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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Hybrid Technique in SCALE for Fission Source Convergence Applied to Used Nuclear Fuel Analysis

The new hybrid SOURCE ConveRgence accelERator (SOURCERER) sequence in SCALE deterministically computes a fission distribution and uses it as the starting source in a Monte Carlo eigenvalue criticality calculation. In addition to taking the guesswork out of defining an appropriate, problem-dependent starting source, the more accurate starting source provided by the deterministic calculation decreases the probability of producing inaccurate tally estimates associated with undersampling problems caused by inadequate source convergence. Furthermore, SOURCERER can increase the efficiency of the overall simulation by decreasing the number of cycles that has to be skipped before the keff accumulation. SOURCERER was applied to a representative example for a used nuclear fuel cask utilized at the Maine Yankee storage site {Scaglione and Ilas}. Because of the time constraints of the Used Fuel Research, Development, and Demonstration project, it was found that using more than 30,000 neutrons per cycle will lead to inaccurate Monte Carlo calculation of keff due to the inevitable decrease in the number of skipped and active cycles used with this problem. For a fixed uncertainty objective and by using 30,000 neutron per cycle, the use of SOURCERER increased the efficiency of the keff calculation by 60%compared to a Monte Carlo calculation that used a starting source distributed uniformly in fissionable regions, even with the inclusion of the extra computational time required by the deterministic calculation. Additionally, the use of SOURCERER increased the reliability of keff calculation using any number of skipped cycles below 350.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Comparison Of Hybrid Methods For Global Variance Reduction In Shielding Calculations

For Monte Carlo shielding problems that calculate a mesh tally over the entire problem, the relative uncertainties computed for each voxel can vary widely. This can lead to unacceptably long run times in order to reduce the uncertainties in all areas of the problem to a reasonably low level. Hybrid methods using estimates from deterministic calculations to create importance maps for variance reduction in Monte Carlo calculations have been successfully used to optimize the calculation of specific tallies. For the global problem, several methods have been proposed that create importance maps that distribute Monte Carlo particles in such a way as to achieve a more uniform distribution of relative uncertainty across the problem. The goal is to compute a mesh tally with nearly the same relative uncertainties in the low flux/dose areas as in the high flux/dose areas. Methods based on only forward deterministic estimates and methods using both forward and adjoint deterministic methods have been implemented the SCALE/MAVRIC package and have been compared against each other by computing global mesh tallies on several representative shielding problems. Methods using both forward and adjoint estimates provide better performance for computing more uniform relative uncertainties across a global mesh tally.

FW-CADIS↗
DOE OSTI2013

A GENERALIZED STOCHASTIC COLLOCATION APPROACH TO CONSTRAINED OPTIMIZATION FOR RANDOM DATA IDENTIFICATION PROBLEMS

We present a scalable, parallel mechanism for stochastic identification/control for problems constrained by partial differential equations with random input data. Several identification objectives will be discussed that either minimize the expectation of a tracking cost functional or minimize the difference of desired statistical quantities in the appropriate $L^p$ norm, and the distributed parameters/control can both deterministic or stochastic. Given an objective we prove the existence of an optimal solution, establish the validity of the Lagrange multiplier rule and obtain a stochastic optimality system of equations. The modeling process may describe the solution in terms of high dimensional spaces, particularly in the case when the input data (coefficients, forcing terms, boundary conditions, geometry, etc) are affected by a large amount of uncertainty. For higher accuracy, the computer simulation must increase the number of random variables (dimensions), and expend more effort approximating the quantity of interest in each individual dimension. Hence, we introduce a novel stochastic parameter identification algorithm that integrates an adjoint-based deterministic algorithm with the sparse grid stochastic collocation FEM approach. This allows for decoupled, moderately high dimensional, parameterized computations of the stochastic optimality system, where at each collocation point, deterministic analysis and techniques can be utilized. The advantage of our approach is that it allows for the optimal identification of statistical moments (mean value, variance, covariance, etc.) or even the whole probability distribution of the input random fields, given the probability distribution of some responses of the system (quantities of physical interest). Our rigorously derived error estimates, for the fully discrete problems, will be described and used to compare the efficiency of the method with several other techniques. Numerical examples illustrate the theoretical results and demonstrate the distinctions between the various stochastic identification objectives.

