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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

FAST-1.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: • heat conduction through the fuel and other materials • heat transfer from the cladding-to-coolant • cladding elastic and plastic deformation, including creep • fuel-cladding mechanical interaction • fission gas release from the fuel • rod internal pressure and void volume • cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat-transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Compact System for Operating and Decommissioned Reactor Used Fuel Management - 17243

Most nuclear power plants in operation today were designed with limited wet storage for their discharged used nuclear fuel (UNF), predicated on the assumption that either recycling or geologic storage would be available within the reactor's lifetime. While that has proven to be the case in some countries such as France with its recycling, it has not for most. In the USA, dry storage of UNF at the plant site at an Independent Spent Fuel Storage Installation (ISFSI) has since the mid-1980s been the primary mode of interim storage so that the plant can off-load its UNF from the pools at the power plants. Initially the space reserved for the ISFSI at a reactor site was based on the assumption that the Department of Energy (DOE) would be transporting that UNF to a geological repository site, such as Yucca Mountain, by 1998. However, due to the delay in the opening of any geological repository the utilities are running out of space on their existing ISFSIs for additional UNF dry storage systems. Since plants prefer to keep their ISFSIs inside the protected area of the plant, options for expanding the ISFSI are becoming limited. At the same time, increasing the capacity of the dry storage systems is also not an option in most cases as the dry storage systems have almost reached their capacity limits with the 37 PWR or 89 BWR fuel assembly canister designs. For plants that have announced plans to shut down, their main priority is to off-load the UNF from the pool to dry storage at an ISFSI so that they can proceed to 'Safe store' or other storage options, and start their decommissioning process as soon as possible. Thus, keeping the size of the ISFSI compact is very important for safety, security and economics. This paper will describe the details of a reduced footprint version of AREVA TN's NUHOMS{sup R} System that addresses the need for maximizing ISFSI capacity. This system not only offers a compact design that stores nearly double the number of UNF assemblies in the same space, it also offers superior radiation protection to the workers and public. In addition, it offers unmatched protection against beyond design basis (BDB) events like the tsunami at Fukushima in Japan, the earthquake at North Anna site, tornadoes and manmade attacks such as the 9/11 attack on the World Trade Center in New York. It also includes significant enhancement for a robust Canister Evolved Safety Program for Aging Management. The utilities in the USA are currently using dry storage systems with canisters of different diameter, different length, and different weights. The fuel assemblies that are stored in these canisters are also different in their fuel types (PWR or BWR),different fuel parameters like burnups, enrichments, and cooling times and different storage orientations in the storage overpacks like horizontal or vertical. The compact version of AREVA TN's NUHOMS{sup R} System offers a 'Universal' Storage Overpack which is designed to accept all these different design dry storage canister available in the industry. As UNF storage continues to drives the industry towards innovations that address the need for consolidated storage, compact ISFSIs and aging management, this new evolution of the AREVA TN NUHOMS{sup R} system offers a significant leap forward in addressing all of these challenges. (authors)

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

Total Integrated Dose to Spent Fuel Pool Level Instrument Probes and Refueling Floor Personnel Dose Rates Meeting NRC Order EA-12-051 for a Typical Boiling Water Reactor - Paper 22

