Results for “Design Basis” — Page 9

DOE OSTI2015

Response Time Analysis and Test of Protection System Instrument Channels for APR1400 and OPR1000

Safety limits are required to maintain the integrity of physical barriers designed to prevent the uncontrolled release of radioactive materials in nuclear power plants. The safety analysis establishes two critical constraints that include an analytical limit in terms of a measured or calculated variable, and a specific time after the analytical limit is reached to begin protective action. Keeping with the nuclear regulations and industry standards, satisfying these two requirements will ensure that the safety limit will not be exceeded during the design basis event, either an anticipated operational occurrence or a postulated accident. Various studies on the setpoint determination methodology for the safety-related instrumentation have been actively performed to ensure that the requirement of the analytical limit is satisfied. In particular, the protection setpoint methodology for the advanced power reactor 1400 (APP1400) and the optimized power reactor 1000 (OPR1000) has been recently developed to cover both the design basis event and the beyond design basis event. The developed setpoint methodology has also been quantitatively validated using specific computer programs and setpoint calculations. However, the safety of nuclear power plants cannot be fully guaranteed by satisfying the requirement of the analytical limit. In spite of the response time verification requirements of nuclear regulations and industry standards, it is hard to find the studies on the systematically integrated methodology regarding the response time evaluation. In cases of APR1400 and OPR1000, the response time analysis for the plant protection system is partially included in the setpoint calculation and the response time test is separately performed via the specific plant procedure. The test technique has a drawback which is the difficulty to demonstrate completeness of timing test. The analysis technique has also a demerit of resulting in extreme times that not actually possible. Thus, the establishment of the systematic response time evaluation methodology is needed to justify the conformance to the response time requirement used in the safety analysis. This paper proposes the response time evaluation methodology for APR1400 and OPR1000 using the combined analysis and test technique to confirm that the plant protection system can meet the analytical response time assumed in the safety analysis. In addition, the results of the quantitative evaluation performed for APR1400 and OPR1000 are presented in this paper. The proposed response time analysis technique consists of defining the response time requirement, determining the critical signal path for the trip parameter, allocating individual response time to each component on the signal path, and analyzing the total response time for the trip parameter, and demonstrates that the total analyzed response time does not exceed the response time requirement. The proposed response time test technique is composed of defining the response time requirement, determining the critical signal path for the trip parameter, determining the test method for each component on the signal path, performing the response time test, and demonstrates that the total test result does not exceed the response time requirement. The total response time should be tested in a single test that covers from the sensor to the final actuation device on the instrument channel. When the total channel is not tested in a single test, separate tests on groups of components or single components including the total instrument channel shall be combined to verify the total channel response. For APR1400 and OPR1000, the ramp test technique is used for the pressure and differential pressure transmitters and the step function testing technique is applied to the signal processing equipment and final actuation device. As a result, it can be demonstrated that the response time requirement is satisfied by the combined analysis and test technique. Therefore, the proposed methodology in this paper plays a crucial role in guaranteeing the safety of the nuclear power plants systematically satisfying one of two critical requirements from the safety analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimize the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new perspective in the post Fukushima -accident era. Accidents at the Fukushima Daiichi reactors in the aftermath of the devastating earthquake and tsunami of March 11, 2011 have slowed down the nuclear renaissance world-wide and may have accelerated decommissioning either because some countries have decided to halt or reduce nuclear, or because the new safety requirements may reduce life-time extensions. Even in countries such as the UK and France that favor nuclear energy production existing nuclear sites are more likely to be chosen as sites for future NPPs. Even as the site recovery efforts continue at Fukushima and any decommissioning decisions are farther into the future, the accidents have focused attention on the reactor designs in general and specifically on the Fukushima type BWRs. The regulatory authorities in many countries have initiated a re-examination of the design of the systems, structures and components and considerations of the capability of the station to cope with beyond-design basis events. Enhancements to SSCs and site features for the existing reactors and the reactors that will be built will also impact the decommissioning phase activities. The newer reactor designs of today not only have enhanced safety features but also take into consideration the features that will facilitate future decommissioning. Lessons learned from past management and operation of reactors as well as the lessons from decommissioning are incorporated into the new designs. However, in the post-Fukushima era, the emphasis on beyond-design-basis capability may lead to significant changes in SSCs, which eventually will also have impact on the decommissioning phase. Additionally, where some countries decide to phase out the nuclear power, many reactors may enter the decommissioning phase in the coming decade. While the formal updating and expanding of existing guidance documents for accident cleanup and decommissioning would benefit by waiting until the Fukushima project has progressed sufficiently for that experience to be reliably interpreted, the development of structured on-line sharing of information and especially the creation of an on-line compendium of methods, tools, and techniques by which damaged fuel and other unique situations have been addressed can be addressed sooner and maintained as new problems and solutions arise and are resolved. The IAEA's new 'WEB 2.0 tool' CONNECT is expected to play a significant role in this and related information-sharing activities. The trend in some countries such as the United States has been to re-license the existing reactors for additional twenty years, beyond the original design life. Given the advances in technology over the past four decades, and considering that the newer designs incorporate significant improvements in safety systems, it may not be economical or technically feasible to retrofit enhancements into some of the older reactors. In such cases, the reactors may be retired from service and decommissioned. Overall, the energy demand in the world continues to rise, with sharp increases in the Asian countries, and nuclear power's role in the world's energy supply is expected to continue. Events at Fukushima have led to a re-examination on many fronts, including reactor design and regulatory requirements. Further changes may occur in these areas in the post-Fukushima era. These changes in turn will also impact the world-wide decommissioning scene and the decommissioning phase of the future reactors. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Quarterly Technical Progress Report Piperazine Advanced Stripper (PZAS™) Front End Engineering Design

