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AbstractIntegral benchmarks for criticality safety and nuclear data validation require expensive uncertainty quantification studies. In general, uncertainty quantification techniques ignore correlations between experiments and shared components. Experiments, such as the Thermal/Epithermal eXperiments (TEX) campaigns, consist of many shared components, such as the Jemima highly enriched uranium (HEU) fuel plates, which create a strong correlation in their uncertainties. While these correlations are known to exist, they are often not estimated because of the complexity of such calculations. This paper describes an intuitive method of determining the covariance for each of the experimental components, providing a correlation for each family of components across the multiple cases examined within a benchmark. A proof-of-principle study using the TEX-HEU experimental campaign was performed and verified that the covariance and correlation matrices can be calculated with information commonly found in the International Criticality Safety Benchmark Evaluation Project benchmarks. This study showed that the introduction of model and experimental covariances reduces the χ2 per degree of freedom from 2.203 to 1.179, indicating that the omission causes overly pessimistic bias quantifications. This technique can be seamlessly integrated to current benchmark evaluations as well as reevaluations of legacy benchmarks. [ABSTRACT FROM AUTHOR]

The American National Standards Institute/American Nuclear Society national standards 8.1 and 8.24 provide guidance on the requirements and recommendations for establishing confidence in the results of the computerized models used to support operation with fissionable materials. By design, the guidance is not prescriptive, leaving freedom to the analysts to determine how the various sources of uncertainties are to be statistically aggregated. Due to the involved use of statistics entangled with heuristic recipes, the resulting safety margins are often difficult to interpret. Also, these technical margins are augmented by additional administrative margins, which are required to ensure compliance with safety standards or regulations, eliminating the incentive to understand their differences. With the new resurgent wave of advanced nuclear systems, e.g., advanced reactors, fuel cycles, and fuel concepts, focused on economizing operation, there is a strong need to develop a clear understanding of the uncertainties and their consolidation methods to reduce them in manners that can be scientifically defended. In response, the current studies compare the analyses behind four notable methodologies for upper subcriticality limit estimation that have been documented in the nuclear criticality safety literature: the parametric, nonparametric, Whisper, and TSURFER methodologies. Specifically, the work offers a deep dive into the various assumptions of the noted methodologies, their adequacies, and their limitations to provide guidance on developing confidence for the emergent nuclear systems that are expected to be challenged by the scarcity of experimental data. To limit the scope, the current work focuses on the application of these methodologies to criticality safety experiments, where the goal is to calculate a bias, a bias uncertainty, and a tolerance limit for keff in support of determining an upper subcriticality limit for nuclear criticality safety. [ABSTRACT FROM AUTHOR]

The nuclear criticality safety issue in spent fuel reprocessing facilities is closely related to the main process and is almost equally important. It is closely related to the design and operation of reprocessing facilities. The requirement of nuclear criticality safety greatly restricts the operational capacity of the spent fuel reprocessing process, thereby affecting the economic efficiency of reprocessing. In the post-processing process, both multiphase uranium-plutonium mixed systems and multi body interaction systems are involved. The system characteristics are complex, and experimental simulation is difficult, which greatly restricts the further improvement of the production and operation capabilities of the pilot plant. However, due to the criticality safety challenges caused by reactivity changes such as the non-uniformity, dynamic complexity, and instability of the solution in the reactor under boiling nitric acid during the dissolution process, it has become a key research topic in various countries. According to the criticality safety problem of nuclear fuel dissolution process, the criticality effect of nitric acid concentration was studied. The criticality experiment data of nuclear fuel dissolution process were obtained by keeping the concentration of nitric acid in the same fuel concentration. Four experiments were conducted with different concentrations of nitric acid. During the experiment, the subcritical extrapolation method, reactivity interpolation method, and stable power method were used to complete the criticality experiment. The experimental results show that with the increase of nitric acid concentration, the relative deviation of the criticality experiment data is 0.068%, and the relative deviation between the criticality experiment results and the theoretical calculated values is 0.39%. The research results show that the reactivity of the system gradually increases as the concentration of nitric acid decreases. Therefore, it is necessary to consider the reactivity changes caused by changes in the criticality safety of the spent fuel dissolution process, and high attention is needed. According to the agreement between the experiment and the theoretical calculation, it is appropriate to use the Monte Carlo code MONK to calculate and analyze the acidity effect of the solid-liquid two-phase solution system, which can be used as a nuclear criticality safety control engineering design process for the solid-liquid two-phase nuclear fuel dissolution system. This series of experiment data can be used for calculation, verification, and safety evaluation of critical analysis under solid-liquid coexistence conditions of nuclear fuel. This paper results provide data support for improving the criticality safety control level of critical post-processing equipment.

