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DOE OSTI2026

Pressurized Water Reactor Containment Aerosol Settling Rate for Accident Source Terms

This report documents results from MELCOR calculations performed to determine the aerosol settling rate in pressurized water reactor containments in support of the Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” which is the methodology for determining alternate source terms in design basis accidents at nuclear power reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Best-Estimate-Plus-Uncertainty-Informed Deterministic Safety Evaluation

Nuclear power plant (NPP) utilities operate contingent on receiving a favorable safety assessment from their national regulatory and safety authority. Among the many facets of NPP safety addressed is verification of integral design-basis performance - a demonstration of a plant's response to a suite of challenging transient and accident scenarios. Thermal-hydraulic system computer codes like RELAP5, TRAC, CATHARE, and ATHLET are used as the simulation engine as part of a broader evaluation model that accounts for a variety of uncertainties. It is common for these evaluation models to apply either a super-bounding/ deterministic or best-estimate plus uncertainty (BEPU) approach. The trend over the past two decades has been away from the classical deterministic methods towards methods explicitly incorporating parametric uncertainty models. These methods center upon the evaluation of parametric and system response uncertainties, where these uncertainties are derived from the convolution of one or more parametric uncertainty models. Acceptance of BEPU methods has provided the nuclear industry benefits regarding operational flexibility and, in some cases, power output. The advantage of BEPU for new NPPs is less obvious. Generally, these new designs include engineered safety features that practically eliminate fuel damage resulting from design-basis events. To substantiate such claims, applying super-bounding assumptions on uncertainties may be a more expedient demonstration of a plant's safety case than those based on BEPU. In addition, the requirements on design inputs for a deterministic evaluation model are substantially less than that for a BEPU evaluation model. As such, applying a deterministic evaluation model is expected to have the advantage of reducing the duration of the licensing review process. For the current operating fleet of light-water reactors (LWRs), the nuclear safety authorities in most countries have followed the U.S. Nuclear Regulatory Commission's rules for the preparation of safety analysis. These rules are specified in the U.S. Code of Federal Regulations, most notably in, Part 10, Section 50.46, entitled 'acceptance criteria for emergency core cooling systems for light-water nuclear power reactors' and its Appendix K. Notably, Appendix K specifies deterministic requirements for analysis of the loss-of-coolant accident (LOCA), which is generally regarded as the maximum credible accident for LWRs. Projecting 10 CFR 50, Appendix K, to new plants is not obviously applicable as it addresses light water fluid flow and heat transfer phenomena associated with core uncovery. Preventing core uncovery is a requirement for several designs in the class of small modular reactors (SMR). In doing so, the majority of phenomenological uncertainties prescribed in Appendix K are eliminated. Similarly, the impact of other phenomena to effect a change in various safety measures, i.e., figures-of-merit, is expected to be different. The implication is that development of a deterministic evaluation model must be derived with an original technical basis rather than solely rely on Appendix K. While a logical conclusion, the expectation of original bases, regardless of the treatment of uncertainties, is embodied in the U.S. NRC's Regulatory Guide (RG) 1.203, Transient and Accident Analysis. Within the evaluation model development and application process framework of RG 1.203, this paper describes a BEPU-informed deterministic evaluation model for NPPs with large inherent safety margins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

EDS V26 Containment Vessel Explosive Qualification Test Report

The objective of the test was to qualify the vessel for its intended use by subjecting it to a 1.25 times overtest. The criteria for success are that the measured strains do not exceed the calculated strains from the vessel analysis, there is no significant additional plastic strain on subsequent tests at the rated design load (shakedown), and there is no significant damage to the vessel and attached hardware that affect form, fit, or function. Testing of the V25 Vessel in 2011 established a precedent for testing V26 [2]. As with V25, two tests were performed to satisfy this objective. The first test used 9 pounds of Composition C-4 (11.25 lbs. TNT-equivalent), which is 125 percent of the design basis load. The second test used 7.2 pounds of Composition C-4 (9 lbs. TNT-equivalent) which is 100 percent of the design basis load. The first test provided the required overtest while the second test served to demonstrate shakedown and the absence of additional plastic deformation. Unlike the V25 vessel, which was mounted in a shipping cradle during testing, the V26 vessel was mounted on the EDS P2U3 trailer prior to testing. Visual inspections of the EDS vessel, surroundings, and diagnostics were completed before and after each test event. This visual inspection included analyzing the seals, fittings, and interior surfaces of the EDS vessel and documenting any abnormalities or damages. Photographs were used to visually document vessel conditions and findings before and after each test event.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Evaluation of Detritiation Strategies for Mitigating Tritium Releases