In reviewing the events after the March 11, 2011 Fukushima accident, the Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Reference 1, requiring new reliable spent fuel pool (SFP) instrumentation for pool level determination under beyond design basis event conditions. In 2013, URS prepared a calculation for a typical Boiling Water Reactor to determine the Total Integrated Dose (TID) potentially to be imparted to radiation sensitive underwater probes (one primary and one backup) planned for meeting this Order. Additionally, URS calculated the radiological dose rates to an individual standing on the refueling floor for various SFP fill levels. NEI 12-02, Reference 2, defines the levels at which SFP condition identification is necessary as the following: Level 1 - level that is adequate to support operation of the normal fuel pool cooling system. Level 1 represents the higher of the following two points: - The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or - The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis. Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. Level 2 is based on either of the following: - 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or - A designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components. Level 3 - level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e. +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. The URS calculation did not define these levels or determine instrumentation setpoints. However, it determined refueling floor dose rates that could be used as a tool to determine the appropriate water depths for these levels. In determining the TID for the SFP probes and the radiological dose rates to the individual, the following radioactive sources were considered: 1. Spent Nuclear Fuel (SNF) stored in the spent fuel storage racks, 2. Local Power Range Monitors (LPRMs) stored above the top of the racks on hangers on the SFP walls 3. Radioactive material laden filters stored on platforms on top of the racks, and 4. Control Rod Blades (CRBs) stored above the top of the racks on hangers on the SFP walls. The TID for the SFP probes includes both the dose from normal operation (pool full of water) for the remaining plant life (an assumed 40 years, conservatively including an allowance for lifetime extension SFP loading) and, consistent with the NEI 12-02 guidelines, Reference 2, the assumed drop in elevation of the pool water level to the top of the spent fuel racks for a 6 hour period. The acceptable TID for the radiation-sensitive probe as defined by the manufacturer was 2x10{sup 9} rad. The individual on the pool deck was assumed to be standing on the edge of the pool with an unobstructed view of its contents, as a worst case scenario. Dose rates to this individual determined from direct shine from the same 4 sources were determined for SFP water levels varying from 1 ft above the SNF racks to 21 ft above the racks in one foot increments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2000

Analysis and Test of a Proton Exchange Membrane Fuel Cell Power System for Space Power Applications

An effort is underway to develop a prototype Proton Exchange Membrane (PEM) Fuel Cell breadboard system for fuhlre space applications. This prototype will be used to develop a comprehensive design basis for a space-rated PEM fuel cell powerplant. The prototype system includes reactant pressure regulators, ejector-based reactant pumps, a 4-kW fuel cell stack and cooling system, and a passive, membranebased oxygen / water separator. A computer model is being developed concurrently to analytically predict fluid flow in the oxidant reactant system. Fuel cells have historically played an important role in human-rated spacecraft. The Gemini and Apollo spacecraft used fuel cells for vehicle electrical power. The Space Shuttle currently uses three Alkaline Fuel Cell Powerplants (AFCP) to generate all of the vehicle's 15-20kW electrical power. Engineers at the Johnson Space Center have leveraged off the development effort ongoing in the commercial arena to develop PEM fuel cel ls for terrestrial uses. The prototype design originated from efforts to develop a PEM fuel cell replacement for the current Space Shuttle AFCP' s. In order to improve on the life and an already excellent hi storical record of reliability and safety, three subsystems were focused on. These were the fuel cell stack itself, the reactant circulation devices, and reactant / product water separator. PEM fuel cell stack performance is already demonstrating the potential for greater than four times the useful life of the current Shuttle's AFCP. Reactant pumping for product water removal has historically been accomplished with mechanical pumps. Ejectors offer an effective means of reactant pumping as well as the potential for weight reduction, control simplification, and long life. Centrifugal water separation is used on the current AFCP. A passive, membrane-based water separator offers compatibility with the micro-gravity environment of space, and the potential for control simplification, elimination of moving parts in an oxygen environment, and long life. The prototype system has been assembled from components that have previously been tested and evaluated at the component level. Preliminary data obtained from tests performed with the prototype system, as well as other published data, has been used to validate the analytical component models. These components have been incorporated into an integrated oxidant fluid system model. Results obtained from both the performance tests and the analytical model are presented.