The nuclear criticality safety issue in spent fuel reprocessing facilities is closely related to the main process and is almost equally important. It is closely related to the design and operation of reprocessing facilities. The requirement of nuclear criticality safety greatly restricts the operational capacity of the spent fuel reprocessing process, thereby affecting the economic efficiency of reprocessing. In the post-processing process, both multiphase uranium-plutonium mixed systems and multi body interaction systems are involved. The system characteristics are complex, and experimental simulation is difficult, which greatly restricts the further improvement of the production and operation capabilities of the pilot plant. However, due to the criticality safety challenges caused by reactivity changes such as the non-uniformity, dynamic complexity, and instability of the solution in the reactor under boiling nitric acid during the dissolution process, it has become a key research topic in various countries. According to the criticality safety problem of nuclear fuel dissolution process, the criticality effect of nitric acid concentration was studied. The criticality experiment data of nuclear fuel dissolution process were obtained by keeping the concentration of nitric acid in the same fuel concentration. Four experiments were conducted with different concentrations of nitric acid. During the experiment, the subcritical extrapolation method, reactivity interpolation method, and stable power method were used to complete the criticality experiment. The experimental results show that with the increase of nitric acid concentration, the relative deviation of the criticality experiment data is 0.068%, and the relative deviation between the criticality experiment results and the theoretical calculated values is 0.39%. The research results show that the reactivity of the system gradually increases as the concentration of nitric acid decreases. Therefore, it is necessary to consider the reactivity changes caused by changes in the criticality safety of the spent fuel dissolution process, and high attention is needed. According to the agreement between the experiment and the theoretical calculation, it is appropriate to use the Monte Carlo code MONK to calculate and analyze the acidity effect of the solid-liquid two-phase solution system, which can be used as a nuclear criticality safety control engineering design process for the solid-liquid two-phase nuclear fuel dissolution system. This series of experiment data can be used for calculation, verification, and safety evaluation of critical analysis under solid-liquid coexistence conditions of nuclear fuel. This paper results provide data support for improving the criticality safety control level of critical post-processing equipment. [ABSTRACT FROM AUTHOR]

BackgroundMicro-reactors can be used as a lunar surface power or spacecraft power source for space exploration. Before launching the reactor, a safety analysis should be conducted to prevent a launch accident. Currently, the safety analysis of the radioactive isotope power system does not fully include the safety analysis of the reactor. The main critical safety analysis scenario is that the reactor falls and hits the concrete ground from a high altitude. The reactor may return to criticality after high-speed impact.PurposeThis study aims to investigate the nuclear safety characteristics of a space reactor subjected to dynamic shock under high-speed impact conditions.MethodsFirst of all, based on internal and surface unstructured grids, two simplified reactor models corresponding to two high-speed impact scenarios, i.e., pure fuel reactor vertical impact with ground, and cylinder reactor with a reflector layer and shielding layer impact the ground at a 30° angle were established. Then, the ABAQUS finite element method and unstructured mesh Monte Carlo method of particle transport were combined to predict the criticality properties of the pure fuel and cylindrical reactor during high-speed impact. Based on the surface and internal unstructured mesh Monte Carlo transport technology, the criticality safety analysis platform of micro-reactor under high speed impact was established.ResultsThe results show that the keff induced by the deformation may increase with time for the above mentioned two simplified reactors. The maximum increase in the keff of the pure fuel reactor can reach 1 000×10-5, whereas for the cylinder reactor, the keff is improved to a maximum of 200×10-5. Considering the non-uniform density effect, reactivities of -666×10-5 and -132×10-5 are introduced into the two reactors.ConclusionsThe critical safety characteristics of the reactor under different impact conditions should be evaluated to ensure sufficient safety margins under such accident conditions.

Current designs of both large reactor units and small modular reactors utilize a nuclear fuel with increasing enrichment. This increasing demand for better nuclear fuel utilization is a challenge for nuclear fuel handling facilities. The operation with higher enriched fuels leads to reduced reserves to legislative and safety criticality limits of spent fuel transport, storage and final disposal facilities. Design changes in these facilities are restricted due to a boron content in steel and aluminum alloys that are limited by rolling, extrusion, welding and other manufacturing processes. One possible solution for spent fuel pools and casks is the burnup credit method that allows decreasing very high safety margins associated with the fresh fuel assumption in spent fuel facilities. This solution can be supplemented or replaced by an alternative solution based on placing the neutron absorber material directly into the fuel assembly, where its efficiency is higher than between fuel assemblies. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the fuel assemblies. The paper summarizes possibilities of fixed neutron absorbers for various nuclear fuel and fuel handling facilities. Moreover, an absorber material was optimized to propose alternative options to boron. Multiple effective absorbers that do not require steel or aluminum alloy compatibility are discussed because fixed absorbers are placed inside zirconium or steel cladding.

