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DOE OSTI2013

System Study: High-Pressure Core Spray 1998–2012

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at 8 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Fukushima Daiichi—A Case Study for BWR Instrumentation and Control Systems Performance During a Severe Accident, Rev 1

This report provides an assessment of the performance of various instrumentation and control (I&C) and monitoring systems, with a focus on the first few days of the Fukushima Daiichi accident starting on March 11, 2011. Similar to what occurred at the Three Mile Island Unit 2 (TMI-2) accident in 1979, operator responses were challenged when information on key parameters was lost, erroneous, or misleading. While numerous improvements in instrumentation capabilities have been made since the TMI-2 event, the accident at Fukushima Daiichi suggests the potential for further improvement. The objective of this research is to review the performance of the Fukushima Daiichi Units 1–3 instrumentation systems that could have helped ensure that plant operators and emergency responders had relevant, accurate, and timely knowledge of plant conditions as they managed the accident response. As was the case with TMI-2, forensic analyses of plant instrumentation failures will be a long-term effort. As an introduction to this long-term research, this report focuses on the information available in the control room during the accident and illustrates the challenges to operators in their understanding of plant conditions during the progression of the accident. I&C systems monitor the success or failure of achievement of light water reactor safety principles in which the protection of the reactor fuel, reactor coolant system, and reactor containment is accomplished. This report (1) provides an overview of the safety principles; (2) notes parameters associated with the boiling water reactor (BWR)/3 and BWR/4 protection, actuation, and monitoring systems as they relate to the achievement of these principles; (3) presents data from past research on predicted instrumentation performance during severe accidents, and (4) assesses performance of the key I&C and monitoring systems associated with BWR/3 and BWR/4 reactors like those at Fukushima Daiichi. Lessons learned or potential improvements associated with the performance of these systems in the context of severe accidents are delineated. As background information, an overview of key BWR/3- and BWR/4-Mark I reactor designs (BWR/3 and BWR/4 with a Mark I containment) reactor protection, engineered safety features actuation, and accident monitoring systems is provided. A brief description of plant safety systems used to prevent or mitigate reactor accidents is also provided. This background information provides context for the brief discussion of previous research that has been conducted on BWR severe accidents and for the discussion of the accident sequences, which led to core damage at Fukushima Daiichi Units 1–3. The progression of the accident sequences at Fukushima Daiichi Units 1–3 occurred very much as predicted in prior severe accident research conducted in the United States and internationally, given the complete station blackout that occurred when the tsunami generated by one of the most powerful earthquakes ever recorded rolled onshore.

Accident Tolerant Equipment↗
DOE OSTI2014

Light Water Reactor Sustainability Program: BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Initiating Event Rates at U.S. Nuclear Power Plants. 1988 - 2013

Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant’s low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC’s Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Risk-Informed Safety Margin Characterization Methods Development Work

This report summarizes the research activity developed during the Fiscal year 2014 within the Risk Informed Safety Margin and Characterization (RISMC) pathway within the Light Water Reactor Sustainability (LWRS) campaign. This research activity is complementary to the one presented in the INL/EXT-??? report which shows advances Probabilistic Risk Assessment Analysis using RAVEN and RELAP-7 in conjunction to novel flooding simulation tools. Here we present several analyses that prove the values of the RISMC approach in order to assess risk associated to nuclear power plants (NPPs). We focus on simulation based PRA which, in contrast to classical PRA, heavily employs system simulator codes. Firstly we compare, these two types of analyses, classical and RISMC, for a Boiling water reactor (BWR) station black out (SBO) initiating event. Secondly we present an extended BWR SBO analysis using RAVEN and RELAP-5 which address the comments and suggestions received about he original analysis presented in INL/EXT-???. This time we focus more on the stochastic analysis such probability of core damage and on the determination of the most risk-relevant factors. We also show some preliminary results regarding the comparison between RELAP5-3D and the new code RELAP-7 for a simplified Pressurized Water Reactors system. Lastly we present some conceptual ideas regarding the possibility to extended the RISMC capabilities from an off-line tool (i.e., as PRA analysis tool) to an online-tool. In this new configuration, RISMC capabilities can be used to assist and inform reactor operator during real accident scenarios.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Thermal-Hydraulic Simulations of Single Pin and Assembly Sector for IVG- 1M Reactor

Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW output power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an inner pin just before the enrichment change. The 600 mm case demonstrated a peak clad surface temperature of 370.4 K, while the 800 mm case had a temperature of 391.6 K. These temperatures are well below the necessary temperatures for boiling to occur at the rated pressure. Fuel temperatures are also well below the melting point. Future bundle work will include simulations of the proposed low-enriched uranium (LEU) design. Two transient scenarios were also investigated for the single-pin geometries. Both were “model” problems that were focused on pure thermal-hydraulic behavior, and as such were simple power changes that did not incorporate neutron kinetics modeling. The first scenario was a high-power, ramp increase, while the second scenario was a low-power, step increase. A cylindrical RELAP model was also constructed to investigate its accuracy as compared to the higher-fidelity CFD. Comparisons between the two codes showed good agreement for peak temperatures in the fuel and at the cladding surface for both cases. In the step transient, temperatures at four axial levels were also computed. These showed greater but reasonable discrepancy, with RELAP outputting higher temperatures. These results provide some evidence that RELAP can be used with confidence in modeling transients for IVG.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Concentration of HLLW from Future SNF Recycling for Efficient Immobilization in a CCIM

Sponsored by the Department of Energy Nuclear Energy’s Fuel Cycle Research and Development Program, the Cold Crucible Induction Melter is being developed as the next generation of melter technology for High Level Liquid Waste’s efficient immobilization in highly durable glass ceramic and ceramic forms. Concentration of the radioactive High Level Liquid Waste generated from the proposed future recycling of spent nuclear fuel, after the fuel’s dissolution in nitric acid, is necessary to take advantage of the inherent attributes of Cold Crucible Induction Melting technology. Based on a provided range of commercial spent nuclear fuel fission product composition data and its expected High Level Liquid Waste raffinate composition data as provided in oxide form, an analysis was completed to concentrate the waste. The analysis involved using nitric acid vapor liquid equilibrium data over a range of boiling temperatures and performing spreadsheet calculations to concentrate the High Level Liquid Waste through evaporation. The calculation results will provide a concentrated nonradioactive surrogate High Level Liquid Waste melter feed recipe for testing in Idaho National Laboratory’s Cold Crucible Induction Melter Pilot Plant. This testing will provide a quantifiable verification of the relatively high feed rates of Cold Crucible Induction Melters compared to those achievable with the current ceramic lined Joule Heated Melters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Irradiation Programs and Test Plans to Assess High-Fluence Irradiation Assisted Stress Corrosion Cracking Susceptibility.

. Irradiation assisted stress corrosion cracking (IASCC) is a known issue in current reactors. In a 60 year lifetime, reactor core internals may experience fluence levels up to 15 dpa for boiling water reactors (BWR) and 100+ dpa for pressurized water reactors (PWR). To support a safe operation of our fleet of reactors and maintain their economic viability it is important to be able to predict any evolution of material behaviors as reactors age and therefore fluence accumulated by reactor core component increases. For PWR reactors, the difficulty to predict high fluence behavior comes from the fact that there is not a consensus of the mechanism of IASCC and that little data is available. It is however possible to use the current state of knowledge on the evolution of irradiated microstructure and on the processes that influences IASCC to emit hypotheses. This report identifies several potential changes in microstructure and proposes to identify their potential impact of IASCC. The susceptibility of a component to high fluence IASCC is considered to not only depends on the intrinsic IASCC susceptibility of the component due to radiation effects on the material but to also be related to the evolution of the loading history of the material and interaction with the environment as total fluence increases. Single variation type experiments are proposed to be performed with materials that are representative of PWR condition and with materials irradiated in other conditions. To address the lack of IASCC propagation and initiation data generated with material irradiated in PWR condition, it is proposed to investigate the effect of spectrum and flux rate on the evolution of microstructure. A long term irradiation, aimed to generate a well-controlled irradiation history on a set on selected materials is also proposed for consideration. For BWR, the study of available data permitted to identify an area of concern for long term performance of component. The efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. The use of existing, available, materials and the generation of additional materials via irradiation in a research reactor are considered.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Methodologies for Reservoir Characterization Using Fluid Inclusion Gas Chemistry