Tritium confinement is performed using different barriers to minimize releases to the environment. Primary confinement is usually performed by process piping and components, secondary confinement typically by inerted gloveboxes connected to a continuously operating tritium stripper system. Some facility designs employ tertiary tritium confinement by using process room confinement with a tritium stripper system activated after an accident to further mitigate tritium releases to the environment. Secondary confinement of some tritium systems such as radiation-hardened or shielded-cell enclosures are not easily achieved using designs for inerted glovebox. These enclosures can be challenging to seal for secondary tritium confinement due to heating, ventilation, and air conditioning (HVAC) needs for temperature control, as well as feedthroughs for equipment and instrumentation, are potential pathways for tritium leaks or loses. Additional tritium leak paths can be created in secondary confinement enclosures due to a Design Basis Accident (DBA) seismic event. These new, larger tritium leak rates from the DBA event are usually not part of the stripper system design basis so all tritium released to the secondary confinement enclosure is assumed released to the environment. Detritiation strategies for shielded enclosures and/or gloveboxes with significant leak rates are necessary to minimize off-site radiological dose consequences. To address this, the relationship between the confinement system leak rate and stripper system (recirculating) flow rate was evaluated to meet prescribed maximum allowable environmental tritium emissions from the event. The simple analysis described assumes a constant leak rate from the confinement system (i.e. shielded enclosure or glovebox) while a recirculating stripper system strips tritium from the system– a competition between tritium release and tritium recovery. At a high level, the resulting expression summarizes the relationship between system leak rate (F L ) and stripper flow rate (F S ) relative to allowable tritium release (Q A ) and initial tritium release (Q 0 ): $\frac{F_L}{F_s}$ ≈ $\frac{Q_A}{Q_0}$. This report derives the relationship between these parameters, examines the impact of finite stripping times followed by purging of the stripped volume. The analysis provides example results for up to 30 gram tritium releases: the maximum tritium inventory of a Hazard Category III (Department of Energy) Nuclear Facility. The detritiation model presented represents a high-level analysis of tritium recovery versus losses for “leaky” confinement enclosures such as shielded cells after large tritium releases. Key parameters for doing the analyses are the system leak rate, stripper flow rate, initial tritium release, and allowable tritium release values. The analyses show a proportional decrease in releases with reduction in initial tritium release but a non-linear increase in tritium recovery with increased stripper flow rate or reduced leak-rate. The analyses also apply to glovebox confinement systems which could have significant increases in leak rates after a DBA. The analysis recognizes but does not include tritium absorption followed by re-emission from the walls of the confinement volume – an analysis which could be pursued in future studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Evaluation of the Performance of the SAVY Container Subjected to Various Impact Events using Finite Element Methods

The SAVY container is a Safety Class Design Feature utilized both inside engineered features and outside engineered features (e.g. gloveboxes) at TA-55 Plutonium Facility Building 4. These containers were certified as low damage ratio containers by demonstrating their ability to retain material after exposure to impacts and elevated temperatures resulting from a design basis seismic event. Original testing of the SAVY did not evaluate the effects of various facility specific insults to the SAVY while within a DOT Type A Container, caused either by design basis accident or an operational mishap. These insults include: 1. Impact of a structural I-beam (drum upright & drum on side); 2. Impact of a section of Positive Pressure Chilled Water Pipe (PPCWS); 3. Forklift fork tine impact; 4. Fuel truck impact. Abaqus Explicit Dynamics Finite Element Analysis was performed on a nested SAVY/DOT Type 7A container configuration to assess the structural integrity of the SAVY under the accident conditions listed.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2012