EXECUTIVE SUMMARY This document summarizes the status of Cooperative Agreement DE-FE0031844, “Piperazine Advanced Stripper (PZAS™) Front-End Engineering Design,” during the reporting period of January 1 through March 31, 2020. The objective of this project is to develop accurate installed costs by conducting a Front-End Engineering Design (FEED) of PZAS™ at Golden Spread Electric Cooperative’s (GSEC) Mustang Station located in Denver City, TX. Complementary benefits include positioning the technology for a commercial project with the 45Q tax credits, qualifying PZAS™ for use on a Natural Gas Combined Cycle (NGCC) Cogen facility, and to provide cost detail to optimize PZAS™ and help guide R&D of second-generation solvent CO2 capture technologies. Results from the FEED will be used to evaluate the economic feasibility of the process at Mustang Station. This project is funded by the U.S. DOE National Energy Technology Laboratory under the aforementioned Cooperative Agreement. Exxon, Total, Chevron, UOP-Honeywell, and the University of Texas (UT) are project co-funders. AECOM and Trimeric are project team members; UT is the prime contractor. Summary of Progress Cooperative Agreement DE-FE0031844 was established in October 2019. The current reporting period, January 1 through March 31, 2020, is the second technical progress reporting period for the project. Several milestones were accomplished during this reporting period, including: • Kickoff Meeting with DOE on February 3, 2020. • Kickoff Meeting with GSEC on March 30, 2020. (Note: Due to Covid-19 travel restrictions and shelter-in-place guidelines, the kickoff meeting was conducted remotely via videoconferencing. See attached notes from that telecon.) • Updated Project Management Plan March 2020, submitted with this quarterly report Other activities during the quarter included progress on contracting and other legal agreements (e.g., non-disclosure agreements), internal kickoff meetings at both AECOM and Trimeric, and development of a Technical Implementation Plan (TIP). The TIP will help the team to make critical, early process decisions and, ultimately, to develop a project and process design basis. Note that all agreements between project team participants are complete as of this submittal, except the vendor agreement with Kiewit (steam cycle modeling). Plans for Next Reporting Period Activities during the next reporting period (April 1, 2020 through June 30, 2020) include: the completion of the Project Design Basis and progress towards the Process Design Basis/Process Design Package (PDP). The Project Design Basis is due as a deliverable and milestone during the next reporting period.

