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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Stochastic Modeling of Radioactive Material Releases

Nonreactor nuclear facilities operated under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculates the radiation dose associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA was developed using the MATLAB coding framework. The software application has a graphical user input. SODA can be installed on both Windows and Mac computers and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC, rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The work was funded through a grant from the DOE Nuclear Safety Research and Development Program.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

NSR&D Program Fiscal Year 2015 Funded Research Stochastic Modeling of Radioactive Material Releases Final Report

Nonreactor nuclear facilities operating under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculates the radiation dose distribution associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. Users can also specify custom distributions through a user defined distribution option. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA, developed using the MATLAB coding framework, has a graphical user interface and can be installed on both Windows and Mac computers. SODA is a standalone software application and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC; rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The SODA development project was funded through a grant from the DOE Nuclear Safety Research and Development Program.

42 ENGINEERING↗
DOE OSTI2026

Reassessing Double-Ended Guillotine Break Requirements: Evidence-Based Analysis of Regulatory Assumptions After Five Decades of Nuclear Operation

After five decades of nuclear power operation encompassing more than 20,000 reactor-years across 35 countries and 647 reactors, zero double-ended guillotine breaks (DEGBs) have been documented in commercial reactor coolant systems—despite DEGB being the fundamental design-basis assumption driving Emergency Core Cooling System (ECCS) sizing, structural protection requirements, and containment design specifications. This report examines the basis for DEGB requirements in nuclear power plant design. The DEGB postulate assumes the instantaneous, complete circumferential severance of the largest diameter pipes in reactor coolant systems, driving major design requirements under 10 Code of Federal Regulations 50.46, General Design Criterion 4 and containment design specifications. The United States (4,880 reactor-years) and France (2,505 reactor-years) contribute the largest operational datasets. Probabilistic assessments estimate direct DEGB occurrence probabilities with extremely low event frequencies, far below the 10-5/reactor-year thresholds typically used to define non-credible events in nuclear-safety analyses; i.e., events with probability this low fall into beyond-design-basis events. Current material-science knowledge demonstrates that the ductile steel materials used in nuclear piping systems exhibit stable crack-growth behavior fundamentally incompatible with instantaneous severance. International regulatory experience, particularly Germany’s comprehensive break-preclusion implementation, and successful leak-before-break (LBB) applications in almost all of U.S. pressurized water reactor units validate that alternatives can maintain safety performance while reducing economic burden. Current DEGB protection systems impose estimated lifetime costs of hundreds of millions of dollars per unit, over the life of a plant across the nuclear industry (including ongoing costs), representing substantial resource allocation toward scenarios with extremely low probability. Although this report acknowledges uncertainties regarding long-term aging effects, potential synergistic degradation mechanisms, and site-specific seismic considerations that warrant continued evaluation as regulatory policy evolves, there remains no documented evidence that a DEGB has occurred as a consequence of the conditions or mechanisms described in this report. This report acknowledges the Nuclear Regulatory Commission’s (NRC’s) recent efforts—outlined in the draft Interim Staff Guidance (ISG) NRC-DSS-ISG-2025-XX (“Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance”)—to reduce overly conservative requirements for large-break loss of coolant accidents through technical justifications and exemptions. However, extensive operating experience and validated methodologies—such as LBB and in-service inspection programs—demonstrate that the probability of a DEGB in reactor coolant-loop piping is extremely low, even under seismic conditions. The authors and reviewers of this report recommend that DEGB be removed as a design-basis event through formal rulemaking, rather than case-by-case exemptions, to better reflect credible failure modes, align with current data, and align with modern, risk-informed safety analysis.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

An Approach for Assessing Consequences of Potential Supply Chain and Insider Contributed Cyber Attacks on Nuclear Power Plants

This paper provides an approach for developing potential attacks on I and C systems of NPPs and assessing their consequences. An important concept is that the NPPs were not designed to cope with Stuxnet-type of attacks (and any other cyber attacks). That is, the plants were only designed for design basis accidents. The safety margins and redundancies built in the design are all based on design basis accidents. They may be helpful in mitigating cyberattacks, but may not be adequate.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Near-Tank Treatment of Hanford Tank Waste: Pilot-Scale Testing - 12107

