Bess, John D | Engineering Papers

DOE OSTI2020

Availability of Shielding Benchmark Experiment Data in the ICSBEP Handbook

The International Criticality Safety Benchmark Evaluation Projects (ICSBEP) has over two decades of experience under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) to identify, evaluate, and compile a comprehensive data set of criticality safety benchmark data [1]. The activities of the ICSBEP have led to the continued success of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) [2]. The contents of the ICSBEP Handbook include benchmark specifications derived from nuclear facilities around the world, which have been utilized to support criticality safety efforts and also integral testing of nuclear data [3]. The 2020 edition of the ICSBEP Handbook will include data from 582 evaluations containing benchmark specifications for 5,053 critical, subcritical, or near-critical configurations. There are seven criticality-alarm-placement/shielding evaluations containing a total of 45 benchmark configurations, and ten fundamental physics evaluations containing a total of 237 measurements relevant to criticality safety applications. The purpose of this paper is to summarize those benchmarks relative to validation needs in shielding applications currently found within the ICSBEP Handbook. The current and future plans of shielding benchmarks with regards to the ICSBEP will also be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Establishing Maximal Core Excess Reactivity Envelope for TREAT Test Vehicle Insertion

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. The following work represents further analyses building upon previously reported work to support larger volume test vehicles within TREAT [2]. The Broad Use Specimen Transient Experiment Rig (BUSTER) provides ample versatility for numerous transient test concepts. However, future testing and experiment needs will require testing environments of different in-core footprints. The versatility of TREAT allows for the development of variegated core layouts with key limitations emplaced as operational safety temperature restrictions and available core reactivity. The focus of this current effort is to gauge core arrangements for TREAT core loadings to support centralized experimental testing of varying dimensional capacity. The total core excess reactivity (ER) was computed for different arrangements to establish comparison tables that enable ready assessment of core ER available for future experiment concepts. Test vehicle worths would then be necessary to establish available ER to support transient testing requirements and to perform reactor performance and safety calculations [3,4].

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

The 2019 Edition of the ICSBEP Handbook

Over the past several decades, innumerable experiments have been performed worldwide to support criticality safety. These experiments represent significant investment in infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods and nuclear data [1]. If the experimental data are lost, the high cost to repeat many of these measurements would likely be prohibitive. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to: 1. Identify a comprehensive set of critical benchmark data and, to the extent possible, verify the data; 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analyses; 3. Compile the data into a standardized format; 4. Perform calculations of each experiment with standard criticality safety codes; and 5. Formally document the work into a single source of verified benchmark critical data. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) [2]. Further information regarding the ICSBEP, including highlighted criticality safety benchmark evaluations, can be found in the September and October 2003 editions of Nuclear Science and Engineering [3]. New and recently revised evaluations have been prepared for inclusion in the 2019 edition of the ICSBEP Handbook.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

The 2019 Edition of the IRPhEP Handbook

The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] was established to preserve and evaluate integral reactor physics experiment data, including separate or special effects measurements for nuclear energy and technology applications. Numerous international experiments have been performed over the past several decades to support reactor operations, safety, measurements, and design. Those experiments represent a significant investment of infrastructure, cost, and expertise that, in many instances, cannot be as easily reproduced. They represent valuable resources of data for present and future research, and provide the basis for recording, development, and validation of methods [2]. The primary focus of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics integral data suitable for reactor designers and safety analysts to not just validate the analytical tools necessary to design next-generation nuclear reactors, but to establish the safety basis for their operation. Contributors worldwide collaborate to evaluate and review selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [3]. More detailed discussion of the IRPhEP, including highlighted reactor physics benchmark evaluations, can be found in the November and December 2014 editions of Nuclear Science and Engineering [1]. Several new evaluations, and revisions to existing benchmarks have been prepared for inclusion in the 2019 edition of the IRPhEP Handbook (see Table I).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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.

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