• A two-dimensional model considering bubble growth, the interaction between pressure and radiolytic gas bubbles, solution movement and the influence of liquid level changes is proposed to predict transient pressure and power pulses during the nuclear criticality in the fuel solution system. • Development of a simulation code to evaluation of criticality transients of fuel solution system. • The impact of pressure on the radiolytic gas release and the power pulse is investigated by comparing the predicted results in the IP model and the CP model. • The model considers the influence of pressure change on bubble nucleation, large bubbles formation and growth. The prediction of criticality excursion in fuel solution system has a significant impact on criticality safety analysis. The behavior of radiolytic gas bubbles, which may be influenced by inertial pressure, is crucial for the progression of criticality excursions. A two-dimensional model is proposed to calculating the bubble growth in order to simulate power and pressure in fuel solution system. The model is verified with experiments conducted at the SILENE facility and the influence of inertial pressure on radiolytic gas bubbles is analyzed by comparing the calculation results with and without the consideration of pressure changes. The comparison results indicate that pressure initially slows down the rate of power decline, while it speeds up the rate of power decline in the later stage of power decrease. Neglecting the pressure change during fast pulse transients may leads to a significant underestimation of the energy in the first burst. However, as the maximum inverse period of those experiments decreases, this underestimation becomes less significant and may even be overestimated slightly in turn. [ABSTRACT FROM AUTHOR]

The International Criticality Safety Benchmark Evaluation Project, a sanctioned program under the auspices of the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development, has been a highly successful and productive collaboration, now encompassing over 5,000 evaluated experimental benchmarks trusted and relied upon throughout the international nuclear communities. The success of this project led to the development of the International Reactor Physics Experiment Evaluation Project, which is dedicated to the evaluation of benchmark experiment data to sustain current and future reactor physics validation needs. These exemplary programs, and their widely utilized handbooks, serve as gold standards to which other databases strive to emulate. The purpose of the two projects is to preserve modern and legacy experimental data and evaluate it in a standardized handbook format to provide quality benchmarks to support modern and future criticality safety and reactor physics validation. These two projects have often served as the mechanism through which historic and modern neutronics experiments are evaluated and shared across international borders, to best provide unique, high-quality peer-reviewed, and often otherwise unavailable, benchmark data. The contents of these handbooks are utilized not only in validating criticality safety, reactor physics, and advanced reactor calculations, but are used to validate neutronics calculations and nuclear data for most other nuclear applications. This manuscript discusses both international programs and available content to enable advanced reactor design validation.

Copper is an important structural material in various nuclear energy applications, therefore the correct knowledge of copper cross sections is crucial. The presented paper deals with a validation of different copper transport libraries by means of activation of selected samples. An intense 252Cf(sf) source with a reference neutron spectrum was used as a neutron source. After irradiation, the samples were measured using a high purity germanium detector and the dosimeter reaction rates were inferred. These experimental data were compared with MCNP6 calculations using CENDL-3.1, JENDL-4.0, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.2 and JEFF-3.3 evaluated Cu transport libraries. The experiment specifically focuses on 58Ni(n,p)58Co, 93Nb(n,2n)92mNb, 197Au(n,g)198Au and 55Mn(n,g)56Mn dosimetry reactions. Evaluated activation cross sections of these dosimetric reactions were taken from the IRDFF-II library. The best library performance depends on the energy region of interest.

Calculated critical and subcritical data for systems of mixed oxides of plutonium and uranium are presented for review and subsequent incorporation into a revision of ANSI/ANS 8.12. The system specifications were prepared in parallel with the ISO standard ISO 11311:2011, involving experience from international mixed oxide fuel fabrication facilities. Data are presented for two mass fractions of plutonium to uranium plus plutonium (at 12.5% and 35%), damp mixed oxide density up to a theoretical density of 11.03 g/cm3 (for the 12.5% mass fraction, ignoring the variation due to plutonium isotopic content) and up to a density of 3.5 g/cm3 (for the 35% mass fraction), three plutonium isotopic mass fractions (100% 239Pu, 95% 239Pu with 5% 240Pu, and a reactor grade composition with 20% 240Pu and the balance 239Pu, 241Pu, and 242Pu), systems at two moderation conditions (water mass fractions at 3% and at optimal conditions) and systems at two water reflector conditions (nominal 2.5 cm and full 30.0 cm). Parameters of interest presented are values of volume (L), infinite cylinder diameter (cm), infinite slab thickness (cm), mass (kg), cylinder linear density (g/cm), and slab areal density (g/cm2). Values are presented at calculated multiplication factors of 1.000, 0.980 and 0.950. Calculations were made by various subject matter experts with various computational programs and cross section libraries. Validation methods were reviewed and discussed herein though data propagated to revision of ANSI/ANS 8.12 will reflect a conservative subcritical margin selected based on expert judgment. [ABSTRACT FROM AUTHOR]

