperature/subcooling while at higher void fraction the power is also of importance. Wadim Jaeger, Victor Hugo Sánchez Espinoza, Francisco Javier Montero Mayorga, and Cesar Queral Copyright © 2013 Wadim Jaeger et al. All rights reserved. Tue, 26 Feb 2013 10:04:30 +0000 http://www.hindawi.com/journals/stni/2013/725687/ The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes. M. Avramova, A. Velazquez-Lozada, and A. Rubin Copyright © 2013 M. Avramova et al. All rights reserved. Tue, 26 Feb 2013 08:17:24 +0000 http://www.hindawi.com/journals/stni/2013/590684/ Solution monitoring (SM) has been used in a nuclear reprocessing plant as an additional measure to provide assurance that the plant is operated as declared. The inline volume and density monitoring equipment with dip tubes is important for safety and safeguards purposes and is a typical example of safeguards by design (SBD). Recently safety, safeguards, and security by design (3SBD) are proposed to promote an efficient and effective generation of nuclear energy. In 3SBD, proliferation risk assessment has the potential to consider likelihood of the incidence and proliferation risk in safeguards. In this study, risk assessment methodologies for safeguards and security are discussed and several mathematical methods are presented to investigate risk notion applied to intentional acts of facility misuse in an uncertainty environment. Proliferation risk analysis with the Markov model, deterrence effect with the game model, and SBD with fuzzy optimization are shown in feasibility studies to investigate the potential application of the risk and uncertainty analyses in safeguards. It is demonstrated that the SM is an effective measurement system using risk-informed and cost-effective SBD, even though there are inherent difficulties related to the possibility of operator’s falsification. Mitsutoshi Suzuki and Norichika Terao Copyright © 2013 Mitsutoshi Suzuki and Norichika Terao. All rights reserved. Wed, 13 Feb 2013 16:04:54 +0000 http://www.hindawi.com/journals/stni/2013/867561/ Computer simulations have been carried out to investigate the possibility of extending operating cycle length in the Pressurised Water Reactor Ringhals 3 by the use of thorium-plutonium oxide fuel. The calculations have been carried out using tools and methods that are normally employed for reload design and safety evaluation in Ringhals 3. The 3-batch reload scheme and the power level have been kept unchanged, and a normal uranium oxide fuel assembly designed for a 12-month operating cycle in this reactor is used as a reference. The use of plutonium as the fissile component reduces the worth of control rods and soluble boron, which makes it necessary to modify the control systems. The delayed neutron fraction is low compared with the reference, but simulations and qualitative assessments of relevant transients indicate that the reactor could still be operated safely. Differences in reactivity coefficients are mainly beneficial for the outcome of transient simulations for the thorium based fuel. A 50% extension of the current 12-month operating cycle length should be possible with thorium-plutonium mixed oxide fuel, given an upgrade of the control systems. More detailed simulations have to be carried out for some transients in order to confirm the qualitative reasoning presented. Klara Insulander Björk, Cheuk Wah Lau, Henrik Nylén, and Urban Sandberg Copyright © 2013 Klara Insulander Björk et al. All rights reserved. Sun, 10 Feb 2013 10:47:49 +0000 http://www.hindawi.com/journals/stni/2013/437854/ The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in order to perform uncertainty analysis on and 2-group homogenized macroscopic cross-sections. The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is a 15 15 pressurized water reactor (PWR) fuel assembly segment with poison and at full power conditions. A statistical methodology is employed for the uncertainty assessment, where cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB library format are considered as normal random variables. Two libraries were created for such purposes, one based on JENDL-4 data and the other one based on the recently released ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries needed to be computed for the different isotopic reactions by means of ERRORJ. The uncertainty assessment performed on and macroscopic cross-sections, that is based on JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much larger at the thermal and resonant regions than, for instance, the covariance matrix based on ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different uncertainty assessments. Augusto Hernández-Solís, Christophe Demazière, and Christian Ekberg Copyright © 2013 Augusto Hernández-Solís et al. All rights reserved. Thu, 07 Feb 2013 15:29:01 +0000 http://www.hindawi.com/journals/stni/2013/549793/ Capabilities for uncertainty quantification (UQ) with respect to nuclear data have been developed at PSI in the recent years and applied to the UAM benchmark. The guiding principle for the PSI UQ development has been to implement nonintrusive “black box” UQ techniques in state-of-the-art, production-quality codes used already for routine analyses. Two complimentary UQ techniques have been developed thus far: (i) direct perturbation (DP) and (ii) stochastic sampling (SS). The DP technique is, first and foremost, a robust and versatile sensitivity coefficient calculation, applicable to all types of input and output. Using standard uncertainty propagation, the sensitivity coefficients are folded with variance/covariance matrices (VCMs) leading to a local first-order UQ method. The complementary SS technique samples