Vasquez, Arturo↗
DOE OSTI2015

Nonlinear Time Domain Seismic Soil-Structure Interaction (SSI) Deep Soil Site Methodology Development

Currently the Department of Energy (DOE) and the nuclear industry perform seismic soil-structure interaction (SSI) analysis using equivalent linear numerical analysis tools. For lower levels of ground motion, these tools should produce reasonable in-structure response values for evaluation of existing and new facilities. For larger levels of ground motion these tools likely overestimate the in-structure response (and therefore structural demand) since they do not consider geometric nonlinearities (such as gaping and sliding between the soil and structure) and are limited in the ability to model nonlinear soil behavior. The current equivalent linear SSI (SASSI) analysis approach either joins the soil and structure together in both tension and compression or releases the soil from the structure for both tension and compression. It also makes linear approximations for material nonlinearities and generalizes energy absorption with viscous damping. This produces the potential for inaccurately establishing where the structural concerns exist and/or inaccurately establishing the amplitude of the in-structure responses. Seismic hazard curves at nuclear facilities have continued to increase over the years as more information has been developed on seismic sources (i.e. faults), additional information gathered on seismic events, and additional research performed to determine local site effects. Seismic hazard curves are used to develop design basis earthquakes (DBE) that are used to evaluate nuclear facility response. As the seismic hazard curves increase, the input ground motions (DBE’s) used to numerically evaluation nuclear facility response increase causing larger in-structure response. As ground motions increase so does the importance of including nonlinear effects in numerical SSI models. To include material nonlinearity in the soil and geometric nonlinearity using contact (gaping and sliding) it is necessary to develop a nonlinear time domain methodology. This methodology will be known as, NonLinear Soil-Structure Interaction (NLSSI). In general NLSSI analysis should provide a more accurate representation of the seismic demands on nuclear facilities their systems and components. INL, in collaboration with a Nuclear Power Plant Vender (NPP-V), will develop a generic Nuclear Power Plant (NPP) structural design to be used in development of the methodology and for comparison with SASSI. This generic NPP design has been evaluated for the INL soil site because of the ease of access and quality of the site specific data. It is now being evaluated for a second site at Vogtle which is located approximately 15 miles East-Northeast of Waynesboro, Georgia and adjacent to Savanna River. The Vogtle site consists of many soil layers spanning down to a depth of 1058 feet. The reason that two soil sites are chosen is to demonstrate the methodology across multiple soil sites. The project will drive the models (soil and structure) using successively increasing acceleration time histories with amplitudes. The models will be run in time domain codes such as ABAQUS, LS-DYNA, and/or ESSI and compared with the same models run in SASSI. The project is focused on developing and documenting a method for performing time domain, non-linear seismic soil structure interaction (SSI) analysis. Development of this method will provide the Department of Energy (DOE) and industry with another tool to perform seismic SSI analysis.

58 GEOSCIENCES↗
DOE OSTI2023

Accident event progression, gaps, and key performance indicators for steam generator tube rupture events in water-cooled SMRs: A review

According to historical records of reactor-related incidents, a steam generator tube rupture (SGTR) is one of the most common occurrences at operating pressurized water reactors (PWRs). Such design-basis accidents (DBAs) could lead to a direct path for radionuclides to be released to the atmosphere via the safety and relief valves, making assessments of radionuclide discharge from operating nuclear power plants (NPPs) into the environment crucial for ensuring safety. In such analyses, the primary focus has been on the extent of radioactive release, not the potential damage to the core. Moreover, significant and timely intervention by operators is needed during the initial stages of SGTRs in order to avert overfilling of the SGs, as well as to restrict the dissemination of radioactive materials. Determining the event sequence and phases that occur during an SGTR incident in an advanced passive (e.g., AP1000) water-cooled nuclear reactor is crucial for implementing effective passive safety systems (PSSs) for a specific water-cooled small modular reactor (SMR) design. Here this study provides a comprehensive review of the accident event progression, associated physical phenomena, knowledge gaps, and key performance indicators that must be evaluated and assessed in thermal-hydraulics models, based on relevant test data, respectively.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Seismic fragility analysis for lateral supports of HTR-PM RPV - 18521