In order to reduce the number of high level waste canisters that will have to be produced by the Hanford Tank Waste Treatment and Immobilization Plant, supplemental waste treatment technologies are being investigated. One such technology is the Near-Tank Treatment System, which uses continuous sludge leaching, filtration, and ion exchange to process the waste in a simple, compact system. The Near Tank Treatment System is ideally suited for handling Hanford tanks with large amounts of boehmite, a difficult aluminum phase to dissolve. A pilot-scale Near Tank Treatment System was constructed and tested with a boehmite and iron oxyhydroxide waste simulant to evaluate the robustness and effectiveness of the system. The data from the pilot-scale tests were also used to assess scale-up from previously performed bench-scale tests. It should be noted that any work involving the use of simulated HLW has inherent limitations. For this work, every attempt was made to ensure that the simulant mimicked the actual waste performance as closely as possible. However, there are always limitations in the ability to The NTTS Demonstration Test established that all the subsystems could be operated concurrently to process the waste simulant. The NTTS had a stable performance during the Demonstration Test that required very little external manipulation once a steady state was achieved. The CSL demonstrated effectively 90% dissolution of the hard to leach boehmite phase. The cross flow filtration system was able to deliver decontaminated salt solution to the NTCR system. The NTCR system exceeded the design basis by providing effluent below the detection limit beyond the design basis required time. The NTTS system was successfully demonstrated and has shown that leaching and decontamination of the subsequent permeate can be readily achieved using relatively small footprint equipment in an at tank application. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Metal Matrix Microencapsulated (M3) fuel neutronics performance in PWRs

Metal Matrix Microencapsulated (M3) fuel consists of TRISO or BISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod (Fig. 1). In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated. In this manner pellet-clad-interactions (PCI), thin tube failure due to oxidation and hydriding, and tube pressurization and burst will be absent. M3 fuel, given the high stiffness of the integral rod design, could as well improve grid-to-rod wear behavior. Overall M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. Multiple barriers to fission product release (ceramic coating layers in the coated fuel particle and te metal matrix) and the high thermal conductivity zirconium alloy metal matrix contribute to the enhancement in fuel behavior. The discontinuous nature of fissile material encapsulated in coated particles provides additional assistance; for instance if the M3 fuel rod is snapped into multiple pieces, only the limited number of fuel particles at the failure cross section are susceptible to release fission products. This is in contrast to the conventional oxide fuel where the presence of a small opening in the cladding provides the pathway for release of the entire inventory of fission products from the fuel rod. While conventional metal fuels (e.g. U-Zr and U-Mo) are typically expected to experience large swelling under irradiation due to the high degree of damage from fission fragments and introduction of fission gas into the lattice, this is not the case for M3 fuels. The fissile portion of the fuel is contained within the coated particle where enough room is available to accommodate fission gases and kernel swelling. The zirconium metal matrix will not be exposed to fission products and its swelling is known to be very limited when exposed solely to neutrons. Under design basis RIA and LOCA, fuel performance will be superior to the conventional oxide fuel since PCMI and rod burst, respectively, have been mitigated. Under beyond design basis accident scenarios where the fuel is exposed to high temperature steam for prolonged periods, larger inventory of zirconium metal in the core could negatively affect the accident progression. A thin steam resistant layer (e.g. alumina forming alloy steel), integrated into the solid rod during fabrication by co-extrusion or hot-isostatic-pressing, offers the potential to provide additional fuel protection from steam interaction: blanketing under a range of boiling regimes and under severe accident conditions up to high temperatures well beyond what is currently possible in the conventional fuel. A crucial aspect to the viability of M3 fuel in light water reactors is the reduced heavy metal load compared to standard pellet fuel. This study evaluated the design requirements to operate a Pressurized Water Reactor (PWR) with M3 fuel in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors comparable to that of standard fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Technical Basis for Design Life Extension of the SAVY-4000 Series Nuclear Material Packaging System from Five Years to Fifteen Years

