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DOE OSTI2016

Use of Criticality Eigenvalue Simulations for Subcritical Benchmark Evaluations

Subcritical multiplication experiments and simulations are important for a variety of applications including nonproliferation, safeguards, and criticality safety monitoring. Recently, subcritical experiments have been evaluated for inclusion in the International Criticality Safety Benchmark Evaluation Process (ICSBEP). In addition, new subcritical experiments have recently been designed. In these experiments, list-mode data (a list of each absorption in the neutron detector) is recorded and neutron noise analysis is used to determine the leakage multiplication (ML) of each configuration. One downside to the simulation of these types of experiments is that they are computationally expensive. This is because analog Monte Carlo with no variance reduction must be performed in order to maintain the time accuracy required in the measurement analysis. This work demonstrates that criticality eigenvalue simulations can be used to estimate sensitivities and uncertainties for three parameters: singles count rate (R{sub 1}), doubles count rate (R{sub 2}), and leakage multiplication (M{sub L}). This will provide a useful aid both in the design and evaluation of new subcritical measurements. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Status and planned updates of the SINBAD evaluation of HPRR shielding experiments [Slides]

The goal is to use available data from Health Physics Research Reactor (HPRR) operation to create a benchmark report for inclusion in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, or SINBAD, as a criticality accident alarm system (CAAS) shielding benchmark.

CAAS↗

DOE OSTI2015

The experiment filtering tool in the MACSENS workbench - 14341

The MACSENS (outils Moret d'Aide a la Caracterisation et aux etudes de SENSibilite) workbench is dedicated to the advanced validation work for criticality studies. The aim of this kind of tool is to determine the experiments from a validation database being the 'closest' to an application case and, from there, to evaluate a calculation bias with a related uncertainty. To achieve this goal, the MACSENS workbench relies on different kind of information stored in various databases depending on the origin of the data. Two main kinds of data can be identified. The first one concerns the data used to describe the experiments of the validation database (e.g. the inputs needed to perform a calculation). These values characterize the chemical and geometrical models. They can be considered as constant, as they do not vary once evaluated, and some of them can be from a non-numerical nature (reflector material for example...). The second kind of data gathers the results obtained with the use of a simulation code, such as reaction rates, fluxes and leakages, sensitivities, k{sub eff}, etc. These values depend on the version of the code and are purely numerical. In the early 2000's, regarding the increasing number of experiments evaluated in the ICSBEP handbook, the OECD/NEA (Organization for Economic Co-operation and Development/Nuclear Energy Agency), under the ICSBEP (International Criticality Safety Benchmark Evaluation Project) working group direction, decided to store such information in a database and to provide a software, called DICE (Database for the International Criticality safety benchmark Evaluation project), allowing users to search easily criticality experiments of interest. Up to now, more than 4400 critical or subcritical configurations have been compiled by ICSBEP, described and stored in the DICE database. Over time, a validation work has been done on the data, with the participation of IRSN, and a quality assurance policy has been defined. For these reasons, the DICE database is highly reliable. These features were fundamental for the MACSENS developers, who were looking for a database containing descriptive data of the experiments included in the MORET code validation database. The paper describes the solution chosen by IRSN to take advantage of the DICE database content and to overcome the DICE limits. It also explains the quality assurance policy applied to guaranty the reliability of the stored data. It will also discuss the use of the experiment-filtering module of the MACSENS workbench to select a set of experiments and to examine numerical parameter trends for an application case. (authors)

42 ENGINEERING↗

DOE OSTI2016

Benchmark Development for TREAT Minimum Critical Mass Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT minimum critical mass core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-001. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Criticality Safety Validation of SCALE 6.1 (Revised)