Since Fukushima accident, there has been increasing interest in China to study the risks of nuclear power plants under seismic events that are beyond the design basis. Two methodologies are commonly being adopted for this purpose, namely the seismic probabilistic risk assessment (SPRA) and the seismic margin assessment (SMA). Fragility analysis plays an important role as it not only provides fragility curves as input for the SPRA, but also can deduce HCLPF values for the SMA. In this paper, seismic fragilities of the lateral supports of the HTR-PM reactor pressure vessel (RPV) are studied. The lateral supports are crucial to the safety of HTR-PM RPV under seismic events. As they are located at a high elevation in the reactor building, it is expected they are one of the major contributors to the seismic risk of HTR-PM. The lateral supports are first briefly introduced in terms of structural features, load conditions and stress acceptance criteria in the engineering design. Their fragilities are then analyzed by identifying basic fragility analysis variables and determining values in terms of the median and standard deviations. The associated fragility curves are finally calculated and a HCLPF value is reported. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Advancements in Multiphysics Microdepletion Analysis of an eVinci TM -like Microreactor Leveraging OpenMC-CRAB Workflow

Nuclear microreactors (MRs) are a class of nuclear reactor technology, characterized by reduced dimensions, modular design, and reduced power output in contrast to conventional Light Water Reactors (LWRs). MRs are proposed for supplying electricity and eventual process heat to remote locations, such as military installations and disaster-affected areas. Current research work sponsored by the US Department of Energy Microreactor Program (MRP) is devoted to the development of novel modeling and simulation tools to better support MR vendors and regulatory bodies. Notably, the NRC is projected to utilize the CRAB multiphysics software driver for executing both design and beyond-design-basis accident analyses. Furthermore, the NRC has been utilizing the MELCOR code to calculate mechanistic source terms during accidents. Since MELCOR relies on isotopic inventory and reactor temperature/power profiles under accident conditions, which theoretically can be derived from CRAB, the goal is to establish a comprehensive CRAB-MELCOR computational framework. Past work was focused on testing and demonstrating CRAB's capability to generate results that can be used to inform mechanistic source term calculations in MELCOR. In particular, a computational workflow leveraging OpenMC-generated microscopic cross sections and CRAB was first applied to perform multiphysics microscopic depletion calculation followed by an accident scenario for a stylized microreactor problem. In fiscal year 2024, the research work has been focused on applying the OpenMC-CRAB workflow, which was first tested in fiscal year 2023, to a realistic 3D heat-pipe cooled MR problem representative of the eVinci TM design. The latter computational problem was developed with inputs from WEC to conserve selected neutronic and thermal characteristics of the eVinci TM design without releasing proprietary data. The results of this simulation, encompassing isotopic inventory, power density distribution, and kinetic parameters, will inform both MELCOR and the WEC-developed FATE code for mechanistic source terms calculations. The results from the two codes will then be compared for code verification purposes. This report contains the design characteristics of the realist heat pipe cooled microreactor developed as a use-case for the verification exercise, and the current results for the multiphysics microscopic depletion performed with the OpenMC-CRAB workflow. The results include eigenvalue as a function of time, power distribution at EOL, in addition to nuclides inventory's time evolution and spatial distribution. Finally, we report improvements to the workflow efficiency achieved through a collaboration with the NEAMS programs. Through this collaborative effort, we were able to strongly decrease the computational time for the multiphysics microdepletion calculation (i.e., from 17.4 hours to 5.7 hours on 280 processors) in addition to simplifying the interface to generate isotopics spatial distribution utilizable by FATE and MELCOR. Future work, including the improvement of the current microscopic cross-sections' library and the simulation of an accident scenario at EOL, is also discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Core Conversion of VVR-K Research Reactors

The conversion of the VVR-K research reactor to low-enrichment uranium fuel includes switching to a compact core with low-enrichment uranium and gradually transitioning from a light-water side reflector to a beryllium reflector with restoration of the working reactivity margin without makeup with fresh fuel. In 2015, the reactor was stopped in order to modernize the main elements of the core, the control and protection system for the channels of the working rods, and the emergency electric power source. First criticality of VVR-K with low-enrichment fuel followed by first start of the reactor occurred in spring 2016. The VVR-KN type FA specially developed for VVR-K make it possible to not only preserve but also improve the neutronphysical characteristics of the core. The design-basis reactivity margin of the working core load ~7%Δk/k is adequate for reactor operation for three 20-day cycles with ~10% average burnup reached in the central FA. The beryllium reflector is built up in stages starting at the fourth cycle. The beryllium reflector will make it possible to preserve the radial dimensions of the core and increase the neutron flux density in the central and peripheral irradiation channels.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Progress Report on Modified Burst Testing and Alternative Test Methodologies