The purpose of this project was to: 1) evaluate the relationship between geothermal fluid processes and the compositions of the fluid inclusion gases trapped in the reservoir rocks; and 2) develop methodologies for interpreting fluid inclusion gas data in terms of the chemical, thermal and hydrological properties of geothermal reservoirs. Phase 1 of this project was designed to conduct the following: 1) model the effects of boiling, condensation, conductive cooling and mixing on selected gaseous species; using fluid compositions obtained from geothermal wells, 2) evaluate, using quantitative analyses provided by New Mexico Tech (NMT), how these processes are recorded by fluid inclusions trapped in individual crystals; and 3) determine if the results obtained on individual crystals can be applied to the bulk fluid inclusion analyses determined by Fluid Inclusion Technology (FIT). Our initial studies however, suggested that numerical modeling of the data would be premature. We observed that the gas compositions, determined on bulk and individual samples were not the same as those discharged by the geothermal wells. Gases discharged from geothermal wells are CO 2 -rich and contain low concentrations of light gases (i.e. H 2 , He, N, Ar, CH4). In contrast many of our samples displayed enrichments in these light gases. Efforts were initiated to evaluate the reasons for the observed gas distributions. As a first step, we examined the potential importance of different reservoir processes using a variety of commonly employed gas ratios (e.g. Giggenbach plots). The second technical target was the development of interpretational methodologies. We have develop methodologies for the interpretation of fluid inclusion gas data, based on the results of Phase 1, geologic interpretation of fluid inclusion data, and integration of the data. These methodologies can be used in conjunction with the relevant geological and hydrological information on the system to create fluid models for the system. The hope is that the methodologies developed will allow bulk fluid inclusion gas analysis to be a useful tool for estimating relative temperatures, identifying the sources and origins of the geothermal fluids, and developing conceptual models that can be used to help target areas of enhanced permeability.

15 GEOTHERMAL ENERGY↗
DOE OSTI2015

Assessment of geothermal resources near Elim, Alaska

The village of Elim, Alaska is 96 miles west of Nome, on the Seward Peninsula. The Darby Mountains north of the village are rich with hydrothermal systems associated with the Darby granitic pluton(s). In addition to the hot springs that have been recorded and studied over the last 100 years, additional hot springs exist. They are known through a rich oral history of the region, though they are not labeled on geothermal maps. This research primarily focused on Kwiniuk Hot Springs, Clear Creek Hot Springs and Molly’s Hot Springs. The highest recorded surface temperatures of these resources exist at Clear Creek Hot Springs (67°C). Repeated water sampling of the resources shows that maximum temperatures at all of the systems are below boiling.

15 GEOTHERMAL ENERGY↗
DOE OSTI2014

Key Parameters for Operator Diagnosis of BWR Plant Condition During a Severe Accident

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems (Rev.2)

This performance specification document addresses use of a standardized transportation, aging, and disposal (STAD) canister system in the spent nuclear fuel (SNF) waste management system. This document was developed to support STAD canister system studies and potential research, development, and demonstration activities. Requirements in this specification may evolve with time based on results from analyses, experiments, design studies, system evaluations and demonstrations, as well as other factors. A STAD canister system consists of a canister, together with a storage or aging overpack/module/vault, a shielded transfer cask (STC), a site transporter, a transportation overpack, and a transportation skid. Three sizes of circular cross section STAD canister variants are specified in this document. The small STAD canister capacity is four pressurized water reactor (PWR) or nine boiling water reactor (BWR) SNF assemblies. The medium STAD canister capacity is 12 PWR or 32 BWR SNF assemblies. The large STAD canister capacity is 21 PWR or 44 BWR SNF assemblies. Using standardized canisters for SNF facilitates the interface between storage at an independent spent fuel storage installation (ISFSI) or a DOEowned interim storage facility (ISF), transportation between facilities, and disposal at a repository. Multiple STAD canisters may be stored in a single storage overpack and/or a single transportation overpack. The US Nuclear Regulatory Commission (NRC) requirements currently used for storage and transportation are adopted for the STAD canister system. Approved approaches (guidance) in NRC storage and transportation documents are also adopted as requirements for the STAD canister system to facilitate obtaining NRC Certificates of Compliance (CoCs) for STAD canister systems. Performance specifications in this document are also informed by STAD canister disposability in various geologic media and consideration of ISF (and geologic repository) handling facility design, operations, and licensing requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Rationale for the Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems (Rev.1)