The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Experiment Program (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). A total of more than 350 cases from this reference have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID) maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the SCALE 6.1 code system with ENDF/B-VII.0 cross-section data in 238-group and continuous energy is assessed using the VALID models of benchmark experiments. The TSUNAMI tools for sensitivity and uncertainty analysis are utilized to examine some systems further in an attempt to identify potential causes of unexpected results. The critical experiments available for validation of the KENO V.a code cover eight different broad categories of systems. These systems use a range of fissile materials including a range of uranium enrichments, various plutonium isotopic vectors, and mixed uranium/plutonium oxides. The physical form of the fissile material also varies and is represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover both fast and thermal spectra. Over 300 of the total cases used utilize the KENO V.a code. The critical experiments available for the validation of the KENO-VI code cover three broad categories of systems. The fissile materials in the systems vary and include high- and intermediate-enrichment uranium and mixed uranium/plutonium oxides. The physical form of the fissile material is either metal or rod arrays in water. As with KENO V.a, both fast and thermal neutron energy spectra are represented in the systems considered. The results indicate generally good performance of both the KENO V.a and KENO-VI codes across the range of systems analyzed. The bias of calculated k eff from expected values is less than 0.9% Δk in all cases. All eight categories of experiments show biases of less than 0.5% Δk in KENO V.a with the exception of intermediate enrichment metal systems using the 238-group library. The continuous energy library generally manifests lower biases than the multigroup data. The KENO-VI results show slightly larger biases, though this may primarily be the result of modeling systems with more geometric complexity, which are more difficult to describe accurately, even with a generalized geometry code like KENO-VI. Several additional conclusions can be drawn from the results of this validation effort. These conclusions include that the TSUNAMI tools can be used successfully to explain the cause of aberrant results, that some evaluations in the IHECSBE should be updated to provide more rigorous expected k eff values and uncertainties, and that potential cross-section errors can be identified by detailed review of the results of this validation. It also appears that the overall cross-section uncertainty as quantified through the SCALE covariance library is overestimated. Overall, the KENO V.a and KENO-VI codes are shown to provide consistent, low-bias results for a wide range of physical systems of potential interest in criticality safety applications.

Marshall, William (BJ) J.↗

DOE OSTI2022

SCALE 6.2.4 Validation: Nuclear Criticality Safety

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Pebble Tanker Model for Nuclear Criticality Safety Needs

This report documents a study performed to investigate the requirements for criticality safety benchmark experiments for high-assay, low-enriched uranium (HALEU) fuel in transportation applications. In this work, an exploratory application model, the “Pebble Tanker,” was developed to represent TRISO fuel in a transportation scenario for an analysis of the validation basis in industrial quantities. An aspect of the criticality validation process involves assessing the “similarity” between application and experimental benchmark systems through an integral index parameter evaluation. Here, this includes propagating nuclear data uncertainties and calculating a correlation coefficient (hereinafter referred to as “c k ”) to evaluate the similarity of benchmark experiments compared with the application Pebble Tanker model. Finding sufficient critical benchmark experiments allows for the evaluation of bias and bias uncertainty, thus determining the upper subcritical limit (USL) of the transportation package. A target k eff of ~0.94 was used in this work to establish appropriate modeling conditions, reflecting a reasonable estimate for a USL. Two container models were investigated: one with the Hermes-type pebble and one with the Pebble Bed Modular Reactor (PBMR)–type pebble. The models were simplified, considering only fuel, containment structure, and either water or air. This allows a focus on the underlying physics of applications involving TRISO fuel pebbles using the Pebble Tanker model. A crucial consideration is the transport package's ability to safely hold pebbles while flooded, maintaining subcritical conditions. Tools available in the SCALE 6.3.1 suite—the CSAS6-Shift, TSUNAMI-3D-Shift, and TSUNAMI-IP sequences—were employed for neutronics and sensitivity and uncertainty (S/U) analysis of the Pebble Tanker. Findings demonstrated sufficient available critical experiment benchmarks to perform a validation of the Pebble Tanker in the most reactive state, i.e., when the Tanker is flooded.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Preliminary Design of Critical Experiments Involving Commercially Available B 4 C Neutron Absorber Plates with Low-Enriched UO 2 Fuel