The DOE Manual, M441.1-1, Nuclear Material Packaging, approved March 7, 2008, provides detailed packaging requirements for protecting workers from exposure to nuclear materials stored outside of an approved engineered contamination barrier. The SAVY-4000 Series nuclear material packaging system is a DOE Manual 441.1-1 compliant storage system used at Los Alamos National Laboratory. The purpose of this report is to document the technical basis and DOE approval for the initial 5-year design life of the SAVY-4000 Series and to summarize, with supporting detailed technical reports as references, the technical basis for extending the design life from 5 years to 15 years.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Heavy metal inventory and fuel sustainability of recycling TRU in FBR design

Nuclear fuel materials from spent fuel of light water reactors have a potential to be used for destructive devices with very huge energy release or in the same time, it can be utilized as a peaceful energy or civil applications, for generating electricity, desalination of water, medical application and others applications. Several research activities showed some recycled spent fuel can be used as additional fuel loading for increasing fuel breeding capability as well as improving intrinsic aspect of nuclear non-proliferation. The present investigation intends to evaluate the composition of heavy metals inventories and fuel breeding capability in the FBR design based on the loaded fuel of light water reactor (LWR) spent fuel (SF) of 33 GWd/t with 5 years cooling time by adopting depletion code of ORIGEN. Whole core analysis of FBR design is performed by adopting and coupling codes such as SLAROM code, JOINT and CITATION codes. Nuclear data library, JFS-3-J-3.2R which is based on the JENDL 3.2 has been used for nuclear data analysis. JSFR design is the basis design reference which basically adopted 800 days cycle length for 4 batches system. Higher inventories of plutonium of MOX fuel and TRU fuel types at equilibrium composition than initial composition have been shown. Minor actinide (MA) inventory compositions obtain a different inventory trends at equilibrium composition for both fuel types. Higher Inventory of MA is obtained by MOX fuel and less MA inventory for TRU fuel at equilibrium composition than initial composition. Some different MA inventories can be estimated from the different inventory trend of americium (Am). Higher americium inventory for MOX fuel and less americium inventory for TRU fuel at equilibrium condition. Breeding ratio of TRU fuel is relatively higher compared with MOX fuel type. It can be estimated from relatively higher production of Pu-238 (through converted MA) in TRU fuel, and Pu-238 converts through neutron capture to produce Pu-239. Higher breeding ratio of MOX fuel and TRU fuel types at equilibrium condition are estimated from converted fertile material during reactor operation into fissile material of plutonium such as converted uranium fuel (converted U-238 into Pu-239) or additional converted fuel from MA into Pu-238 and changes into Pu-239 by capturing neutron. Loading LWR SF gives better fuel breeding capability and increase inventory of MA for doping material of MOX fuel; however, it requires more supply MA inventory for TRU fuel type.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Sensitivity Analysis of Passive Heat Removal Capability of FHR Based on Efficient Surrogate Model and GenFlow