Webster, Clayton G↗
DOE OSTI2016

Impact of delayed spark restrike on the dynamics of cyclic variability in dilute SI combustion

Spark-ignition (SI) engines can derive substantial efficiency gains from operation at high dilution levels. Additionally, the use of exhaust gas recirculation (EGR) for charge dilution also maintains compatibility with three-way catalysts by allowing stoichiometric operation. However, running high dilution levels increases the occurrence of misfires and partial burns, which induce higher levels of cyclic-variability in engine operation. This variability has been shown to have both stochastic and deterministic components. Factors such as in-cylinder turbulence and mixing-variations can be classified as stochastic; while, charge composition is the major source of the deterministic component through its non-linear effect on ignition and flame propagation characteristics. The use of these deterministic components has been previously explored to construct next-cycle control approaches that would allow stable operation near the edge of stability. Building on that work, this paper aims to understand the effect of spark strategies, specifically the use of a second spark (restrike) after the main spark, on engine operation at high dilution levels that were achieved using both excess air (i.e. lean combustion) and EGR.

Kaul, Brian C↗
DOE OSTI2018

Probabilistic Short-Term Wind Forecasting Based on Pinball Loss Optimization: Preprint

Probabilistic wind power forecasts that quantify the uncertainty in wind output have the potential to aid in the economic grid integration of wind power at large penetration levels. In this paper, a novel probabilistic wind forecasting approach based on pinball loss optimization is proposed, in conjunction with a multi-model machine learning based ensemble deterministic forecasting framework. By assuming the pointforecasted value as the mean at each point, one unknown parameter (i.e., standard deviation) of a predictive distribution at each forecasting point is determined by minimizing the pinball loss. A surrogate model is developed to represent the unknown distribution parameter as a function of deterministic forecasts. This surrogate model can be used together with deterministic forecasts to predict the unknown distribution parameter and thereby generate probabilistic forecasts. Numerical results of case studies show that the proposed method has improved the pinball loss by up to 35% compared to a baseline quantile regression forecasting model.

17 WIND ENERGY↗
DOE OSTI2018

APPLICATION OF THE FAVOR-OCI FRACTURE MECHANICS COMPUTER PROGRAM TO ASME CODE SECTION XI, IWB-3610 FLAW ACCEPTANCE CRITERIA EVALUATIONS

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Appli-cations of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is main-tained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all prev-iously-existing probabilistic capabilities of FAVOR.The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws.The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These com-pute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected fracture toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV. Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Dickson, Terry L.↗
DOE OSTI2018