Performance of accident-tolerant fuel candidates must be assessed during the pellet-cladding mechanical interaction (PCMI) phase of a design-basis reactivity-initiated accident (RIA). Therefore, in this report, we describe Fiscal Year (FY) 2019 research activities on the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions comparable to the PCMI. The first section of this report provides modified-burst test (MBT) results of unirradiated FeCrAl tubes at various pressurization-rates and an RIA-relevant temperature (275°C). The failure strain of FeCrAl was observed and calculated using a high-speed digital-image correlation (DIC) technique developed recently under the Department of Energy (DOE)’s Advanced Fuel Campaign (AFC). To overcome the complexities of performing MBT in a hot-cell environment on irradiated samples, a “plane-strain tension” (PST) test was proposed as an alternative methodology for PCMI testing. The relevance of the PST test as an alternative to MBT for irradiated specimens was explicated using the data available in the literature. The design of hot-cell compatible PST fixtures and a specific holder for PST hot-cell specimen machining were developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

OPSAID improvements and capabilities report.

Process Control System (PCS) and Industrial Control System (ICS) security is critical to our national security. But there are a number of technological, economic, and educational impediments to PCS owners implementing effective security on their systems. Sandia National Laboratories has performed the research and development of the OPSAID (Open PCS Security Architecture for Interoperable Design), a project sponsored by the US Department of Energy Office of Electricity Delivery and Energy Reliability (DOE/OE), to address this issue. OPSAID is an open-source architecture for PCS/ICS security that provides a design basis for vendors to build add-on security devices for legacy systems, while providing a path forward for the development of inherently-secure PCS elements in the future. Using standardized hardware, a proof-of-concept prototype system was also developed. This report describes the improvements and capabilities that have been added to OPSAID since an initial report was released. Testing and validation of this architecture has been conducted in another project, Lemnos Interoperable Security Project, sponsored by DOE/OE and managed by the National Energy Technology Laboratory (NETL).

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2025

Regulatory Considerations for Domestic Reprocessing Facility Physical Security

U.S. advanced non-light-water reactor vendors may pursue collocated on-site reprocessing activities. Therefore, these facilities are likely to possess formula quantities, or Category I quantities, of special nuclear material (SNM) during normal operations. The U.S. Nuclear Regulatory Commission (U.S. NRC) has yet to formally establish a regulatory framework for commercial reprocessing. While Category I requirements would explicitly not apply in this circumstance under current regulatory requirements, regulatory certainty does not exist. A novel framework should be developed to ensure public health and safety while also risk-informing the physical security requirements. This report reviews the relevant background of related rulemaking activities and proposes risk-informed physical protection requirements to satisfy these objectives. Insights from NRC security-related rulemaking activities provide a substantial technical basis to approach potential establishment of physical security requirements for reprocessing facilities. If a licensee can provide justification that the material satisfies a sufficient self-protecting radiation dose threshold, the material may not be subject to theft or diversion requirements and only potential sabotage requirements would apply. Furthermore, if the material can be justified to be moderately dilute, a set of risk-informed requirements could provide adequate protection of public health and safety. A revised performance objective for prevention of theft of moderately dilute Category I SNM may be detection to allow prompt recovery by a local law enforcement agency. However, a significant caveat to the proposed categorization scheme is the unknown integration of radiological sabotage with requirements for the protection against theft. Future licensees should consult with the NRC regarding treatment of this regulatory topic. Additionally, the self-protecting radiation dose threshold (either the existing or a proposed future threshold) would need to be considered. An integrated approach may apply graded potential requirements for protection against the design basis threat of radiological sabotage currently applicable to commercial nuclear power plants and Category I SNM facilities defined within 10 CFR 73.1(a).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Development of Accident Tolerant Fuel Options For Near Term Applications