NYSEG Seneca Compressed Air Energy Storage (CAES) Demonstration Project

Compressed Air Energy Storage (CAES) is a hybrid energy storage and generation concept that has many potential benefits especially in a location with increasing percentages of intermittent wind energy generation. The objectives of the NYSEG Seneca CAES Project included: for Phase 1, development of a Front End Engineering Design for a 130MW to 210 MW utility-owned facility including capital costs; project financials based on the engineering design and forecasts of energy market revenues; design of the salt cavern to be used for air storage; draft environmental permit filings; and draft NYISO interconnection filing; for Phase 2, objectives included plant construction with a target in-service date of mid-2016; and for Phase 3, objectives included commercial demonstration, testing, and two-years of performance reporting. This Final Report is presented now at the end of Phase 1 because NYSEG has concluded that the economics of the project are not favorable for development in the current economic environment in New York State. The proposed site is located in NYSEG’s service territory in the Town of Reading, New York, at the southern end of Seneca Lake, in New York State’s Finger Lakes region. The landowner of the proposed site is Inergy, a company that owns the salt solution mining facility at this property. Inergy would have developed a new air storage cavern facility to be designed for NYSEG specifically for the Seneca CAES project. A large volume, natural gas storage facility owned and operated by Inergy is also located near this site and would have provided a source of high pressure pipeline quality natural gas for use in the CAES plant. The site has an electrical take-away capability of 210 MW via two NYSEG 115 kV circuits located approximately one half mile from the plant site. Cooling tower make-up water would have been supplied from Seneca Lake. NYSEG’s engineering consultant WorleyParsons Group thoroughly evaluated three CAES designs and concluded that any of the designs would perform acceptably. Their general scope of work included development of detailed project construction schedules, capital cost and cash flow estimates for both CAES cycles, and development of detailed operational data, including fuel and compression energy requirements, to support dispatch modeling for the CAES cycles. The Dispatch Modeling Consultant selected for this project was Customized Energy Solutions (CES). Their general scope of work included development of wholesale electric and gas market price forecasts and development of a dispatch model specific to CAES technologies. Parsons Brinkerhoff Energy Storage Services (PBESS) was retained to develop an air storage cavern and well system design for the CAES project. Their general scope of work included development of a cavern design, solution mining plan, and air production well design, cost, and schedule estimates for the project. Detailed Front End Engineering Design (FEED) during Phase 1 of the project determined that CAES plant capital equipment costs were much greater than the $\$$125.6- million originally estimated by EPRI for the project. The initial air storage cavern Design Basis was increased from a single five million cubic foot capacity cavern to three, five million cubic foot caverns with associated air production wells and piping. The result of this change in storage cavern Design Basis increased project capital costs significantly. In addition, the development time required to complete the three cavern system was estimated at approximately six years. This meant that the CAES plant would initially go into service with only one third of the required storage capacity and would not achieve full capability until after approximately five years of commercial operation. The market price forecasting and dispatch modeling completed by CES indicated that the CAES technologies would operate at only 10 to 20% capacity factors and the resulting overall project economics were not favorable for further development. As a result of all of these factors, the Phase 1 FEED developed an installed CAES plant cost estimate of approximately $\$$2,300/KW for the 210MW CAES 1A and 2 cycles. The capital cost for the 136 MW CAES 1 cycle was even higher due to the lower generating capacity of the cycle. Notably, the large equipment could have generated additional capacity (up to 270MW) which would have improved the cost per KW; however, the output was limited by the night time transmission system capability. The research herein, therefore, is particular to the site-specific factors that influenced the design and the current and forecasted generation mix and energy prices in Upstate New York and may not necessarily indicate that CAES plants cannot be economically constructed in other places in New York State or the world.

25 ENERGY STORAGE↗

DOE OSTI2021

UKy-CAER Heat Integrated Transformative CO 2 Capture Process in Pulverized Coal Power Plants