Rochelle, Gary T.↗

DOE OSTI2015

NSR&D Program Fiscal Year (FY) 2015 Call for Proposals Mitigation of Seismic Risk at Nuclear Facilities using Seismic Isolation

Seismic isolation (SI) has the potential to drastically reduce seismic response of structures, systems, or components (SSCs) and therefore the risk associated with large seismic events (large seismic event could be defined as the design basis earthquake (DBE) and/or the beyond design basis earthquake (BDBE) depending on the site location). This would correspond to a potential increase in nuclear safety by minimizing the structural response and thus minimizing the risk of material release during large seismic events that have uncertainty associated with their magnitude and frequency. The national consensus standard America Society of Civil Engineers (ASCE) Standard 4, Seismic Analysis of Safety Related Nuclear Structures recently incorporated language and commentary for seismically isolating a large light water reactor or similar large nuclear structure. Some potential benefits of SI are: 1) substantially decoupling the SSC from the earthquake hazard thus decreasing risk of material release during large earthquakes, 2) cost savings for the facility and/or equipment, and 3) applicability to both nuclear (current and next generation) and high hazard non-nuclear facilities. Issue: To date no one has evaluated how the benefit of seismic risk reduction reduces cost to construct a nuclear facility. Objective: Use seismic probabilistic risk assessment (SPRA) to evaluate the reduction in seismic risk and estimate potential cost savings of seismic isolation of a generic nuclear facility. This project would leverage ongoing Idaho National Laboratory (INL) activities that are developing advanced (SPRA) methods using Nonlinear Soil-Structure Interaction (NLSSI) analysis. Technical Approach: The proposed study is intended to obtain an estimate on the reduction in seismic risk and construction cost that might be achieved by seismically isolating a nuclear facility. The nuclear facility is a representative pressurized water reactor building nuclear power plant (NPP) structure. Figure 1: Project activities The study will consider a representative NPP reinforced concrete reactor building and representative plant safety system. This study will leverage existing research and development (R&D) activities at INL. Figure 1 shows the proposed study steps with the steps in blue representing activities already funded at INL and the steps in purple the activities that would be funded under this proposal. The following results will be documented: 1) Comparison of seismic risk for the non-seismically isolated (non-SI) and seismically isolated (SI) NPP, and 2) an estimate of construction cost savings when implementing SI at the site of the generic NPP.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program (Rev. 2)

The beyond design basis accident at Fukushima Daiichi in March 2011 and the Three Mile Island accident in 1979 imply that severe nuclear accidents may occur at somewhat higher frequencies than previously predicted, and that the financial liabilities of such accidents can cripple a utility. While current fuel is more than adequate for design basis accidents, the potential for harm from beyond design basis accidents (BDBA) has been a driver for Westinghouse development efforts since 2003. After Fukushima, other fuel vendors also began pursuing new fuel materials that can provide significant increases in the time, enabling reactor operators to respond to unforeseen events before significant releases of fuel materials and fission products occur. In order to be commercially successful, accident tolerant fuel (ATF) products must also provide significant fuel cost savings, operating cost improvements as well as safety improvements. This report documents the engineering, design and testing that occurred from October 1, 2016, through September 30, 2018, by the Westinghouse team. In addition, it includes: 1. The results of modeling and simulation work completed to date that supports technical feasibility of the ATF concept in meeting the goals of the enhanced ATF program including a core level analysis predicting how an existing LWR will perform when loaded with ATF. 2. Regulatory Analysis showing path to LTA/LTR acceptance. 3. Business Analysis showing commercial viability of the ATF concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Moving Beyond Traditional Tornado Missile Impact Analysis