In this section, the basis for the proposal followed by description of the work scope and methodologies are discussed. Following the Fukushima disaster in 2011, US congress mandated the department of energy to start a R&D program on accident tolerant fuel (ATF) concepts for the existing light water reactor fleet. By 2015, the start date of this IRP, several ATF concepts were being pursued by various entities. In general, the ATF concepts were divided into two categories: near term and long term. The near term concepts included: Coated Zircaloy clad, fuels with additives and dopants and FeCrAl/steel based claddings. The long term concepts included: SiC composite cladding, high density fuels (U 3 Si 2 , UN) and TRISO type fuel forms (e.g. FCM). It is of particular importance for any DOE funded effort to give background information and original Funding Opportunity Announcement (FOA) goals in its final report to evaluate whether the FOA goals and the proposed work were actually addressed. The FOA for this IRP sought for modeling and simulation capability to predict various ATF performances during normal operation, design basis accident and severe accident conditions. Particularly, the FOA was concern with the state of the core performance during a severe accident. It was postulated that if an ATF retains its structural integrity, the performance of non-fuel structures in the core needs to be addressed. The IRP team focused the proposal on Pressurized Water Reactor (PWR) performance for near term ATF concepts: Coated cladding, Fuel with additives and dopants to accelerate adoption of such concepts and address the unanswered questions regarding their performance. The project also gave limited focus on another near term concept, FeCrAl cladding, by restricting its scope to time-to-failure prediction since substantial effort supported by other programs including ORNL ATF program, GEH ATF program and NEAMS High Impact Problem were already made or underway during the IRP tenure. Despite two previous IRPs funded in 2012 with focus on experimental study of metallic and ceramic Zircaloy coated claddings, this IRP still included a coating program. This was a testament to the project leadership’s belief that the coated clad concept will be the primarily initial pathway for near term industrialization of ATF technology due to presence of enrichment penalty and licensing hurdles of the alternative clad concepts. In doing so, the IRP was positioned to provide key findings on coated Zircaloy as a uniquely funded external program to DOE’s industrial ATF campaign. Currently, in 2019, the coated clad concept is being pursued by all three vendors for licensed application and Lead Test Rods and Lead Test Assemblies (LTRs/LTAs) have already been inserted or planned to be inserted in the existing fleet. At the start of the IRP project, based on the meeting with the projects technical and federal point of contacts, it was also agreed upon if the contractual obligations are met, then the scope of the simulation work will be extended to Boiling Water Reactors as included in Section 2.10.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Engineering Design of a Linde-BASF Advanced Post-Combustion CO 2 Capture Technology at a Linde Steam Methane Reforming H 2 Plant

Linde carried out an initial engineering design study for a Linde-BASF advanced post combustion CO 2 capture technology to be installed at a commercial-scale steam methane reforming (SMR) hydrogen plant located in the US Gulf Coast. This pre-FEED equivalent study included following: (1) basic design, including specific project scope definition and design basis, (2) basic engineering, including development of process flow diagrams and heat & material balances, (3) inside the battery limit (ISBL) equipment and systems specification, (4) balance of plant outside the battery limit (OSBL) equipment and systems specifications, (5) technology maturation plan, (6) hazard and operability (HAZOP) review, (7) environmental, health and safety (EH&S) assessment and environmental permitting analysis, (8) constructability review, (9) ISBL and OSBL EPC cost estimation, and (10) commercial-scale techno-economic analysis including capital expenditures (CAPEX) and operating expenditures (OPEX) and CO 2 capture cost estimates.