The goal of this Phase II Topical Report is to summarize the work for public domain conducted on project DE-FE0031583. Building upon the Design Basis completed during Phase I and in accordance with the Statement of Project Objectives (SOPO), the University of Kentucky (UK) Center for Applied Energy Research (CAER) (Recipient) along with team members has completed the Front End Engineering Design (FEED) for its advanced, versatile post-combustion CO 2 capture system (CCS) at the 10 MWe large pilot scale to be installed at the Wyoming (WY) Integrated Test Center (ITC), using a heat integrated process with split-rich fed, two-stage stripping and any advanced solvent to enhance the CO 2 absorber performance and lower the cost of CO 2 capture. The proposed project (Phases I-III) involves the design, fabrication, installation, testing and evaluation of a large pilot scale facility that will demonstrate the UK CAER transformative CCS integrated with an operating power plant. The work performed in Phase I, detailed in the Phase I Topical Report that was submitted in March 2019, included formation of a project team, securing the host site, and completing pre-FEED design basis, a cost estimation, a preliminary techno-economic analysis (TEA) and an environmental information volume (EIV). The work performed in Phase II, detailed in this Phase II Topical Report, includes completion of FEEDs, completion of a process risk analysis (PRA), obtaining necessary permits, updating the EIV and TEA, updating the cost and schedule estimate for the Phase III scope of work (SOW) and securing the necessary cost share. Updated information is provided on the current status of the UK CAER CCS technology and the project is ready to begin immediately detailed design, construction and operation of the proposed pilot.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2020

Evaluation of ASCE 4-16 and AISC 43-18 (Draft) for use in the Risk-Informed Performance-Based Seismic Design of Nuclear Power Plant Structures, Systems, and Components

This report describes the assessment of ASCE Standards 4-16 and 43-18 (Draft) for use in the Risk-Informed Performance-Based (RIPB) seismic design of structures, systems, and components (SSCs) at nuclear power plants. This work was performed for the U.S. Nuclear Regulatory Commission (NRC) Office of Regulatory Research (RES), to support potential endorsement of these industry standards for the design of nuclear power plants based on the RIPB approach. Currently, the NRC endorses a deterministic design approach for demonstrating the design adequacy of SSCs based on the Standard Review Plan (NUREG-0800) and NRC Regulatory Guides. In the RIPB approach, the design criteria are developed to achieve a target performance goal, which is defined by the annual frequency of occurrence of the design basis earthquake (i.e., Seismic Design Category) and the acceptable level of structural performance (i.e., Limit State) for the SSCs. ASCE 4-16 provides methods for performing a seismic analysis of structures to obtain the seismic response of these structures (e.g., building displacements, accelerations, in-structure response spectra) which are used in the design of the SSCs. It also provides methods for performing seismic analysis of SSCs to determine the seismic demands (e.g. member forces and displacements) needed to design individual SSCs. ASCE 43-18 (Draft) provides the criteria for the seismic design of SSCs using the seismic demands developed in ASCE 4-16. The use of ASCE Standard 43-18 (Draft), along with ASCE 4-16, provides the criteria for the seismic design of the SSCs. ASCE 43-18 (DRAFT) relies on other consensus codes and standards such as ACI 349 for reinforced concrete, AISC/N690 for steel structures, ASME Section III for pressure-retaining mechanical components and Containments, and IEEE-344 for Class 1E equipment. The goal of this technical review was to assess the adequacy of the provisions in these standards for use by the NRC in developing regulatory guidance for design of SSCs in nuclear power plants, based on the RIPB approach. The research reported herein describes the basis for acceptance of the new standards and identifies areas where additional staff guidance is needed for the seismic design of SSCs at nuclear power plants. This technical review has determined that ASCE 4-16 and ASCE 43-18 (Draft) provide an appropriate framework for the seismic design of SSCs at nuclear power plants using a Risk-Informed Performance-Based approach. However, some of the criteria therein warrant exceptions, qualifications, and/or clarifications.

42 ENGINEERING↗

DOE OSTI2015

Severe Accident Test Station Design Document

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Risk and Cost Analysis of Utilizing FLEX Equipment for O&M Cost Reduction in Nuclear Power Plants