On March 11, 2011 a 9.0 magnitude earthquake occurred off the coast of Japan creating a tsunami, which together severely damaged northeastern Japan including the Fukushima Dai-ichi nuclear power plant site. In response to this event, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events'. In order to meet the guidance of the NRC Order EA-12- 049, the a central U.S. nuclear generating station (Plant) was required to demonstrate that the currently installed Borated Water Storage Tank (BWST), Refueling Water Tank (RWT), and Qualified Condensate Storage Tank (QCST) could provide the necessary capacity following a Beyond Design Basis External Event (BDBEE). These tanks were not originally designed to meet current site design basis requirements for tornado wind and missile impacts. To save the Plant from having to build costly new storage tanks off site, the existing tanks were credited as robust for tornado missile impact through advanced finite element impact analyses using the ANSYS Autodyn solver and fluid-structure interaction (FSI). This procedure is the first known NRC acceptance of advanced analysis and FSI for to satisfy the FLEX tornado missile impact criteria. Extensive verification and validation (V and V) was performed to verify perforation of steel plates which included FSI. These V and V analyses were confirmed against large databases of experimental and theoretical results to ensure accurate modeling and results. The use of this advanced analytical method benefitted the Plant in several ways. A new storage tank was not required to be built, saving not only initial construction costs but also ongoing maintenance costs and monitoring activities. By determining specific vulnerable components of each tank a set of modifications and protection schemes were developed. These modifications were much quicker to implement saving the Plant on downtime costs. (authors)

42 ENGINEERING↗

DOE OSTI2014

Physical Property Measurements of Laboratory Prepared Saltstone Grout

The Saltstone Production Facility (SPF) built two new Saltstone Disposal Units (SDU), SDU 3 and SDU 5, in 2013. The variable frequency drive (VFD) for the grout transfer hose pump tripped due to high current demand by the motor during the initial radioactive saltstone transfer to SDU 5B on 12/5/2013. This was not observed during clean cap processing on July 5, 2013 to SDU 3A, which is a slightly longer distance from the SPF than is SDU 5B. Saltstone Design Authority (SDA) is evaluating the grout pump performance and capabilities to transfer the grout processed in SPF to SDU 3/5. To assist in this evaluation, grout physical properties are required. At this time, there are no rheological data from the actual SPF so the properties of laboratory prepared samples using simulated salt solution or Tank 50 salt solution will be measured. The physical properties of grout prepared in the laboratory with de-ionized water (DI) and salt solutions were obtained at 0.60 and 0.59 water to premix (W/P) ratios, respectively. The yield stress of the DI grout was greater than any salt grout. The plastic viscosity of the DI grout was lower than all of the salt grouts (including salt grout with admixture). When these physical data were used to determine the pressure drop and fluid horsepower for steady state conditions, the salt grouts without admixture addition required a higher pressure drop and higher fluid horsepower to transport. When 0.00076 g Daratard 17/g premix was added, both the pressure drop and fluid horsepower were below that of the DI grout. Higher concentrations of Daratard 17 further reduced the pressure drop and fluid horsepower. The uncertainty in the single point Bingham Plastic parameters is + 4% of the reported values and is the bounding uncertainty. Two different mechanical agitator mixing protocols were followed for the simulant salt grout, one having a total mixing time of three minutes and the other having a time of 10 minutes. The Bingham Plastic parameters were essentially the same for the salt grout without admixture. When Daratard 17 was added, the Bingham Plastic yield stress increased for the 10 minute mix. The simulant salt used in this task had similar physical properties of the Tank 50 3Q13 salt grout and is recommended for future use, if the salt solution in Tank 50 does not change. The design basis physical properties used to size the pumps and mixers at SPF were obtained from DPST-85-312. The grouts characterized in this report are bounded by the design basis density and Bingham Plastic yield stress. The opposite is true for the plastic viscosity. Steady state pressure drop calculations were performed for the design basis values using the flow rate for the clean cap and salt grouts and they bound the pressure drop of the grouts characterized in this report. A comparison of the lab prepared samples to PI ProcessBook data, specifically average pressure drop, indicate that the lab prepared samples are more viscous in nature than what is processed in the facility. This difference could be due to the applied shear rates which could be lower in the lab as compared to the facility and that fact the SPF added flush water, making this comparison more difficult. A perfunctory review of the PI ProcessBook data was discussed. It may be possible that the frequency that the distributed control system alters the grout pump speed to maintain grout hopper volume can negatively affect the efficiency of the grout pump.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Design Load Basis Guidance for Distributed Wind Turbines

Aeroelastic modeling (AM) is the primary methodology for structural and performance assessment of any wind turbine. Nonetheless, the use of AM in the distributed wind (DW) industry sector is limited due to several challenges (Damiani, Davis, & Summerville, 2022). One of these challenges lies in the perceived complexity of generating a proper set of numerical simulations to extract and process the key outputs for component design and verification, and, ultimately, achieve certification. This makes it difficult to reliably predict the structural and performance response of small wind turbines. From the investigation carried out in (Damiani & Davis, 2022), it is apparent that many stakeholders in this sector believe that a comprehensive guide for developing a design load basis (DLB) for distributed wind turbines (DWTs) is necessary.