This work presents two different methods for quantifying and propagating the uncertainty associated with fuel composition at end of life for cask criticality calculations. The first approach, the computational approach uses parametric uncertainty including those associated with nuclear data, fuel geometry, material composition, and plant operation to perform forward depletion on Monte-Carlo sampled inputs. These uncertainties are based on experimental and prior experience in criticality safety. The second approach, the data-driven approach relies on using radiochemcial assay data to derive code bias information. The code bias data is used to perturb the isotopic inventory in the data-driven approach. For both approaches, the uncertainty in keff for the cask is propagated by performing forward criticality calculations on sampled inputs using the distributions obtained from each approach. It is found that the data driven approach yielded a higher uncertainty than the computational approach by about 500 pcm. An exploration is also done to see if considering correlation between isotopes at end of life affects keff uncertainty, and the results demonstrate an effect of about 100 pcm.

A key challenge for the introduction of any design changes, e.g., advanced fuel concepts, first-of-a-kind nuclear reactor designs, etc., is the cost of the associated experiments, which are required by law to validate the use of computer models for the various stages, starting from conceptual design, to deployment, licensing, operation, and safety. To achieve that, a criterion is needed to decide on whether a given experiment, past or planned, is relevant to the application of interest. This allows the analyst to select the best experiments for the given application leading to the highest measures of confidence for the computer model predictions. The state-of-the-art methods rely on the concept of similarity or representativity, which is a linear Gaussian-based inner-product metric measuring the angle—as weighted by a prior model parameters covariance matrix—between two gradients, one representing the application and the other a single validation experiment. This manuscript emphasizes the concept of experimental relevance which extends the basic similarity index to account for the value accrued from past experiments and the associated experimental uncertainties, both currently missing from the extant similarity methods. Accounting for multiple experiments is key to the overall experimental cost reduction by prescreening for redundant information from multiple equally-relevant experiments as measured by the basic similarity index. Accounting for experimental uncertainties is also important as it allows one to select between two different experimental setups, thus providing for a quantitative basis for sensor selection and optimization. The proposed metric is denoted by ACCRUE, short for Accumulative Correlation Coefficient for Relevance of Uncertainties in Experimental validation. Using a number of criticality experiments for highly enriched fast metal systems and low enriched thermal compound systems with accident tolerant fuel concept, the manuscript will compare the performance of the ACCRUE and basic similarity indices for prioritizing the relevance of a group of experiments to the given application.

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies. [ABSTRACT FROM AUTHOR]

The present work is concerned with the problem of determining the reactivity of an assembly from which an unknown number of fuel pins are removed at random locations. Because it is usually designed to be under-moderated, the reactivity of an incomplete assembly will pass by an optimum when the number and position of the missing rods vary. Estimating this maximal reactivity is paramount for criticality safety but is difficult to achieve in practice, because of a prohibitive combinatorial factor, that forbids any direct approach of the problem. It is shown that the maximum reactivity of an isolated PWR assembly with missing fuel rods can be estimated with very few calls on time expensive computer simulations. Drawing a parallel with the field of neutron transport in random geometries, it is indeed shown that the problem can be reduced to an optimum search in an appropriate bounded, bi-dimensional space. [ABSTRACT FROM AUTHOR]

To proceed with the decommissioning of the Fukushima Daiichi Nuclear Power Station, analyses of unexpected fuel debris criticality accidents are needed. Supercritical transient analyses have been conducted for fuel debris using the Multiregion Integral Kinetic (MIK) code, which can take the space dependence of fuel debris into account. In those analyses, reactivity is assumed as stepwise insertion because the MIK code does not include delayed neutron effects, which might be negligible. However, reactivity insertion may not always be stepwise. Therefore, it is important to clarify an applicable range of the MIK code for nonstepwise insertion, such as ramp reactivity insertion. To show that kinetics codes without delayed neutron effects could be applied for a supercritical transient induced by ramp reactivity insertion, we established a method to clarify its applicable range. An analysis using the point reactor kinetics model was introduced as a pre-analysis to clarify this range in the case of ramp reactivity insertion in terms of the contribution of delayed neutrons. We applied the methodology to a simple cylindrical fuel debris system and successfully demonstrated a supercritical transient analysis for ramp reactivity insertion using the MIK code. [ABSTRACT FROM AUTHOR]