Code Scaling, Applicability and Uncertainty Evaluation Methodology (CSAU) is used for quantifying reactor safety margins and works as an alternative method for licensing nuclear power plant. In the framework of CSAU, sensitivity and uncertainty analysis (SA and UA) should be performed. However, for large-scale complicated thermal hydraulics analysis simulated by code (e.g. RELAP5, CFD and subchannel analysis, the sensitivity and uncertainty analysis is really time-consuming. Furthermore, sometimes it is not feasible to perform SA and UA. Specifically, the computation time for Monte Carlo sampling or advanced LHS sampling can be neglected, but the computation time for thermal hydraulics code makes up main part of time during SA and UA process. In order to accelerate the speed of SA and UA, thermal hydraulics code can be improved based on advanced mathematic algorithm or using parallel method. However, the development for new mathematic method needs long time and strict verification and validation process is needed for a new code, which usually takes a couple of years or even longer time to have an approved code that can be used in nuclear industry. An alternative way to improve the speed of SA and UA is using surrogate model. The surrogate model is derived from the input as well as output data of the original code, assuming the code as a black box. The most common method is using response surface method (RSM) as a surrogate for the code, the input parameters are then sampled from their assumed distribution and the corresponding value of the response surface function is calculated. However, a couple of input parameters need hundreds or even thousands of original code calculations. The large number of runs is an obvious disadvantage of this method. Besides, the second disadvantage is an appropriate response surface function must be selected (e.g. quadratic, cubic, quartic, with or without cross product terms, or using a non-polynomial form) to avoid the overfit and underfit problem. This process is time consuming and expensive. Polynomial chaos expansion method is a reasonably fast and attractive alternative, stochastic model. The chaos expansion provides an efficient high-order method of including non-linear effects in stochastic analysis. Usually, the information about the model parameters and their probability distribution is limited and the dependence between them is not known. However, if arbitrary polynomial chaos is used, instead of assuming probability distribution for input parameters which may introduces additional subjectivity into the analysis, the available information can directly and most purely be used in stochastic analysis based on arbitrary data distribution. Thus, it is not necessary to know the exact probability density functions to perform SA and UA based on arbitrary polynomial chaos (aPC). The aPC has been applied for uncertainty quantification in many fields (e.g. in solid mechanics, fluid flows, thermal sciences), and become widely used in nuclear industry recently. Mk1 PB-FHR is a small modular fluoride-salt-cooled, high-temperature reactors (FHRs) that is a concept design by University of California, Berkeley (UCB). It has inherently safe features like natural-circulation-driven systems (e.g. direct reactor auxiliary cooling system (DRACS)) to passively remove decay heat from the reactor core during design basis and beyond design. It is necessary to perform SA and UA, since the licensing process of this new kind of reactor is under way and there are many design parameters that have not been fixed. Current nuclear commercial codes, such as RELAP5, have mainly been developed for light water reactors, which require the use of comprehensive two-phase flow models. This capability is not required to simulate single-phase flow phenomena in FHRs, whose baseline coolant, flibe, remains liquid up to 1430 deg. C at atmospheric pressure. The FHR advanced natural circulation analysis (FANCY) code has been developed specifically for FHR technology modeling and validation work has been performed by using CIET data in UCB. The code has been renamed General Flow (GenFlow) because it has been further developed to have the capability to model complicated pipework system based on object-oriented concept for user friendly purpose. Besides, it can be used to simulate both the force circulation as well as natural circulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A summary of the Hanford single-shell tank structural analysis of record - 15526

Washington River Protection Solutions, LLC (WRPS), the US DOE Hanford Site tank contractor, developed an enhanced Single-Shell Tank Integrity Project (SSTIP) in 2009 to better understand the integrity of the single-shell tanks (SSTs). A panel of experts from industry and academia was created to recommend project activities in four main areas: structural integrity, liner degradation, leak integrity and prevention, and contamination migration. One of the panel's primary recommendations was to perform a detailed structural integrity analysis of the SSTs. WRPS and the U.S. Department of Energy, Office of River Protection commissioned the Pacific Northwest National Laboratory (PNNL) to perform this structural integrity analysis of record for the Hanford single shell tanks. This is one of seventeen panel recommendations that form the M-45-10-1 Change Package for the Hanford Federal Agreement and Compliance Order; also known as the Tri-Party Agreement. Finite element analysis was used to predict the structural response of the SSTs to the historical thermal and operating loads, plus design basis seismic loads. Bounding thermal histories were established from waste temperature records and applied to the model to include the thermal degradation of concrete modulus and strength, plus cracking due to differential thermal expansion while under in situ static loading conditions. PNNL contracted with Becht Engineering to evaluate the dynamic response of the tanks to a design basis earthquake. The degraded stiffness of the reinforced concrete was also incorporated in the seismic analysis. The combined response to static and seismic loads was then evaluated against the design requirements of the American Concrete Institute standard, ACI-349-06, for nuclear safety-related concrete structures. Specific analyses were performed for the four major tank designs.This paper presents an overview of the analysis methods used and it summarizes the major analytical findings that are common to the tanks and others that are specific to the individual tank designs. The analyses of record determined that including the bounding effects of their operating histories, each tank design currently satisfies the ACI-349-06 structural integrity design requirements. Although leak tightness is not directly addressed in the structural integrity analysis, this analysis will be a key input in the formal tank integrity assessment due in 2018. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Energetic Basis and Design of Enzyme Function Demonstrated Using GFP, an Excited-State Enzyme