17 WIND ENERGY↗

DOE OSTI2012

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. Key building design elements include (1) below grade siting to minimize vulnerability to aircraft impact, (2) multiple natural circulation decay heat rejection chimneys, (3) seismic base isolation, and (4) decay heat powered back-up electricity generation.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2017

Thermal properties of simulated Hanford waste glasses

Abstract The Hanford Tank Waste Treatment and Immobilization Plant will vitrify the mixed hazardous wastes generated from 45 years of plutonium production at the Hanford Site in Washington State. The molten glasses will be poured into stainless steel containers or canisters and subsequently cooled for storage and disposal. For appropriate facility design and operations to handle such highly energy‐consuming processes, knowledge of the material properties is required. The thermal properties (heat capacity, thermal diffusivity, and thermal conductivity) of representative high‐level and low‐activity waste glasses were studied as functions of temperature in the range of 200°C‐800°C (relevant to the cooling process). Simultaneous differential scanning calorimetry‐thermal gravimetry (DSC‐TGA), Xe‐flash diffusivity, pycnometry, and dilatometry were implemented. The study showed that simultaneous DSC‐TGA would be a reliable method for obtaining the heat capacity of various glasses in the temperature range of interest. Accurate thermal properties from this study were shown to provide a more realistic guideline for capacity and time constraints of the heat removal process when compared to the original conservative design‐basis engineering estimates. The estimates, though useful for design in the absence of measured physical properties, can now be supplanted and the measured thermal properties can be used in design verification activities.

Rodriguez, Carmen P.↗

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 OSTI2011

DETERMINATION OF REPORTABLE RADIONUCLIDES FOR DWPF SLUDGE BATCH 6 (MACROBATCH 7)

The Waste Acceptance Product Specifications (WAPS) 1.2 require that 'The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115'. As part of the strategy to comply with WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the heel from Sludge Batch 5 (SB5) with H-Canyon Np transfers completed after the start of processing SB5, and Sludge Batch 6 (SB6) that was transferred to Tank 40 from Tank 51. The blend of sludge in Tank 40 is also referred to as Macrobatch 7 (MB7). This report develops the list of reportable radionuclides and associated activities as a function of time. The DWPF will use this list and the activities as one of the inputs for the development of the Production Records that relate to radionuclide inventory. This work was initiated through Technical Task Request (TTR) HLW-DWPF-TTR-2009-0014; Rev. 2 entitled Sludge Batch 6 SRNL Shielded Cells Testing. Specifically, this report details results from performing Subtask III, Item 2 of the TTR and, in part, meets Deliverable 7 of the TTR. The work was performed following the Task Technical and Quality Assurance Plan (TTQAP), SRNL-RP-2009-00473, Rev. 15 and Analytical Study Plan (ASP), SRNL-RP-2009-00474, Rev. 1. In order to determine the reportable radionuclides for SB6 (MB7), a list of radioisotopes that may meet the criteria as specified by the Department of Energy's (DOE) WAPS was developed. All radioactive U-235 fission products and all radioactive activation products that could be in the SRS HLW were considered. In addition, all U and Pu isotopes identified in WAPS 1.6 were included in the list. This list was then evaluated and some isotopes were excluded from the projection calculations. Based on measurements and analytical detection limits, 30 radionuclides have been identified as reportable for DWPF SB6 as specified by WAPS 1.2. The WCP and WQR require that all of the radionuclides present in the Design Basis glass be considered as the initial set of reportable radionuclides. For SB6, all of the radionuclides in the Design Basis glass are reportable except for three radionuclides: Pd-107, Cs-135, and Th-230. At no time during the 1100-year period between 2015 and 3115 did any of these three radionuclides contribute to more than 0.01% of the radioactivity on a Curie basis. Two additional uranium isotopes (U-235 and -236) must be added to the list of reportable radionuclides in order to meet WAPS 1.6. All of the Pu isotopes (Pu-238, -239, -240, -241, and -242) and other U isotopes (U-233, -234, and -238) identified in WAPS 1.6 were already determined to be reportable according to WAPS 1.2 This brings the total number of reportable radionuclides for SB6 to 32. The radionuclide measurements made for SB6 are the most extensive conducted to date. Some method development/refinement occurred during the conduct of these measurements, leading to lower detection limits and more accurate measurement of some isotopes than was previously possible. Improvement in the analytical measurements will likely continue, and this in turn should lead to improved detection limit values for some radionuclides and actual measurements for still others.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2021