Several Nuclear Power Plant (NPP) operators in USA have announced to close their plants despite there are remaining years on their operating licenses. The cause of these closures is because the plant operators suffer financial losses. Approximately 80% of USA NPPs would have undergone financial losses if they were situated in a deregulated electricity market. In a deregulated market, electricity generators are independent entities separate from the transmission companies and Independent System Operator (ISO) which manages the market. NPPs in this market are termed merchant NPPs, and their profitability is determined solely from electricity generation revenues. Meanwhile, regulated and public power nuclear units’ profitability is determined by the regulatory process in place for each unit. Due to the complexity of differentiating each regulated and public power NPP’s regulations 36 NPPs are at an immediate risk of retiring early to stop losses, while 27 units will likely to follow in the long run. Today, cost is well-recognized and important factor in NPP decision-making that cannot be ignored. Therefore, it is essential to consider risk and cost together in decision-making to strengthen the position of the nuclear industry in an economically challenging energy market. With operational and maintenance (O&M) cost at 70% of the total operating expenditure of an NPP and some plants closing prematurely despite excellent safety record, there is an urgent need to address the high O&M cost. One of the novel entities that provide additional safety resulting in reduced risk of nuclear power plants is Flex equipment. Flex strategies were postulated by the US NRC in the wake of Fukushima Dai-chi accident to address beyond design-basis accidents and improve plant flexibility. Flex strategy comprise both onsite and off-site component storage for the provision of additional materials and equipment. This work explored innovative ways of utilizing the onsite Flex equipment during regular plant operation, for purposes in addition to the intended use of beyond design basis accidents. This report describes the risk and cost analysis of using Flex equipment during regular or unplanned maintenance.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Enduring Transuranic Liquid Waste Volume Forecast and Constituent Evaluation (FY20)

The “FY11 Radioactive Liquid Waste (RLW) Influent Boundary Condition Assessment and Forecast” report provided influent volume forecasts that were critical in establishing the design capacity for the new Radioactive Liquid Waste Transuranic (RLW-TRU) capability. The design for the new RLW-TRU capability was paused as the final design was being completed and prior to the transition of the new management and operations contract for Los Alamos National Laboratory (LANL). Recently, NNSA has directed the project to restart and complete the design effort. As part of the restart process, LANL was asked to validate or document any material changes in the influent volumes or constituents that formed the design basis for the new capability. This analysis updates the transuranic liquid waste influents volumes, incorporating the changes in the LANL program of record since the FY11 assessment, and includes a cursory validation of influent constituents. The results of the FY19 update to influent volume forecasts shows: 1) the influent volume of 29,000 L should continue to be used as the project design-basis; 2) There have been no material changes in the operational processes that would change the constituents of concern from either a treatment or material compatibility perspective.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

BISON Capability, Validation and Demonstration for Reactivity-Initiated Accidents

Reactivity Initiated Accidents (RIA) are design basis accidents that could have adverse impacts on the core coolability. In the unlikely event that sufficient reactivity is inserted into the reactor core by the ejected/dropped control rod, prompt energy deposition into the fuel can occur, which when sufficiently high can lead to fuel rod failure or, at large energy deposition levels, expulsion of UO2 fragments or molten UO2 material from the fuel rod. This results in a release of fission product and fuel into the coolant and has the potential to compromise core coolability and threatening the pressure boundary of the primary coolant system. The design basis RIA is one of the industry challenging problems that the CASL aimed to address; the CASL RIA Challenge Problem Charter [1] states, ?The Pressurized Water Reactor (PWR) Rod Ejection Accident (REA) and Boiling Water Reactor (BWR) Control Rod Drop Accident (CRDA) are postulated accidents with consequences that are important to nuclear safety (fuel rod integrity and core coolability). Currently each reload core design must be analyzed to meet regulatory acceptance criteria. The goal of CASLs ModSim capability for RIA is to model the event at a higher fidelity, with validation to existing tests, to better model the transient neutronics and the progression of the fuel and cladding thermal-mechanical behavior. These improved analytical capabilities can be used to better inform reload core design, limits on fuel assembly discharge burnup, restrictions on placement of fuel in the reactor, control rod insertion limits, operating margin, and performance sensitivities.? In support of that charter, BISON, the fuel performance code has been used to demonstrate the simulation of thermal-mechanical behavior of LWR fuels during a reactivity-initiated accident. The combination of mechanical, thermal, and thermal-hydraulic phenomena all present during an RIA makes a multi-physics code such as BISON a valuable tool for modeling these scenarios. This paper will highlight many of the activities associated with BISON relevant to reactivity-initiated accident capability development and validation efforts. These efforts have been performed under both CASL and the Nuclear Energy Advanced Modeling and Simulation (NEAMS) programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Deterministic Selection of Design Extension Condition without Core Melt

The concept of beyond design basis accidents (BDBA) recently has been changed to design extension conditions (DEC). DEC is required to be considered in the design process, and it comprises conditions in events without core melt and conditions in events with core melt. DEC with core melt is a part of severe accident. A systematic and deterministic approach to select DEC without core melt for a pressurized water reactor (PWR) is introduced.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel

The work presented in this report applies the MELCOR code to evaluate potential accidents in non-reactor nuclear facilities, focusing on Design Basis Accidents. Ten accident scenarios were modeled using NRC's best-estimate severe accident analysis code, MELCOR 2.2. The accident scenarios simulated a range of explosions and/or fires related to a nuclear fuel reprocessing facility. The objective was to evaluate the radionuclide source term to the environment following initiating explosion and/or fire events. The simulations were performed using a MELCOR model of the Barnwell Nuclear Fuel Plant, which was decommissioned before beginning reprocessing operations. Five of the accident scenarios were based on the Class 5 Design Basis Accidents from the Final Safety Analysis Report. Three of the remaining accident scenarios include sensitivity studies on smaller solvent fires. The final two accidents included an induced fire from an initial explosion. The radionuclide inventory was developed from ORIGEN calculations of spent PWR fuel with an initial enrichment of 4.5% U-235 by weight. The fuel aged for five years after a final 500-day irradiation cycle. The burn-up was conservatively increased to 60 GWd/MTU to bound current US operations. The results are characterized in terms of activity release to the environment and the building decontamination factor, which is related the leak path factor used in Department of Energy safety analyses. The MELCOR 2.2 results consider adverse consequences to the filters, ventilation system, and structures as a result of the explosions and fires. The calculations also include best-estimate models for aerosol transport, agglomeration, and deposition. The new calculations illustrate best-estimate approaches for predicting the source term from a reprocessing facility accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Thermal conductance measurements of bolted copper joints for SuperCDMS

Joint thermal conductance testing has been undertaken for bolted copper to copper connections from 60 mK to 26 K. This testing was performed to validate an initial design basis for the SuperCDMS experiment, where a dilution refrigerator will be coupled to a cryostat via multiple bolted connections. Copper used during testing was either gold plated or passivated with citric acid to prevent surface oxidation. Results obtained are well fit by a power law regression of joint thermal conductance to temperature and match well with data collected during a literature review.

Schmitt, R.↗

DOE OSTI2024

Temperature-Time Modeling of Spent Fuel Cladding in Dry Storage Casks

This report uses thermal and decay heat modeling to investigate spent fuel performance and the potential for cladding to anneal in dry storage conditions. Annealing is an important feature to investigate in SNF cladding because it has a direct relationship to cladding response in storage, transportation and disposal conditions. Annealed cladding may have lower strength than unannealed cladding, however its increased ductility would provide better protection against rupture in high strain rate situations such as severe accidents. These consequences are the driver for this work however are outside of the scope of this report. The modeling focused on three representative storage systems, the TN-32B, MAGNASTOR with TSC-37 canister, and NUHOMS AHSM with a 32PTH2 canister. This covers the vertical dual-purpose, vertical ventilated and horizontal ventilated casks respectively. Decay heat modeling using high and low enrichment assemblies was used to bound the decay heat curves that might be expected in dry storage. To bound the temperature relationship, the storage casks were modeled starting at the design basis heat loads with heat decaying through time. Although the results are bounding there is not an attempt to maximize conservatism, rather the intent to form a reasonable upper limit on temperature that will be broadly applicable to the U.S. cask fleet. This will allow materials testing to focus on relevant conditions for annealing that may affect the U.S. spent fuel inventory. The results show a clear dependence on heat load pattern in time in Figure S-1 and Figure S-2. This dependence is due to the different assembly decay heat curves for different assemblies in preferentially loaded casks. It shows the need for careful decay heat modeling when examining in service fuel temperatures that are less than the cask design basis heat load. The body of the report shows percent cladding cutoffs of 300°C and 350 °C as well. These results can be used to inform testing and conclusions about cladding performance through time and the potential for cladding annealing during dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