RISK-INFORMED EXTERNAL HAZARDS ANALYSIS BY EEVE TOOLKIT

Introducing risk-informed safety analyses for external events can bring sensible benefits to the nuclear power industry easing the regulatory burden introduced in the aftermath of the Fukushima accident. In the framework of the US-DOE Light-Water Reactor Sustainability program, Risk-Informed Margin Characterization project (LWRS-RISMC), research activities were conducted at INL for developing and applying a toolkit for performing risk-informed external hazards analysis. The “EEVE” (External EVEnts) toolkit, a chain of INL and state-of-the-art codes, was set-up and applied for simulating earthquake-induced transients with internal flooding in a PWR. This was done using methodology set-up for evaluating new data and determining critical areas that would benefit most from advanced analysis methods. The methodology is agnostic and it can be applied to other external hazards than just seismic and flooding (e.g., high winds, intense precipitation, etc.). The first step was the deterministic calculations for evaluating the propagation of a set of earthquakes and for determining the effects on a fire suppression system in an auxiliary building. These deterministic calculations were performed using the LS-DYNA and the OPENSEE structural mechanics codes. Probability failures and event trees of a generic 3-loops Westinghouse PWR were derived from a generic SAPHIRE model. Seismically induced Loss of Offsite Power, Station Blackout, and fire suppression piping failure events were considered. SAPHIRE code allowed a pre-screening of all possible branches of the event tree and a selection of those with the largest Core Damage Frequency increase for further analysis using dynamic PRA methods. INL's EMRALD code performed dynamic PRA calculations of the identified sequences, integrating risk analysis with on-line deterministic safety analysis results (seismically-induced internal flooding calculations by NEUTRINO code and system analysis by RELAP5-3D code). RELAP5-3D Best-Estimate calculations coupled with RAVEN statistical code were used for deriving limit surfaces as function of relevant transient parameters. This data was used by EMRALD for determining the final core status (core damage/core safe). Finally, the coupling of seismic, flooding, system thermal-hydraulic, uncertainty quantification and probabilistic risk analysis for an advanced risk-informed external hazards analysis was demonstrated. Results showed the sensible decrease in the level of conservatism achieved using the developed risk-informed methodology, which allowed also improving the evaluation of the failure modes and failure probabilities of different components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Revised methods for few-group cross sections generation in the Serpent Monte Carlo code

This paper presents new calculation methods, recently implemented in the Serpent Monte Carlo code, and related to the production of homogenized few-group constants for deterministic 3D core analysis. The new methods fall under three topics: 1) Improved treatment of neutron-multiplying scattering reactions, 2) Group constant generation in reflectors and other non-fissile regions and 3) Homogenization in leakage-corrected criticality spectrum. The methodology is demonstrated by a numerical example, comparing a deterministic nodal diffusion calculation using Serpent-generated cross sections to a reference full-core Monte Carlo simulation. It is concluded that the new methodology improves the results of the deterministic calculation, and paves the way for Monte Carlo based group constant generation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Comparison of hybrid methods for global variance reduction in shielding calculations

For Monte Carlo shielding problems that calculate a mesh tally over the entire problem, the statistical uncertainties computed for each voxel can vary widely. This can lead to unacceptably long run times in order to reduce the uncertainties in all areas of the problem to a reasonably low level. Hybrid methods - using estimates from deterministic calculations to create importance maps for variance reduction in Monte Carlo calculations - have been successfully used to optimize the calculation of specific tallies. For the global problem, several methods have been proposed to create importance maps that distribute Monte Carlo particles in such a way as to achieve a more uniform distribution of relative uncertainty across the problem. The goal is to compute a mesh tally with nearly the same relative uncertainties in the low flux/dose areas as in the high flux/dose areas. Methods based on only forward deterministic estimates and methods using both forward and adjoint deterministic methods have been implemented in the SCALE/MAVRIC package and have been compared against each other by computing global mesh tallies on several representative shielding problems. Methods using both forward and adjoint estimates provide better performance for computing more uniform relative uncertainties across a global mesh tally. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Accelerated Monte Carlo Fission Source Convergence with Fission Matrix and Kernel Density Estimators

Accurate solution of the k-eigenvalue form of the Boltzmann radiation transport equation is required in many application areas, including nuclear reactor analysis and criticality safety. Monte Carlo calculations are free of the discretization errors inherent to deterministic methods and therefore offer the most accurate solutions. However, converging the fission source can be difficult for many problems, resulting in either a substantial computational overhead or potential inaccuracies that negate many of the benefits of using Monte Carlo methods. Difficulty in converging the fission source occurs for two distinct reasons. The first is the existence of a high dominance ratio in the problem (i.e. the largest and second largest eigenvalues have approximately the same magnitude). The second difficulty arises due to undersampling of the fission source because an insufficient number of particle histories are simulated. The first difficulty is shared by deterministic solvers when using power iteration, while the second difficulty is unique to Monte Carlo solvers. In practice, both difficulties may be present in the same problem, compounding the difficulty in achieving a converged distribution. In this paper, we consider a new approach to accelerate the convergence of the fission source based on combining the use of fission matrix methods with kernel density estimators (KDEs). It has been observed that deterministic fission matrix methods show the potential to accelerate the convergence of the fission source in Monte Carlo criticality calculations but may introduce significant additional noise for difficult problems. In this paper, we have demonstrated the use of a KDE approach in combination with the fission matrix acceleration. Although the addition of the KDE initially slows the convergence of the fission matrix acceleration, it decreases the magnitude of the noise while still offering significantly improved convergence relative to analog Monte Carlo. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Position-Dependency of Fuel Assembly Homogenization in a Pressurized Water Reactor