This work presents an assessment of the applicability of existing benchmark critical experiments to the criticality safety code validation for a large-capacity high-assay low-enriched uranium (HALEU) transportation package concept. Numerous next-generation nuclear reactor designs require HALEU fuel, which is characterized by an enrichment between 5 and 20 wt% 235U. The U.S. Department of Energy (DOE) has proposed to recover and downblend highly enriched uranium from DOE-owned used nuclear fuel to accelerate the demonstration of commercially viable microreactor technologies. One element of the infrastructure needed to demonstrate HALEU-fueled reactors is the ability to safely transport enriched product to be used for fuel fabrication. There is uncertainty as to whether existing critical benchmark experiment data are sufficient to support criticality safety code validation for HALEU transportation applications. The anticipated chemical form of the HALEU in the proposed transportation concept is UO2 with 20 wt% 235U/U. The concept uses a combination of an existing transportation packaging design and a novel basket design, including borated aluminum flux traps. The basket provides space for 18 reusable, stainless steel canisters that contain the HALEU. In 10 CFR 71, normal conditions of transport (NCTs) and hypothetical accident conditions (HACs) are defined for fissile material transportation packages. NCT and HAC KENO-VI models of the transportation package were developed using the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.2.3 computer code package, and optimum moderation conditions were determined using the SCALE SAMPLER sequence. The SCALE Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) sequences were then used to compare the neutronic characteristics of 1584 International Criticality Safety Benchmark Evaluation Project benchmark critical experiments with the NCT and HAC HALEU transportation models. The TSUNAMI integral correlation coefficient ck was the criterion used to rank neutronic similarity. Thirty-four experiments were identified as similar (ck ≥ 0.9) to the NCT model, and 55 experiments were identified as similar to the HAC model. Hundreds of experiments were also identified as at least marginally similar (ck ≥ 0.8) to both models. The results indicate that additional critical experiments are unlikely to be needed to support HALEU transportation criticality safety analyses for package concepts similar to the concept package analyzed. [ABSTRACT FROM AUTHOR]

In order to support the development and deployment of uranium fuels with enrichment beyond 5%, additional criticality safety methodologies are needed to prevent the possibility of criticality accidents. Specifically, improved methodologies for computer code validation using evaluated critical experiments, particularly for dissolver systems, need to be developed. Potential candidate evaluations and methodologies are presented to evaluate the effect of heterogeneity for intermediate-enrichment uranium systems. [ABSTRACT FROM AUTHOR]

Transient analysis for possible prompt supercritical accidents of fuel debris in the Fukushima Daiichi Nuclear Power Station is quite important. However, unlike solution fuel systems, there is little knowledge about supercritical transient analysis in fuel debris systems. In particular, reactivity feedback effects, which may have a significant impact on the results of the analysis, are important and require further study. In particular, the impacts of radiolysis gas void and moderator boiling should be discussed. Thus, the purpose of this study is to clarify whether the reactivity feedback effects of radiolysis gas and boiling of the moderator impact the supercritical transient analysis in fuel debris systems. To accomplish this, we used a power profile obtained by the MIK code with the Doppler reactivity feedback effect; radiolysis gas analysis and heat transfer analysis were performed. For the radiolysis gas analysis, the AGNES2 model was modified to consider the difference between solution fuel and fuel debris systems. The heat transfer analysis used an OpenFOAM solver to perform conjugate heat transfer calculations. We found that the radiolysis gas void was negligible when probable G values, which are the generation number of molecules per absorbed energy, were used. In addition, the results showed that boiling could be also negligible under most conditions. However, we found that the boiling time may be earlier than the peak time of the power when the radius of the fuel debris particle is small. In this case, ignoring the boiling may give conservative results. These considerations should be included in future analyses. [ABSTRACT FROM AUTHOR]