This document provides the rationale for the specifications for the standardized transportation, aging, and disposal (STAD) canister systems provided in Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, FCRD-NFST-2014-000579, Rev. 2, referred to as STAD Spec herein. The STAD Spec was developed to support STAD canister system studies and potential research, development, and demonstration activities. Requirements in the specification may evolve with time based on results from analyses, experiments, design studies, system evaluations and demonstrations, as well as other factors. The STAD canister system consists of a canister, together with a storage or aging overpack/module/vault, a shielded transfer cask (STC), a site transporter, a transportation overpack, and a transportation skid. Three sizes of circular cross section STAD canister variants are specified. The small STAD canister capacity is four pressurized water reactor (PWR) or nine boiling water reactor (BWR) spent nuclear fuel (SNF) assemblies. The medium STAD canister capacity is 12 PWR or 32 BWR SNF assemblies. The large STAD canister capacity is 21 PWR or 44 BWR SNF assemblies. There are 60 specifications in the STAD Spec that cover the STAD canister general design attributes, canister safety functional requirements, operational considerations, materials, storage and aging system requirements, and transportation system requirements. The 60 specifications are derived from regulatory requirements, regulatory guidance, STAD canister disposability considerations, and interim storage facility (ISF) and geologic repository handling facility considerations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs). Researchers have investigated the effects of thermal conductivity on design basis accidents, investigated silicon carbide (SiC) cladding, as well as the effects of ATF concepts on the late stage accident progression. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 and the South Korean OPR-1000 pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl). With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6, and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Extensions to Dynamic System Simulation of Fissile Solution Systems

Previous reports have documented the results of applying dynamic system simulation (DSS) techniques to model a variety of fissile solution systems. The SUPO (Super Power) aqueous homogeneous reactor (AHR) was chosen as the benchmark for comparison of model results to experimental data for steadystate operation.1 Subsequently, DSS was applied to additional AHR to verify results obtained for SUPO and extend modeling to prompt critical excursions, ramp reactivity insertions of various magnitudes and rate, and boiling operations in SILENE and KEWB (Kinetic Experiment Water Boiler).2 Additional models for pressurized cores (HRE: Homogeneous Reactor Experiment), annular core geometries, and accelerator-driven subcritical systems (ADAHR) were developed and results reported.3 The focus of each of these models is core dynamics; neutron kinetics, thermal hydraulics, radiolytic gas generation and transport are coupled to examine the time-based evolution of these systems from start-up through transition to steady-state. A common characteristic of these models is the assumption that (a) core cooling system inlet temperature and flow and (b) plenum gas inlet pressure and flow are held constant; no external (to core) component operations that may result in dynamic change to these parameters are considered. This report discusses extension of models to include explicit reference to cooling structures and radiolytic gas handling. The accelerator-driven subcritical generic system model described in References 3 and 4 is used as a basis for this extension.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Advanced probabilistic risk analysis using RAVEN and RELAP-7

RAVEN, under the support of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program [1], is advancing its capability to perform statistical analyses of stochastic dynamic systems. This is aligned with its mission to provide the tools needed by the Risk Informed Safety Margin Characterization (RISMC) path-lead [2] under the Department Of Energy (DOE) Light Water Reactor Sustainability program [3]. In particular this task is focused on the synergetic development with the RELAP-7 [4] code to advance the state of the art on the safety analysis of nuclear power plants (NPP). The investigation of the probabilistic evolution of accident scenarios for a complex system such as a nuclear power plant is not a trivial challenge. The complexity of the system to be modeled leads to demanding computational requirements even to simulate one of the many possible evolutions of an accident scenario (tens of CPU/hour). At the same time, the probabilistic analysis requires thousands of runs to investigate outcomes characterized by low probability and severe consequence (tail problem). The milestone reported in June of 2013 [5] described the capability of RAVEN to implement complex control logic and provide an adequate support for the exploration of the probabilistic space using a Monte Carlo sampling strategy. Unfortunately the Monte Carlo approach is ineffective with a problem of this complexity. In the following year of development, the RAVEN code has been extended with more sophisticated sampling strategies (grids, Latin Hypercube, and adaptive sampling). This milestone report illustrates the effectiveness of those methodologies in performing the assessment of the probability of core damage following the onset of a Station Black Out (SBO) situation in a boiling water reactor (BWR). The first part of the report provides an overview of the available probabilistic analysis capabilities, ranging from the different types of distributions available, possible sampling strategies, and post processing analysis capabilities. The first part of the report provides an extensive description of two major developments introduced this year: adaptive sampling for limit surface sampling and multi variate distributions. The document concludes with a description of the demo case (BWR-SBO) and a discussion of the results obtained.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

ASME Code Considerations for the Compact Heat Exchanger

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a working fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