RCCS Testing with the Water-based NSTF: Year-1 Single-Phase Results

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large scale test facility has been constructed to generate validation data for passive decay heat removal systems of advanced reactors. Located at Argonne National Laboratory, the Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS). This report serves as a summary of the single-phase testing results obtained during the programs first year of matrix testing. An introduction to the high-level program objectives is first presented, followed by a brief summary of facility design, instrumentation, and control systems. Details of the NQA-1 driven test method are included, which govern the programs quality assurance requirements and provide a background for the testing strategy and procedures. Checkout activities, including verification of gross system mass and energy balances, heat loss quantification, and isothermal facility characterization, were performed early in the first year and results presented in this report. These tests have identified relevant facility characteristics such as K-factors, frictional pressure losses, parasitic heat loss rates, and transient heating response. Results are then presented from four accepted matrix tests where the facility established a steady-state, single-phase mode of flow operation. All matrix tests were performed with an identical top-level procedure: first initiating a heater power ramp to target levels, and then holding power levels constant until the facilities structural members and network piping reached thermal equilibrium. Upon obtaining stable thermal and hydraulic conditions, active cooling was initiated off of the bulk storage tank to maintain constant riser inlet temperatures of 30°C. This steady-state period was then held for a minimum required duration, at which point the heaters were ramped down and test concluded. The two baseline tests completed reflect conditions prototypic to the full scale plant at both normal operation, and design basis Depressurized Loss of Forced Cooling (DLOFC) conditions. These parameters are defined by full RCCS decay heat loads of 1.4 MWt and 2.1 MWt, which translate to NSTF scale thermal power levels of 34.4 kWt and 51.6 kWt, respectively. Two additional tests were also conducted, which repeated baseline cases but included modification by single parameter variations in decay heat load, or heater power, at 700 kWt, and 2.8 MWt. Finally, a scoping test was performed to demonstrate the facilities ability to operate in a two-phase mode of operation. The facility successfully operated for a 3-hour period at this mode, reaching peak inventory temperatures of 105°C and condensing approximately 50 gallons of boil-off inventory. The testing parameters for this demonstration were based off the design basis DLOFC condition, and will be the reference for new baseline cases as the program moves forward with two-phase matrix testing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

LONG TERM OPERATION ISSUES FOR ELECTRICAL CABLE SYSTEMS IN NUCLEAR POWER PLANTS

Nuclear power plants contain hundreds of kilometers of electrical cables including cables used for power, for instrumentation, and for control. It is essential that safety-related cable systems continue to perform following a design-basis event. Wholesale replacement of electrical cables in existing plants facing licensing period renewal is both impractical and cost-prohibitive. It is therefore important to understand the long term aging of cable materials to have confidence that aged cables will perform when needed. It is equally important in support of cable aging management to develop methods to evaluate the health of installed cables and inform selective cable replacement decisions. The most common insulation materials for electrical cables in nuclear power plants are cross-linked polyethylene and ethylene-propylene rubber. The mechanical properties of these materials degrade over time in the presence of environmental stresses including heat, gamma irradiation, and moisture. Mechanical degradation of cable insulation beyond a certain threshold is unacceptable because it can lead to insulation cracking, exposure of energized conductors, arcing and burning or loss of the ability of the cable system to function during a design-basis accident. While thermal-, radiation-, and moisture-related degradation of polymer insulation materials has been extensively studied over the last few decades, questions remain regarding the long-term performance of cable materials in nuclear plant-specific environments. Identified knowledge gaps include an understanding of the temperature-dependence of activation energies for thermal damage and an understanding of the synergistic effects of radiation and thermal stress on polymer degradation. Many of the outstanding questions in the aging behavior of cable materials relate to the necessity of predicting long-term field degradation using accelerated aging results from the laboratory. Materials degrade faster under more extreme conditions, but extension of behavior to long term degradation under more mild conditions, such as those experienced by most installed cables in nuclear power plants, is complicated by the fact that different degradation mechanisms may be involved in extreme and mild scenarios. The discrepancy in predicted results from short term, more extreme exposure and actual results from longer term, more mild exposures can be counter intuitive. For instance, due to the attenuation of oxidation penetration in material samples rapidly aged through exposure to high temperatures, the bulk of the samples may be artificially protected from thermal aging. In another example, simultaneous exposure of cable insulation material to heat and radiation may actually lead to less damage at higher temperatures than may be observed at lower temperatures. The Light Water Reactor Sustainability program of the United States (US) Department of Energy Office (DOE) of Nuclear Energy is funding research to increase the predictive understanding of electrical cable material aging and degradation in existing nuclear power plants in support of continued safe operation of plants beyond their initial license periods. This research includes the evaluation and development of methods to assess installed cable condition.

Fifield, Dr Leonard S↗