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is an initiative to provide high-quality benchmark data in a standardized format for criticality safety analysts to use to validate calculational tools and nuclear data. In the last released 2022 version, the ICSBEP handbook contains over 5,000 critical experiment descriptions, results, and associated models. This paper describes the critical configurations proposed and calculations performed as part of the conceptual design phase of Integral Experiment Request (IER) 554 (IER-554). The goal of IER-554 is to add to the benchmark experiments available to the criticality safety community by designing critical experiments that can analyze how adding commercially available Boralcan plates to an assembly of low-enriched UO2 fuel rods affects the effective neutron multiplication factor (k eff ). Boralcan, a neutron absorber product developed by Rio Tinto, is made of B4C in an Al 1100 matrix and is commonly used for criticality suppression in fuel storage pools. In the current version of the ICSBEP handbook, only a few dated critical experiments involve B4C materials, and none involve Boralcan in a thin plate shape. The experiments designed as part of IER-554 are intended to be performed at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. SPRF/CX is a well-characterized assembly considered trustworthy by the benchmarking community because of the numerous high-quality evaluations with very low experimental uncertainties included in the ICSBEP handbook. Figure 1 illustrates the assembly for a particular configuration from LCT-078, one of the published benchmarks in the ICSBEP handbook. The experiments will be moderated and reflected by light water at ambient atmospheric pressure. Before the plates are inserted in the critical assembly, they will be characterized with x-ray computed tomography (XCT) to identify the sizes and distribution of the B 4 C powder particles inside the plates. The preliminary design calculations were performed with SCALE 6.3.0 using the KENO V.a sequence for the criticality calculations, the CE-TSUNAMI-3D sequence for the sensitivity and uncertainty studies, and the ENDF/B-VII.1 Continuous Energy cross section library. All the calculation results presented have 10 pcm statistical uncertainties. A set of critical experiments with Boralcan plates could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material. Experiments would also help validate k eff calculations, and the industry could use these validations to change the B loading credit limits from the US Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. This paper summarizes only part of the analysis documented in the preliminary design report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Criticality Safety Validation of SCALE 6.2.2

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

UMo Benchmark Experiment Data Needed to Support Computational and Nuclear Data Validation

Currently there are almost no benchmark experiment data available directly supporting neutronics modeling and simulation of ~20 wt.% 235U, UMo fuel contained within the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) international benchmark handbooks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and International Reactor Physics Experiment Evaluation Project (IRPhEP). The benchmarks within these two handbooks represent the international standards used to support computational and nuclear data validation. Recent activities in nuclear data development supporting new neutronics thermal scattering cross sections for UMo fuel indicate the necessity to provide quality benchmark experiment data that can be utilized to support their validation and development. Preparation of high-integrity neutronics benchmarks from UMo experimental data will ensure proper integral testing of these cross sections, as well as serve to support computational validation of contemporary modeling and simulation software.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Benchmarking of the Updated Resolved Resonance Region Evaluations of Copper

A new resolved resonance region (RRR) evaluation for the two naturally occurring isotopes of copper, {sup 63}Cu and {sup 65}Cu was performed at Oak Ridge National Laboratory (ORNL), and the benchmarking results based on the four highly enriched uranium (HEU), intermediate spectrum, copper-reflected Zeus experiments were presented in June 2014. Since then, two significant updates to the evaluations have been performed: first, the data for the experimental measurement of the capture cross section between 220 keV and 300 keV were renormalized and the resonance parameters were refitted. Second, the angular distribution of elastic scattering initially reported in point-wise form in File 4 of the ENDF file format with high fidelity was significantly reduced in size to allow for better computer storage. Twenty-seven criticality safety benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) were calculated using these updated resolved resonance region evaluations for the two isotopes of copper. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system Benchmark experiments at the CEA VALDUC silene facility - 14454

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'energie Atomique et aux energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available that show much better agreement with the measured values. (authors)

42 ENGINEERING↗

DOE OSTI2024

Delayed Critical Highly Enriched Uranium Metal Cylinders with Thin Graphite Top and Bottom Reflectors

Nine 7, 11, and 15 in. diam highly enriched uranium (HEU, 93.15 wt % 235 U) metal cylinders were assembled on the vertical assembly machine in the Oak Ridge Critical Experiments Facility (ORCEF) and had 1, 2, or 3 in. thick HLM graphite reflectors on the top and bottom. The experiments, which were performed between April 3, 1970, and February 18, 1971, used 23 operational days at ORCEF. Before those experiments were carried out, unreflected and unmoderated, graphite- and polyethylene-reflected, and polyethylene-moderated HEU metal cylinders had been assembled to obtain delayed criticality at ORCEF in the 1960s and reported by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) at the Nuclear Energy Agency (NEA)*. The data from the nine critical experiments are acceptable for use as criticality safety benchmark experiments for the NEA’s ICSBEP once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with HEU metal at ORCEF, the uncertainties in k eff are expected to be as low as ±0.0004.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Intrinsic value of the international benchmark projects, ICSBEP and IRPhEP, for advanced reactor development