We report the past decades have witnessed an explosion of de novo protein designs with a remarkable range of scaffolds. It remains challenging, however, to design catalytic functions that are competitive with naturally occurring counterparts as well as biomimetic or nonbiological catalysts. Although directed evolution often offers efficient solutions, the fitness landscape remains opaque. Green fluorescent protein (GFP), which has revolutionized biological imaging and assays, is one of the most redesigned proteins. While not an enzyme in the conventional sense, GFPs feature competing excited-state decay pathways with the same steric and electrostatic origins as conventional ground-state catalysts, and they exert exquisite control over multiple reaction outcomes through the same principles. Thus, GFP is an “excited-state enzyme”. Herein we show that rationally designed mutants and hybrids that contain environmental mutations and substituted chromophores provide the basis for a quantitative model and prediction that describes the influence of sterics and electrostatics on excited-state catalysis of GFPs. As both perturbations can selectively bias photoisomerization pathways, GFPs with fluorescence quantum yields (FQYs) and photoswitching characteristics tailored for specific applications could be predicted and then demonstrated. The underlying energetic landscape, readily accessible via spectroscopy for GFPs, offers an important missing link in the design of protein function that is generalizable to catalyst design.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2010

Application of the DG-1199 methodology to the ESBWR and ABWR.

Appendix A-5 of Draft Regulatory Guide DG-1199 'Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors' provides guidance - applicable to RADTRAD MSIV leakage models - for scaling containment aerosol concentration to the expected steam dome concentration in order to preserve the simplified use of the Accident Source Term (AST) in assessing containment performance under assumed design basis accident (DBA) conditions. In this study Economic and Safe Boiling Water Reactor (ESBWR) and Advanced Boiling Water Reactor (ABWR) RADTRAD models are developed using the DG-1199, Appendix A-5 guidance. The models were run using RADTRAD v3.03. Low Population Zone (LPZ), control room (CR), and worst-case 2-hr Exclusion Area Boundary (EAB) doses were calculated and compared to the relevant accident dose criteria in 10 CFR 50.67. For the ESBWR, the dose results were all lower than the MSIV leakage doses calculated by General Electric/Hitachi (GEH) in their licensing technical report. There are no comparable ABWR MSIV leakage doses, however, it should be noted that the ABWR doses are lower than the ESBWR doses. In addition, sensitivity cases were evaluated to ascertain the influence/importance of key input parameters/features of the models.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2012

Design study on safety protection system of JSFR

Development of Japan Sodium-cooled Fast Reactor (JSFR) has been progressed in Fast Reactor Cycle Technology Development (FaCT) project aiming at realizing high level of safety, reliability and economic competitiveness. For JSFR, design consideration on safety protection system has also been performed, which is essential for reactor shutdown in the case of design basis events (DBEs). In the design activity, consideration of safety protection system includes logic circuits configuration, selection of trip signals, and its setting values for reactor trip. In addition, it is necessary to evaluate the performance of the safety protection system by safety analysis taking into account the comprehensive parameter ranges. For this purpose, it has been evaluated whether adequate reactor trip signals can be ensured for satisfying safety standard regarding the fuel integrity (e.g., maximum fuel clad temperature) for DBEs. In this paper, results obtained from the design study on safety protection system of JSFR is presented focusing on the evaluation results of satisfaction of safety protection system for representative events of transient over power (TOP), loss of coolant flow (LOF) and loss of heat sink (LOHS). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