2-step nuclear reactor core analysis has long been established as the conventional calculational procedure for the pressurized water reactor (PWR) due to its efficiency. This is because the procedure significantly avoids the potentially huge computational burden arising from an explicit geometrical modeling and consideration of over ∼ 500 energy groups. It instead approaches the problem with spatial homogenization and group condensation: a 2-D lattice transport calculation to homogenize the fuel assembly cross-sections and a 3-D whole core nodal diffusion. One notes that the few-group constant generation using deterministic method could be complicated due to its limited flexibilities in geometrical modeling and inherent drawbacks of the multi-group approximation. Furthermore, accuracy of the deterministic few-group constant generation can especially be compromised during the resonance treatment calculation. As such, an alternative lattice calculation using the Monte Carlo (MC) approach, in place of the deterministic fuel assembly homogenization (few-group constant generation), has previously been proposed. During the assembly homogenization, directions and magnitudes of the current at boundaries are unknowns. Therefore, zero net current boundary conditions (infinite lattice) are instead used as the assembly node. The assumption is quite reasonable considering heterogeneities inside the node are more dominant than heterogeneities due to nodal location in the core. In fact, this approximation has successfully been applied to many LWR core analyses to reasonable accuracy. It should nonetheless be noted that the net current zero boundary assumption effectively ignores position-dependency of the fuel assembly homogenization in the general 2-step method. As such, there are a number of well-known researches on the extended assembly homogenization techniques, which introduce correction factors to compensate effects of the assembly environment, such as Generalized Equivalence Theory (GET), Super-homogenization method (SPH), and color set calculation method. Even though the assembly environment effect has long been acknowledged, its effect has not properly been investigated due to the limited computing power in the past and due to the reasonable accuracy of the extended assembly homogenization techniques. Nonetheless, quantification of position-dependency of the assembly homogenization in the core could still potentially offer additional value in terms of calculation accuracy to the 2-step reactor analysis. It is upon this motivation that the paper is prepared. In this research, position-dependent few-group constants were calculated and compared against those of infinite lattices. All calculations were completed using the MC Serpent2 code with ENDF/B7.0 nuclear data library. In this study, position-dependency of the fuel assembly homogenized cross sections in a PWR core has been investigated through comparison of the conventional 2-step reactor analysis with the reference high fidelity Monte Carlo simulation. It was found that the homogenized two-group constants are clearly position-dependent as the discrepancies between infinite lattice and whole-core calculations are quite significant especially in a rodded assembly. As such, the corresponding correction required due to the assembly environment effect is not negligible compared to the correction required due to assembly heterogeneity itself. It is therefore quite clear that position-dependency of the group cross sections should be considered in the 2-step procedure for a more accurate PWR core analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the 80's at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Preliminary criticality calculations have been performed using both models (deterministic and stochastic). Results show the good agreement between different models. It should also be noted that deterministic solution could still be further improved by implementing super-homogenization factors and by increasing the transport calculation order.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A Comparison of the FW-CADIS and MR-CADIS Variance Reduction Methods