Nuclear energy has been one of the sustainable energy sources, but after the Fukushima Daiichi nuclear accident, large-scale light water reactors are losing price competitiveness due to the rising costs to meet elevated safety standards. On the other hand, small modular reactors (SMRs) have been developed by various teams and are expected to provide not only electricity but also heat for small communities, chemical plants, factories, mines, and hydrogen production. Since 2017, a multipurpose very small modular reactor (vSMR), namely, Mobile-Very-small reactor for Local Utility in X-mark (MoveluXTM), has been studied at Toshiba Energy Systems and Solutions Corporation as a feasible distributed energy source. The main concept to MoveluX is a heat pipe–cooled calcium hydride–moderated core to simplify the reactor system while increasing inherent safety and nuclear security. Portable vSMRs are useful for remote places; therefore, criticality safety during their transport is essential for vSMRs to gain popularity. In a previous paper, we discuss positive temperature reactivity coefficients of the hydride-moderated core and its control method. The phenomenon is caused by thermal-neutron spectrum shifts at increased temperatures. In the current paper, we show that a positive temperature reactivity coefficient can be utilized to maintain subcriticality during transport. The reactor core requires preheating to achieve criticality, which means the core does not become critical even though safety rods have been extracted in the low-temperature range. The positive reactivity in the low-temperature range establishes inherent criticality safety during transport of the reactor system. [ABSTRACT FROM AUTHOR]

Removal of fuel debris is regarded as one of the most important operations in the decommissioning of the Fukushima Daiichi nuclear power station (1F-NPS) to decrease long-term risk. To begin the operation, the consequences of possible criticality accidents must be evaluated in advance. In this work, we evaluated radiation doses during possible criticality accidents at 1F-NPS in assumptive fuel debris systems. In particular, the relationship between the water level surrounding the fuel debris and the radiation dose was investigated. This is because the water level surrounding the fuel debris is thought to have an impact on radiation dose during accidents as it affects both the reactivity and shielding of radiation. A combination of space-dependent kinetic analysis and radiation transport analysis was carried out in order to consider the special characteristics of fuel debris systems in water. Instead of traditional point-kinetics analysis, we used the Multi-region Integral Kinetic (MIK) code, which is a unique method based on Monte Carlo neutron transport calculations. The radiation transport calculation code Particle and Heavy Ion Transport Code System (PHITS) was used as well. The analyses revealed that the dose caused by criticality accidents may be the largest in systems in which part of the fuel debris is exposed to the air. [ABSTRACT FROM AUTHOR]

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate plus uncertainty approach and includes burnup credit. Burnup credit is applied by means of a computational scheme called BUCSS-R (Burnup Credit System for the Swiss Reactors–Repository case) which is complemented by the quantification of uncertainties from various sources. BUCSS-R consists in depletion, decay and criticality calculations with CASMO5, SERPENT2 and MCNP6, respectively, determining the keff eigenvalues of the disposal canister loaded with the Swiss spent nuclear fuel assemblies. However, the depletion calculation in the first and the criticality calculation in the third step, in particular, are subject to uncertainties in the nuclear data input. In previous studies, the effects of these nuclear data-related uncertainties on obtained keff values, stemming from each of the two steps, have been quantified independently. Both contributions to the overall uncertainty in the calculated keff values have, therefore, been considered as fully correlated leading to an overly conservative estimation of total uncertainties. This study presents a consistent approach eliminating the need to assume and take into account unrealistically strong correlations in the keff results. The nuclear data uncertainty quantification for both depletion and criticality calculation is now performed at once using one and the same set of perturbation factors for uncertainty propagation through the corresponding calculation steps of the evaluation method. The present results reveal the overestimation of nuclear data-related uncertainties by the previous approach, in particular for spent nuclear fuel with a high burn-up, and underline the importance of consistent nuclear data uncertainty quantification methods. However, only canister loadings with UO2 fuel assemblies are considered, not offering insights into potentially different trends in nuclear data-related uncertainties for mixed oxide fuel assemblies.

• Performed double differential scattering cross section measurements for lucite. • Measured dynamic structure factor for lucite at VISION spectrometer. • Developed a new methodology for creation of thermal scattering libraries. • Produced a new improved thermal scattering library for lucite. With the advancements in technology (both experimental and computational) the determination of the"true" experimental phonon spectrum became more accessible. In this work a methodology for producing thermal scattering libraries from the experimental data (namely the DFT + oClimax method) for lucite (C 5 O 2 H 8) n is discussed. Double differential scattering cross section (DDSCS) experiments were performed at the Spallation Neutron Source of Oak Ridge National Laboratory (SNS ORNL). New scattering kernel evaluations, based on the phonon spectrum for (C 5 O 2 H 8) n ,were created using oClimax and NJOY2016 codes. In order to compare and asses the performance of the newly created library, the experimental setup was simulated using MCNP6.1. Compared to the current ENDF/B-VIII.0, the resulting RPI (C 5 O 2 H 8) n library improved the calculation of both double differential scattering and total scattering cross sections. A set of criticality benchmarks containing (C 5 O 2 H 8) n from HEU-MET-THERM resulted in an overall improved calculation of K eff . The DFT + oClimax method is shown to be the most comprehensive method for analysis of moderator materials, due to the fact that it can be verified against all data measured at VISION, ARCS and SEQUOIA neutron spectrometers at SNS ORNL, and experimental total scattering cross section measurements. This method also provides a new technique for calculating any phonon spectrum-related quantities such as scattering law kernel, specific heat capacity, thermal conductivity, etc. for any solid state material. [ABSTRACT FROM AUTHOR]