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 paper discusses both international programs and available content to enable advanced reactor design validation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Thermal/Epithermal Experiments with Hafnium (TEX-Hf)

Critical experiment benchmarks are essential for validating calculational techniques used by criticality safety engineers to guarantee safe fissile material operations, yet there remain significant areas of need where existing benchmarks are insufficient or entirely absent. There is a particularly notable shortage of intermediate benchmarks (where the majority of fissions are caused by neutrons with energy between 0.625 eV and 100 keV). This work describes the design of a set of critical experiments that address this gap by developing a highly enriched uranium (HEU) test bed into which various materials of interest (diluents) can be substituted. The first diluent to be investigated with this test bed is hafnium, a neutron poison used in nu clear applications for which there are no sufficiently sensitive benchmarks. Twenty-one critical configurations made up of stacked disks of HEU, polyethylene, and hafnium were designed and simulated in MCNP6 to calculate their fission energy spectra and the sensitivity of k eff to changes in various cross sections. It was shown that configurations spanned the energy range from thermal (below 0.625 eV), through the intermediate energy range (0.625 eV to 100 keV), to fast (above 100 keV). Furthermore, it was determined that this set of experiments is sensitive to the hafnium and 235 U cross section data. These test bed experiments were designed to allow substitution of various diluent plates and can therefore be used to quickly and easily produce many more benchmarks.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Graphite and Beryllium Reflector Critical Assemblies of UO 2 (Benchmark Experiments 2 and 3)

A series of experiments was carried out in 1962-65 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for use in space reactor research programs. A core containing 93.2 wt% enriched UO 2 fuel rods was used in these experiments. The first part of the experimental series consisted of 252 tightly packed fuel rods (1.27-cm triangular pitch) with graphite reflectors, the second part used 252 graphite-reflected fuel rods organized in a 1.506-cm triangular-pitch array, and the final part of the experimental series consisted of 253 beryllium-reflected fuel rods in a 1.506-cm-triangular-pitch configuration and in a 7-tube-cluster configuration. Fission rate distribution and cadmium ratio measurements were taken for all three parts of the experimental series. Reactivity coefficient measurements were taken for various materials placed in the beryllium reflected core. All three experiments in the series have been evaluated for inclusion in the International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbooks. The evaluation of the first experiment in the series was discussed at the 2011 ANS Winter meeting. The evaluations of the second and third experiments are discussed below. These experiments are of interest as benchmarks because they support the validation of compact reactor designs with similar characteristics to the design parameters for a space nuclear fission surface power system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available that show much better agreement with the measured values.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Validation for {sup 233}U-Fueled Systems in KENO V.a in SCALE 6.2

Validation of the KENO V.a Monte Carlo code has been performed in recent years using benchmark experiments contained in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The models used in these recent validation efforts have been drawn from the Verified, Archived Library of Inputs and Data (VALID). Models must be generated and reviewed by qualified individuals to be added to the VALID library, which leads to high confidence in the models and reduces the probability that apparent validation issues are a result of modeling errors. The contents of the VALID library are somewhat limited because of the effort involved in generating and reviewing models, so no {sup 233}U-fueled experiments are included in VALID. A large number of models of {sup 233}U experiments have been modeled in the past in support of the criticality safety of {sup 233}U in storage at Oak Ridge National Laboratory (ORNL). In June, 2016, two midshipmen from the United States Naval Academy were assigned to ORNL, and they reviewed and updated these models for execution in SCALE 6.2. The validation of {sup 233}U is of interest because some discrepancies in SCALE results were identified by a beta tester at the Nuclear Energy Agency modeling these systems. These discrepancies were related to a data thinning procedure in the continuous-energy nuclear data process, and had been resolved prior to the release of SCALE 6.2. (authors)

07 ISOTOPE AND RADIATION SOURCES↗