The Consistent Adjoint Driven Importance Sampling (CADIS) method, which uses an adjoint deterministic estimate for making variance reduction parameters that optimize Monte Carlo (MC) shielding calculations, was introduced in 1998. This method was designed to optimize a single tally, typically an integrated response such as a detector count rate or a dose rate. Because both deterministic and stochastic methods are employed, CADIS is referred to as a hybrid method. In 2007, the forward-weighted CADIS (FW-CADIS) method was introduced to optimize several tallies at once, giving more uniform relative uncertainties among the tallies for both high-flux and low-flux areas of the problem. This method can be used for several small detectors, for semi-global problems (mesh tally over a large fraction of the problem), or for global problems (mesh tally over the entire problem). Optimization can be done for an integrated response such as a dose rate D(r) as a function of position, the total flux {phi}(r) as a function of position, or the space/energy flux {phi}(r,E). Both the CADIS and FW-CADIS methods have been in the MAVRIC sequence of SCALE for three versions (6.0 in 2009, 6.1 in 2011, and 6.2 in 2016). The ADVANTG package released in 2015 uses the CADIS and FW-CADIS methods to accelerate MCNP neutron/photon shielding calculations. In 2014, the multi-response CADIS (MR-CADIS) method was presented as a way to obtain even more uniform relative uncertainties for problems with multiple tallies than those obtained with FW-CADIS. The method does not use a forward deterministic calculation but instead uses one adjoint calculation for every tally to be optimized in the Monte Carlo calculation. This work compares the above methods.

42 ENGINEERING↗
DOE OSTI2015

Implementation and Validation of Reference Sensitivity Profile Calculations in TRIPOLI 4 - 14247

In current neutron transport calculations, particularly on continuous-energy Monte-Carlo codes, the major source of uncertainty is coming from nuclear data. In order to improve nuclear data evaluations, integral experiments are used. Feedback is then precisely obtained by re-estimating cross-section levels and correlations according to their sensitivity. k{sub eff}-sensitivity coefficients are traditionally obtained using deterministic codes since hands-on methods were not yet implemented in continuous energy Monte Carlo codes. In this paper, an Iterated-Fission-Probability-based method newly developed in the continuous energy Monte Carlo code TRIPOLI-4 will be used to calculate sensitivity coefficient breakdown on isotopes and reactions and results will be compared to those of ERANOS deterministic code. Objective is twofold: to show the reliability of the new TRIPOLI-4 method and to identify where and why profiles are slightly different. Sensitivity profiles, forward and adjoint fluxes are thus meticulously compared using a fine 1968-group energy structure. An OECD international benchmark (Uncertainty Analysis for Criticality Safety Assessment - UACSA - Phase III) is taken as working example to enforce validation and to compare to independent results (e.g. MCNP-6). This benchmark is constituted of two bare spheres filled with homogeneous mixture of uranium and polyethylene (respectively 2% and 50% U-235 enrichments) and gives an overview of issues encountered in thermal reactor cores. Several energy-group structures (1968G, 172G, 33G) are also tested in ERANOS to identify multi-group and implicit effects. Lessons learnt are that continuous Monte Carlo methods seem to be mature and compare well with deterministic codes. Only damper the U-238 elastic - and on certain conditions the U-238 capture - sensitivity coefficient for which difference can reach 30% between TRIPOLI-4 and ERANOS. Indeed concerning U-238, implicit effects are emphasized and multi-group theory remains limited by hypothesis; even a 1968-group structure is not refined enough to properly calculate U-238 elastic scattering k{sub eff}-sensitivity. This reference 3D continuous-energy sensitivity capability in TRIPOLI4-IFP will be extensively used in the uncertainty automated tool RIB of the French Criticality-Safety package CRISTAL. (authors)

42 ENGINEERING↗
DOE OSTI2015

Incorporating realistic conservatism into criticality alarm placement and emergency response: a regulatory perspective - 14425