Criticality safety analyses rely on the availability of relevant benchmark experiments to determine justifiable margins of subcriticality. When a target application lacks neutronically similar benchmark experiments, validation studies must provide justification to the regulator that the impact of modeling and simulation limitations is well understood for the application and often must provide additional subcritical margin to ensure safe operating conditions. This study estimated the computational bias in the critical eigenvalue for several criticality safety applications supported by only a few relevant benchmark experiments. The accuracy of the following three methods for predicting computational biases was evaluated: the Upper Subcritical Limit STATisticS (USLSTATS) trending analysis method; the Whisper nonparametric method; and TSURFER, which is based on the generalized linear least-squares technique. These methods were also applied to estimate computational biases and recommended upper subcriticality limits for several critical experiments with known biases and for several cases from a blind benchmark study. The methods are evaluated based on both the accuracy of their predicted computation bias and upper subcriticality limit estimates, as well as on the consistency of the methods' estimates, as the model parameters, covariance data libraries, and set of available benchmark data were varied. Data assimilation methods typically have not been used for criticality safety licensing activities, and this study explores a methodology to address concerns regarding the reliability of such methods in criticality safety bias prediction applications. [ABSTRACT FROM AUTHOR]

• Accuracy of the MPS method is investigated by comparison of experimental and simulation results. • A calculation system was developed by coupling the MPS method and neutron transport calculation. • Numerical analysis of criticality of fuel debris during falling down in water became possible by this calculation system. The dynamic behavior of fuel debris in water raises the possibility of criticality accidents during debris removal. Hence, accurate evaluations of criticality that consider the dynamic behavior of fuel debris are an essential part of the decommissioning process. In this study, we develop a calculation system that combines the moving particle semi-implicit method and Monte Carlo neutron transport calculation code. We then clarify that it is possible to evaluate criticality safety using the actual dynamic behavior of fuel debris in water with this calculation system. We first carry out validation experiments to compare the experimental and simulation results of solid-fluid multiphase flows calculation. We demonstrate that the MPS method could calculate the dynamic behavior of fuel debris in water, mostly within the 95% confidence interval of experimental results. We then show it could also evaluate criticality during the sedimentation of cubic fuel debris. Moreover, in the case of fundamental conditions such as a large amount of cubic debris falling in water, it was possible that the effective multiplication factor was maximized in the state of falling in water rather than with complete sedimentation. [ABSTRACT FROM AUTHOR]

• Homogenization is non-conservative and only applicable for very small particles sizes. • The regular lattice model is suitable if an adequate equivalent diameter d eq is chosen. • A unique d eq value may not represent adequately the debris in the whole porosity range. • Taking very high d eq do not always guarantee conservative results. • The BCC lattice is preferred due to its simplicity and the covered porosity range. After the Fukushima accident, the interest of the scientific community in severe accident research has been renewed. One of the severe accident research issues that needs to be further investigated is the recriticality potential of the debris bed that is formed after the core meltdown. The uncertainty regarding fuel debris conditions is very high. Consequently, one of the challenges of the criticality evaluation is the neutronic modeling of the debris itself. Conservative assumptions and simplifications have to be performed to overcome the uncertainties and to achieve a computationally feasible debris bed model. This paper presents a suitability analysis of several debris bed models for a criticality evaluation. The objective is to identify the most suitable model, i.e. the model with the best compromise between accuracy and simplicity, from a set of candidates. For that, a detailed near-to-reality model was developed whose higher fidelity results have been used as a reference to evaluate the adequacy of various simplifications. This work focuses on the modeling of the porous internal structure of the debris, concretely on the fuel particles, which are characterized by their shape, size and spatial distribution. The final proposed neutronic model represents the debris bed as a regular 3D arrangement of monosized spherical particles. This is a very suitable model delivering accurate results as long as an appropriate equivalent diameter d eq is chosen. However, it was proved that not always the same d eq value is adequate for representing the debris in the whole porosity range. The Monte Carlo codes MCNP6.1 and Serpent 2.1 were used to construct the debris bed models and to compute the infinite multiplication factor (k ∞). [ABSTRACT FROM AUTHOR]