uncertain inputs according to their joint probability distributions and provides a global, all-order UQ method. This paper describes both DP and SS implemented in the lattice physics code CASMO-5MX (a special PSI-modified version of CASMO-5M) and a preliminary SS technique implemented in MCNPX, routinely used in criticality safety and fluence analyses. Results are presented for the UAM benchmark exercises I-1 (cell) and I-2 (assembly). W. Wieselquist, T. Zhu, A. Vasiliev, and H. Ferroukhi Copyright © 2013 W. Wieselquist et al. All rights reserved. Thu, 07 Feb 2013 14:23:23 +0000 http://www.hindawi.com/journals/stni/2013/573697/ The OECD UAM Benchmark was launched in 2005 with the objective of determining the uncertainty in the simulation of Light Water Reactors (LWRs) system calculations at all the stages of the coupled reactor physics—thermal hydraulics modeling. Within the framework of the “Neutronics Phase” of the Benchmark the solutions of some selected test cases at the cell physics and lattice physics levels are presented. The SCALE 6.1 code package has been used for the neutronics modeling of the selected exercises. Sensitivity and Uncertainty analysis (S/U) based on the generalized perturbation theory has been performed in order to assess the uncertainty of the computation of some selected reactor integral parameters due to the uncertainty in the basic nuclear data. As a general trend, it has been found that the main sources of uncertainty are the 238U (n,) and the 239Pu nubar for the UOX- and the MOX-fuelled test cases, respectively. Moreover, the reference solutions for the test cases obtained using Monte Carlo methodologies together with a comparison between deterministic and stochastic solutions are presented. Luigi Mercatali, Kostadin Ivanov, and Victor Hugo Sanchez Copyright © 2013 Luigi Mercatali et al. All rights reserved. Sun, 03 Feb 2013 10:43:51 +0000 http://www.hindawi.com/journals/stni/2013/508485/ Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement. Maria Avramova and Diana Cuervo Copyright © 2013 Maria Avramova and Diana Cuervo. All rights reserved. Mon, 28 Jan 2013 09:33:03 +0000 http://www.hindawi.com/journals/stni/2013/790206/ The aim of this work is to present the Exercise I-1b “pin-cell burn-up benchmark” proposed in the framework of OECD LWR UAM. Its objective is to address the uncertainty due to the basic nuclear data as well as the impact of processing the nuclear and covariance data in a pin-cell depletion calculation. Four different sensitivity/uncertainty propagation methodologies participate in this benchmark (GRS, NRG, UPM, and SNU&KAERI). The paper describes the main features of the UPM model (hybrid method) compared with other methodologies. The requested output provided by UPM is presented, and it is discussed regarding the results of other methodologies. O. Cabellos Copyright © 2013 O. Cabellos. All rights reserved. Tue, 22 Jan 2013 16:20:08 +0000 http://www.hindawi.com/journals/stni/2013/293792/ Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content) and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed. A. Schwenk-Ferrero Copyright © 2013 A. Schwenk-Ferrero. All rights reserved. Tue, 22 Jan 2013 15:16:24 +0000 http://www.hindawi.com/journals/stni/2013/687494/ We investigate the present capabilities of CFD for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. Very similar modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant nondimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, liquid temperature, and bubble size. Robust predictive capabilities of the modeling require that it is validated for a wide range of parameters. It is known that a careful calibration of correlations used in the wall boiling model is necessary to obtain agreement with the measured data. We here consider tests under a variety of conditions concerning liquid subcooling, flow rate, and heat flux. It is investigated to which extent a set of calibrated model parameters suffices to cover at least a certain parameter range. Roland Rzehak and Eckhard Krepper Copyright © 2013 Roland Rzehak and Eckhard Krepper. All rights reserved. Mon, 14 Jan 2013 12:09:07 +0000 http://www.hindawi.com/journals/stni/2013/426356/ The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC) transient PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets. Gerhard Strydom Copyright © 2013 Gerhard Strydom. All rights reserved. Mon, 31 Dec 2012 15:18:00 +0000 http://www.hindawi.com/journals/stni/2012/698019/ The mixed oxide (MOX) fuel is one of the most important fuels for the advanced reactors in the future. It is flexible to be applied either in the thermal reactor like pressurized water reactor (PWR) or in the fast reactor (FR). This paper compares the two approaches from the view of fuel cost. Two features are involved. (1) The cost of electricity (COE) is investigated based on the simulation of realistic operation of a practical PWR power plant and a typical fast breeder reactor design. (2) A new economic analysis model is established, considering the discount rate and the revenue of the reprocessed plutonium besides the traditional costs in the processes of fuel cycle. The sensitivity of COE to the changing parameters is also analyzed. The results show that, in the closed fuel cycle, the fuel cost of applying MOX fuels in the FBR is about 25% lower than that in the PWR at the current operating and fuel cycle level. Youqi Zheng, Hongchun Wu, Liangzhi Cao, and Shizhuang Jia Copyright © 2012 Youqi Zheng et al. All rights reserved. Mon, 24 Dec 2012 16:45:01 +0000 http://www.hindawi.com/journals/stni/2012/215469/ Yan Yang, Jin Huang, and Xing Chen Copyright © 2012 Yan Yang et al. All rights reserved.