08 HYDROGEN↗
DOE OSTI2017

Diversity and defense-in-depth analysis for I and C systems of research reactors: A case study on two research reactors - 376

This paper provides the analysis results of the diversity and defense-in-depth (D3) applied to two research reactors designed by the Korea Atomic Energy Research Institute. The D3 assessments for instrumentation and control (I and C) systems of two research reactors are performed according to the guidelines documented in NUREG/CR-6303. The I and C systems are classified into four echelons of defense: control echelon, reactor trip echelon, engineered safety feature (ESF) actuation echelon, and monitoring and indication echelon. The diversity between the echelons of defense is assessed to determine that a single software common cause failure (CCF) does not impair the functions of multiple echelons of defense. In addition, the effects of a software CCF of the reactor protection system on the design basis events of the research reactors are analyzed by assuming that the automated reactor trip and ESF actuation functions of the reactor protection system are inoperable due to the software CCF. From the D3 analyses for the research reactors, it is verified that (1) adequate diversity has been provided, (2) adequate defense-in-depth has been provided, and (3) the displays and manual controls for critical safety functions initiated by operator action are diverse from digital systems used in the automatic portion of the protection system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Criticality Safety Analysis of Spent Nuclear Fuel Canisters Using As-Loaded Configurations

Dual purpose canisters (DPCs) used for storage and transportation of spent nuclear fuel (SNF) are typically designed and evaluated using bounding (enveloping) fuel characteristics such as fuel type, fuel dimensions, initial enrichment, discharge burnup, and cooling time. This is a design basis, bounding licensing approach for SNF storage and transportation systems, as licensing and supporting safety analysis reviews are performed prior to the actual fuel loading. The bounding fuel characteristics for a system are developed by fully utilizing the regulatory and regulators recommended safety limits such as neutron multiplication factor (keff) approaching 0.95 to maximize the system utilizations. In reality, there are wide variations in SNF assembly burnups, initial enrichments, and cooling times. Therefore, dry storage systems are typically loaded with assemblies that satisfy the bounding fuel characteristics defined in the certificate of compliance (CoC), with some amount of unquantified and uncredited margin. This paper presents an as-loaded canister-specific criticality analysis approach for quantifying inherent uncredited margins in already loaded DPCs. The as-loaded analysis approach has been implemented in a new SNF management and analysis tool - The Used Nuclear Fuel-Storage, Transportation, and Disposal Analysis Resource and Data System (UNF ST&DARDS). The paper presents as-loaded criticality analysis of 616 canisters at 28 US reactor sites. Additionally, the paper discusses as-loaded analysis methodology validation approach using detailed reactor operational and fuel design data.

Banerjee, Kaushik↗
DOE OSTI2010

Risk-Informed Safety Margin Characterization (RISMC): Integrated Treatment of Aleatory and Epistemic Uncertainty in Safety Analysis

The concept of “margin” has a long history in nuclear licensing and in the codification of good engineering practices. However, some traditional applications of “margin” have been carried out for surrogate scenarios (such as design basis scenarios), without regard to the actual frequencies of those scenarios, and have been carried out with in a systematically conservative fashion. This means that the effectiveness of the application of the margin concept is determined in part by the original choice of surrogates, and is limited in any case by the degree of conservatism imposed on the evaluation. In the RISMC project, which is part of the Department of Energy’s “Light Water Reactor Sustainability Program” (LWRSP), we are developing a risk-informed characterization of safety margin. Beginning with the traditional discussion of “margin” in terms of a “load” (a physical challenge to system or component function) and a “capacity” (the capability of that system or component to accommodate the challenge), we are developing the capability to characterize probabilistic load and capacity spectra, reflecting both aleatory and epistemic uncertainty in system response. For example, the probabilistic load spectrum will reflect the frequency of challenges of a particular severity. Such a characterization is required if decision-making is to be informed optimally. However, in order to enable the quantification of probabilistic load spectra, existing analysis capability needs to be extended. Accordingly, the INL is working on a next-generation safety analysis capability whose design will allow for much more efficient parameter uncertainty analysis, and will enable a much better integration of reliability-related and phenomenology-related aspects of margin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