NEUP Final Report: Multilayer Composite Fuel Cladding and Core Internals for LWR Performance Enhancement and Severe Accident Tolerance

Accident tolerance of LWR fuels and structures is of paramount importance, as highlighted by the accident at the Fukushima Daiichi nuclear power station. The ability of fuel cladding and core internals to resist runaway oxidation during a beyond design basis accident (BDBA), as well as to minimize corrosion during steady state operation and design basis accidents (DBAs), determines its degree of accident tolerance. A gap was identified in the accident tolerant fuel (ATF) concepts under consideration, as most consisted either of coatings which may not hold their integrity under prolonged operation, or different materials altogether which incur significant neutronic penalties. A compromise was therefore sought – by minimizing the amount of coating material while ensuring a strong microstructural bond, it was thought that a Zircaloy-steel layered composite would achieve the best of both approaches.The goal of this NEUP project was to develop a multi-metallic layered composite (MMLC) tailored to addressing accident tolerance of LWR fuel cladding and core internal structures using an innovative fabrication technology. The MMLC developed in this program is expected to enhance the accident tolerance of LWRs, thereby reducing the cost by recovering lost operating margins and/or increasing operating windows of peak cladding temperature, peak linear power, reducing steady-state corrosion, and enhancing severe accident tolerance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Performance of Waterlogged LWR fuel during RIA

Fuel pins with small defects could allow for the gradual ingress of water into the cladding during normal operation. If these 'waterlogged' fuel pins are subjected to a rapid power change, the internal pressurization could greatly exceed that of a normal fuel pin due to rapid conversion of the water to steam. This could induce either a premature or a more energetic failure during a design basis RIA than in non-defected fuel pins. Experimental studies on this type of fuel are very limited and do not appear to have been fully reported in the open literature. Results from a variety of tests conducted at the Transient Reactor Test Facility (TREAT), Special Power Excursion Reactor Test (SPERT) facility, Nuclear Safety Research Reactor (NSRR), and Power Burst Facility (PBF) are reviewed in this paper. These tests were intended to evaluate both the physical consequences of failure and the threshold and causes for waterlogged fuel failure during RIA. The available data suggests that although waterlogged fuel pins may burst at a lower energy deposition threshold, they do not pose a significant new hazard during design basis accidents. However, data is limited to tests conducted on fresh fuel samples only.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Recent Progress on the Stochastic Objective Decision Aide (SODA) Application

Idaho State University, in collaboration with Idaho National Laboratory, is engaged in development of a simple to use software application that utilizes stochastic techniques for the calculation of radiation doses associated with hypothetical radioactive material releases. This work is being performed through a grant from the U.S. Department of Energy (DOE) Nuclear Safety Research and Development (NSR and D) Program. Initial progress on the application was reported previously. Recent additions to the application are reported here. The Stochastic Objective Decision Aide (SODA) provides an easy to use software application to help decision makers support their judgment about the need or lack of need for safety structures, systems, and components. SODA is not intended to replace traditional design basis accident dose calculation methodology, nor is it intended to replace existing software used to perform formal dose consequence calculations. Rather, SODA is intended to supplement information obtained from existing codes to give a decision maker more information about an accident scenario. SODA also provides a simple method to perform parametric or sensitivity studies on individual input parameters. Non-reactor nuclear facilities operated under the approval authority of the US DOE use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge dose evaluation guidelines. Unmitigated events that sufficiently challenge dose evaluation guidelines merit the selection of safety structures, systems, or components (SSCs) to prevent or mitigate the hazard. Dose consequence calculations traditionally involve use of the 'five-factor' formula: ST = MAR.DR.ARF.RF.LPF where ST is the source term (Bq), MAR is the total available material-at-risk (Bq), DR is the damage ratio (no units), ARF is the airborne release fraction (no units), RF is the respirable fraction (no units), and LPF is the leak path factor (no units). Potential radiation doses are then calculated using: CED = χ/Q.BR.ST.DCF (2) where CED is the committed effective dose (Sv), χ/Q is the plume dispersion (s/m{sup 3}), BR is the breathing rate (m{sup 3}/s), ST is the source term (Bq), and DCF is the dose conversion factor (Sv/Bq). Conservative single value input parameters are typically used to represent ARF, RF, LPF and BR. The traditional methodology, while conservative, can lead to skewed conclusions in the balance between cost and risk reduction resulting in over engineered systems with greater design, construction and operating costs. Rather than using a bounding single point value for each parameter in the dose consequence calculation, SODA uses distributions for some or all of the parameters. Each input parameter distribution is stochastically sampled and the resulting dose consequence calculation is repeated many times to develop a dose consequence distribution. The resulting distribution provides more information for the decision maker to rely upon when determining the need for safety SSCs. The SODA application was developed using MATLAB and it incorporates the use of Monte Carlo techniques as well as a graphical user interface (GUI). The application includes user selection of the governing distribution for parameters such as the MAR, DR, ARF, RF, LPF,χ/Q, BR, and DCF. While MATLAB was used for the development work, the application is distributed as a self-contained executable program. Features of the application include pull down menus with available distributions for the various parameters. The user has the option to plot the input parameter distribution that results from the random sampling process. The resulting dose consequence distribution is automatically analyzed to determine its similarity to well know distributions. This was done using the Bayesian Information Criterion method. If the resulting distribution is sufficiently similar to a well-known distribution, the application provides information about the distributions such as the mean value and other parameters. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