Demonstrate Capability of NEAMS Tools to Generate Reactor Kinetics Parameters for Pebble-Bed HTGRs Transient Modeling

The system analysis of anticipated operating occurrences and design basis accident for pebblebed reactor systems requires knowledge of the neutron kinetic parameters. These parameters take into account various types of feedback from the pebble-bed core and are typically integrated into system analysis tools such as the System Analysis Module (SAM) in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) tool package. These parameters are typically generated by a higher fidelity full-core coupled neutronics/thermal fluids analysis, such as using Griffin and Pronghorn, in 2D or 3D. The delayed neutron fractions (beta) and the neutron lifetimes (lambda) are typically generated by applying the adjoint solution of the neutron flux. Temperature reactivity coefficients are the most important feedback for all anticipated operating occurrences and design basis accident, then depending on the transient other effects might be important as rector control system worth, s-curves, and Xenon generation. This work package will establish a methodology and process for calculating the reactor kinetics parameters for high-temperature gas-cooled reactor and provide them in a suitable form for system level analyses of typical anticipated operating occurrences and design basis accident.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1951

Performance of single-stage compressor designed on basis of constant total enthalpy with symmetrical velocity diagram at all radii and velocity ratio of 0.7 at rotor hub / Jack R. Burtt and Robert J. Jackson

A typical inlet axial-flow compressor inlet stage, which was designed on the basis of constant total enthalpy with symmetrical velocity diagram at all radii, was investigated. At a tip speed of 1126 feet per second, a peak pressure ratio of 1.28 was obtained at an efficiency of 0.76. At a tip speed, the highest practical flow was 28 pounds per second per square foot frontal area with an efficiency of 0.78. Data for a rotor relative inlet Mach number range of from 0.5 to 0.875 indicates that the critical value for any stage radial element is approximately 0.80 for the stage investigated.

FANS↗

DOE OSTI2017

Impacts of common cause failure regulatory requirements on protection system architectures - 304

Regulatory-based design considerations for potential common cause failures (CCF) in digital instrumentation and control (I and C) systems have evolved over time and have impacted the development of protection system I and C architectures. The single failure criterion used to be the main design driver for a protection system I and C architecture and resulted in the familiar four-fold and three-fold system architectures. The single failure criterion has now been replaced by the conservative assumptions associated with digital CCF concerns (i.e., assumed digital CCF coincident with an anticipated operational transient or postulated accident) which are classified as 'beyond design basis events' by regulatory authorities. It is unusual for a beyond design basis event to be the controlling factor for a safety system design but that is now the reality for I and C system designers. The preferred regulatory solution is to introduce diversity into I and C system designs to guard against the general concern of digital CCF. However, the lack of clear criteria on how to define sufficient diversity (i.e., how much, what kind, and whether it is linked to identified digital CCF vulnerabilities) had led to more complex I and C architectures. The trend has been towards lengthy and more difficult reviews for the treatment of digital CCF vulnerabilities and I and C system architectures because of the subjective definition of digital CCF vulnerabilities and the lack of clear acceptance criteria for diversity strategy. Improvement in the treatment of digital CCF is needed to reverse the trend of increased I and C system architecture complexity and longer regulatory reviews. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