Review of historical process plant criticality accidents shows that their consequences vary widely and are extremely difficult to predict accurately. Because of the wide variability and inherent uncertainty in estimating consequences, criticality detector placement analysis and consequence estimation for emergency response typically utilize very conservative assumptions with regard to the total fissions, time-dependent fission rate, and intervening shielding. For detector placement, a 'minimum accident of concern' is typically defined, whereas emergency consequence evaluations typically define a maximum fission source term. These require different mindsets because what is conservative for one type of analysis may be non-conservative for the other. While some conservatism is commonly practiced in both cases, excessive conservatism can result in a net increase in risk, such as resulting from inadvertent alarm activations, that can be costly to operations and counterproductive to safety. Conversely, reliance on very detailed calculational models to reduce conservatism could increase the number and complexity of controls, such as shielding for dose mitigation. Deterministic regulatory criteria have also been applied for determining where criticality detectors are to be deployed and what areas must be evacuated in the event of an accident. This paper explores several issues associated with criticality alarm placement and emergency response from a regulatory perspective. This paper will discuss insights gained from historical fuel facility accidents and significant events regarding the inherent risk of criticality, the most likely locations for criticality to occur, and the most probable characteristics of a critical excursion. This will include discussion of the previously disputed relevance of the United States Nuclear Regulatory Commission's (USNRC's) longstanding model for a critical excursion. While USNRC regulations deterministically require dual alarm coverage in all areas having greater than a 'critical mass' of fissionable material, a process is provided to grant exemptions when a risk-informed argument is technically justified. Similar risk-informed considerations have been used to justify compensatory measures allowing limited operation during periods of alarm system outage due to component failure or maintenance. This paper will explore the application of risk information and performance history to address requirements for criticality alarm system coverage, granting exemptions from deterministic criteria, allowing compensatory measures during system outages, determining evacuation areas and routes, and assessing the significance associated with alarm system issues. Finally, the impact of shielding on the USNRC's regulatory risk basis and approved subcritical margin will be reviewed. (authors)

42 ENGINEERING↗
DOE OSTI2015

A Methodology for defining intermediate level waste package characteristics that ensure criticality safety in a geological disposal facility - 14599

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. The basis for ensuring criticality safety of intermediate level waste (ILW) packages is to specify controls on the fissile material content of individual packages through the application of deterministic criticality safety assessments (CSAs). The approach is defined and controlled through RWM's Disposability Assessment process, which aims to ensure that: criticality is prevented during waste package transport to a GDF; the risk of criticality is tolerable and as low as reasonably practicable (ALARP) during GDF operations; and the likelihood and consequences of criticality arising from the evolution of waste packages after GDF closure are low. The most restrictive phase determines the package fissile material limit and, until now, in many cases the GDF post-closure phase has provided the most restrictive limit. Currently, the GDF post-closure limits are derived deterministically using highly stylised and conservative assumptions about fissile material accumulations occurring on very long timescales into the future. RWM, waste producers and regulators do not currently consider this approach to be yielding an appropriate balance of risk. Therefore, RWM is modifying the process for setting fissile material limits on ILW packages by making use of the outcomes of its recent research projects to establish the likelihood of post-closure criticality in a GDF and the consequences of hypothetical post-closure criticality events (presented in companion papers). This paper presents the revised methodology, which is based on the requirement that criticality in a GDF can be demonstrated to be of low and tolerable consequences with regard the performance of a GDF. A probabilistic 'low-likelihood package envelope' is proposed that establishes the packaging and disposal facility conditions under which post-closure criticality is considered to be of low likelihood and low consequence. The bounding conditions for establishing the extent of the package envelope have been evaluated through application of the methodologies and models developed in RWM's Likelihood of Criticality research project. It is expected that the vast majority of ILW packages would meet the package envelope requirements. The methodology also allows the use of other options, such as deterministic CSAs, for waste packages that fall outside the low-likelihood package envelope. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Stochastic level-set method for shape optimisation

We present a new method for stochastic shape optimisation of engineering structures. The method generalises an existing deterministic scheme, in which the structure is represented and evolved by a level-set method coupled with mathematical programming. The stochastic element of the algorithm is built on the methods of statistical mechanics and is designed so that the system explores a Boltzmann–Gibbs distribution of structures. In non-convex optimisation problems, the deterministic algorithm can get trapped in local optima: the stochastic generalisation enables sampling of multiple local optima, which aids the search for the globally-optimal structure. The method is demonstrated for several simple geometrical problems, and a proof-of-principle calculation is shown for a simple engineering structure.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