NUHOMS{sup R} EOS (Extended Optimized Storage): Robust, Advanced, and Dual Purpose Dry Storage System - 16472

AREVA TN has completed the design, analysis, and US NRC license submittal for the NUHOMS{sup R} Extended Optimized Storage (EOS) system. NUHOMS{sup R} EOS offers a large capacity and high heat load dry storage and transportation solution for UNF. EOS is not a new system, but an evolution of the already-proven safe and low-risk horizontal NUHOMS{sup R} storage system, which is one of the most widely used UNF dry storage systems in the United States. The EOS system features a 37PTH dry shielded canister (DSC) for 37 PWR UNF assemblies and an 89BTH DSC for the 89 BWR UNF assemblies. The EOS DSCs will be stored inside a horizontal storage module (HSM) that is a robust, fortress-like concrete storage overpack. Designed with the lessons learned from the Fukushima event in Japan in mind, EOS has robust margins for beyond design basis (BDB) events and offers the lowest risk and highest safety margin of any dry storage system design for recovery from BDB events. EOS also offers innovative aging management features with an aging management program (AMP) to assure safe and low dose long-term storage at the utility sites. EOS incorporates ports for AMP inspections and offers the option of using duplex stainless steel (DSS) for the DSC shell in order to negate corrosion issues. EOS is stored horizontally and above ground, making it easy and quick to access for visual inspection, monitoring, and mitigating actions during AMP. NUHOMS{sup R} EOS is optimized for high capacity, high burnup, high enrichments, and very low cooling times for UNF assemblies. It enables the highest heat load per canister (50 kW) and low dose rates. Even with high heat loads, EOS has significant margins to design basis limits. As the future of used fuel disposition will require the need to move fuel off-site to interim or consolidated storage, EOS is the ideal solution as it enables on-site loading and unloading operations that mitigate risk when transferring fuel on-site or transporting fuel off-site. The EOS design can be safely located in a traditional outdoor setting with fortress-like concrete reinforced modules that offer superior resistance to attack or can be safely operated inside of a building enclosure for those utilities desiring a more discreet siting for its used fuel storage systems. EOS has an innovative basket design that is made of high strength low alloy (HSLA) material and is easy to fabricate due to its non-welded design. Its innovative inlet and outlet air vent design, which incorporates dose reduction hardware, allows for optimal ambient airflow for highly efficient heat transfer with minimal radiation dose. The NUHOMS{sup R} EOS storage modules are interlocked in a side-by-side configuration offering superior and robust seismic performance. The license application for NUHOMS{sup R} EOS is currently under US NRC review and this system will be available for deployment in 2017. (authors)

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