In this study, an analysis on burnup credit for cask criticality safety in BWR spent fuel is conducted. Accurate burnup credit can be used to reduce overly conservative safety margins to increase shipping and storage efficiency while maintaining criticality level within regulatory limits. This analysis is based on advanced lattice depletion models that capture various complexities associated with BWR operation. This paper describes the second part of the two-part study which performs an out-of-core analysis of spent fuel in a transportation/storage cask. The first part of the study (Radaideh et al., 2019) developed the set of depletion models used here. In this paper, the spent fuel compositions resulting from these depletion models are used for cask criticality calculations. Uncertainty quantification of cask k e f f is performed by combining the uncertainty in isotope inventory, nuclear data, and the statistical sampling in KENO-V.a. The uncertainty in isotopic inventory is quantified by performing a validation analysis by comparing spent fuel compositions calculated by 2D TRITON to experimentally determined spent-fuel assay data for three reactors: Fukushima Daini-2, Cooper-1, and Gundermmingen-A. The validation results demonstrate good agreement for the uranium isotopes as compared to the plutonium isotopes. Also, it was found that the uncertainty in cask k e f f is dominated by the isotopic uncertainty and can reach about 2500 pcm, and as low as about 1700 pcm. Final results show that axial power profile, axial coolant density, control rod modeling, and the presence of gadolinium in 3D simulations have the largest effects on BWR burnup credit. This implies the need for detailed 3D modeling for accurate BWR burnup credit analysis. In addition, based on the UQ analysis considering both actinide only and actinide and fission products sets, the cask remains subcritical within 2 σ for all depletion cases analyzed (C0-C9), even though the cask is assumed to be flooded with water and the lattices are discharged at their peak reactivity. [ABSTRACT FROM AUTHOR]

The purpose of this technical note is to consolidate the notations used for describing parameters that pertain to neutron multiplicity mathematics relevant to various applications including nonproliferation, international safeguards, and criticality safety among others. The nomenclatures used in these techniques vary widely depending on the origin of the work and their applications. We aim to consolidate many of the previously used notations in a single document to enhance past, present, and future technical exchanges pertaining to neutron multiplicity. This will help avoid confusion in future publications and will facilitate wider application-independent advancements and utility of peer-reviewed findings. A brief introduction and history of neutron multiplicity counting is presented, followed by a summary of commonly used techniques in a variety of different applications. In each section, we present the notations used in previous publications for the reader's reference. [ABSTRACT FROM AUTHOR]

Abstract Due to the increased amounts of BWR spent fuel that will need to be transported and stored, additional research is needed in the area of BWR criticality safety and burnup credit. This study performs an advanced depletion analysis for a BWR lattice capturing various BWR operating complexities simultaneously. This is done to address the compounding effects of these complexities on lattice reactivity, k ∞ uncertainty, and isotopic inventory. A set of BWR lattice models is developed with evolving complexity, these complexities include presence of a gadolinium absorber, control rod modeling, variable radial enrichment, a nonuniform axial burnup profile, non-uniform axial coolant density, control rod partial insertion, variable axial enrichment, part-length rods, and control rod movement during operation. These models have been developed rigorously and benchmarked using different codes to ensure modeling accuracy. The lattice configuration, coolant density profile, control blade history, and other operating data were based on real-world data collected from literature. It was found that averaging radial enrichment had minimal impact on the k ∞ value and reactivity peak location. The effect of axial burnup and coolant density profiles was significant on the time of peak reactivity, making the lattice reaching the peak reactivity earlier in cycle. The 3D models show slower U-235 and Gd-155 depletion compared to the 2D cases, making the 3D lattices less reactive in general. The k ∞ uncertainty for the studied design is driven by the uncertainties of U-235, U-238, Gd-155, and Gd-157 at beginning of cycle, these are replaced later by plutonium isotopes after depletion. The effect of variable axial enrichment and part-length rods showed a significant impact on U-235 and Gd-155 depletion, making those design complexities important for criticality safety considerations. It was found that BWR modeling require many complexities which make the depletion calculations very expensive even for a single lattice. Furthermore, this adds more difficulty on the brute-force sampling-based uncertainty analysis. [ABSTRACT FROM AUTHOR]