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DOE OSTI2012

Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites

This plan incorporates U.S. Department of Energy (DOE) Office of Legacy Management (LM) standard operating procedures (SOPs) into environmental monitoring activities and will be implemented at all sites managed by LM. This document provides detailed procedures for the field sampling teams so that samples are collected in a consistent and technically defensible manner. Site-specific plans (e.g., long-term surveillance and maintenance plans, environmental monitoring plans) document background information and establish the basis for sampling and monitoring activities. Information will be included in site-specific tabbed sections to this plan, which identify sample locations, sample frequencies, types of samples, field measurements, and associated analytes for each site. Additionally, within each tabbed section, program directives will be included, when developed, to establish additional site-specific requirements to modify or clarify requirements in this plan as they apply to the corresponding site. A flowchart detailing project tasks required to accomplish routine sampling is displayed in Figure 1. LM environmental procedures are contained in the Environmental Procedures Catalog (LMS/PRO/S04325), which incorporates American Society for Testing and Materials (ASTM), DOE, and U.S. Environmental Protection Agency (EPA) guidance. Specific procedures used for groundwater and surface water monitoring are included in Appendix A. If other environmental media are monitored, SOPs used for air, soil/sediment, and biota monitoring can be found in the site-specific tabbed sections in Appendix D or in site-specific documents. The procedures in the Environmental Procedures Catalog are intended as general guidance and require additional detail from planning documents in order to be complete; the following sections fulfill that function and specify additional procedural requirements to form SOPs. Routine revision of this Sampling and Analysis Plan will be conducted annually at the beginning of each fiscal year when attachments in Appendix D, including program directives and sampling location/analytical tables, will be reviewed by project personnel and updated. The sampling location/analytical tables in Appendix D, however, may have interim updates according to project direction that are not reflected in this plan. Deviations from location/analytical tables in Appendix D prior to sampling will be documented in project correspondence (e.g., startup letters). If significant changes to other aspects of this plan are required before the annual update, then the plan will be revised as needed.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Information findability: An informal study to explore options for improving information findability for the systems analysis group

A Systems Analysis Group has existed at Sandia National Laboratories since at least the mid-1950s. Much of the group s work output (reports, briefing documents, and other materials) has been retained, along with large numbers of related documents. Over time the collection has grown to hundreds of thousands of unstructured documents in many formats contained in one or more of several different shared drives or SharePoint sites, with perhaps five percent of the collection still existing in print format. This presents a challenge. How can the group effectively find, manage, and build on information contained somewhere within such a large set of unstructured documents? In response, a project was initiated to identify tools that would be able to meet this challenge. This report documents the results found and recommendations made as of August 2013.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2014

AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC-3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC-3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. The report documents qualification of AGC-3 experiment irradiation monitoring data following MCP-2691. This report also documents whether AGC-3 experiment irradiation monitoring data meet the requirements for data collection as specified in technical and functional requirements documents and quality assurance (QA) plans. Data handling is described showing how data are passed from the data collection experiment to the Nuclear Data Management and Analysis System (NDMAS) team. The data structure is described, including data batches, components, attributes, and response variables. The description of the approach to data qualification includes the steps taken to qualify the data and the specific tests used to verify that the data meet requirements. Finally, the current status of the data received by NDMAS from the AGC-3 experiment is presented with summarized information on test results and resolutions. This report addresses all of the irradiation monitoring data collected during the AGC-3 experiment.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Criteria Considered in Selecting Feed Items for Americium-241 Oxide Production Operations

The analysis in this document serves the purpose of defining a number of attributes in selection of feed items to be utilized in recovery/recycle of Pu and also production operations of 241 AmO 2 material intended to meet specification requirements. This document was written in response to a specific request on the part of the 2014 annual program review which took place over the dates of October 28-29, 2014. A number of feed attributes are noted including: (1) Non-interference with existing Pu recovery operations; (2) Content of sufficient 241 Am to allow process efficiency in recovery operations; (3) Absence of indications that 243 Am might be mixed in with the Pu/ 241 Am material; (4) Absence of indications that Cm might be mixed in with the Pu/ 241 Am material; (5) Absence of indications of other chemical elements that would present difficulty in chemical separation from 241 Am; (6) Feed material not expected to present difficulty in dissolution; (7) Dose issues; (8) Process efficiency; (9) Size; (10) Hazard associated with items and package configuration in the vault; (11) Within existing NEPA documentation. The analysis in this document provides a baseline of attributes considered for feed materials, but does not presume to replace the need for technical expertise and judgment on the part of individuals responsible for selecting the material feed to be processed. This document is not comprehensive as regards all attributes that could prove to be important. The value of placing a formal QA hold point on accepting feed items versus more informal management of feed items is discussed in the summation of this analysis. The existing planned QA hold points on 241 AmO 2 products produced and packaged may be adequate as the entire project is based on QA of the product rather than QA of the process. The probability of introduction of items that would inherently cause the 241 AmO 2 products produced to be outside of specification requirements appears to be rather small.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Probabilistic Seismic Hazards Update for LLNL

Fugro Consultants, Inc. (FCL) completed the Probabilistic Seismic Hazard Analysis (PSHA) performed for Building 332 at the Lawrence Livermore National Laboratory (LLNL), near Livermore, CA. The study performed for the LLNL site includes a comprehensive review of recent information relevant to the LLNL regional tectonic setting and regional seismic sources in the vicinity of the site and development of seismic wave transmission characteristics. The Seismic Source Characterization (SSC), documented in Project Report No. 2259-PR-02 (FCL, 2015b), and Ground Motion Characterization (GMC), documented in Project Report No. 2259-PR-06 (FCL, 2015a) were developed in accordance with ANS/ANSI 2.29- 2008 Level 2 PSHA guidelines. The ANS/ANSI 2.29-2008 Level 2 PSHA framework is documented in Project Report No. 2259-PR-05 (FCL, 2016a). The Hazard Input Document (HID) for input into the PSHA developed from the SSC and GMC is presented in Project Report No. 2259-PR-04 (FCL, 2016b). The site characterization used as input for development of the idealized site profiles including epistemic uncertainty and aleatory variability is presented in Project Report No. 2259-PR-03 (FCL, 2015c). The PSHA results are documented in Project Report No. 2259-PR-07 (FCL, 2016c).

58 GEOSCIENCES↗
DOE OSTI2013

Emergency Response Capability Baseline Needs Assessment Compliance Assessment

This document is the second of a two-part analysis of Emergency Response Capabilities of Lawrence Livermore National Laboratory. The first part, 2013 Baseline Needs Assessment Requirements Document established the minimum performance criteria necessary to meet mandatory requirements. This second part analyses the performance of Lawrence Livermore Laboratory Emergency Management Department to the contents of the Requirements Document. The document was prepared based on an extensive review of information contained in the 2009 BNA, the 2012 BNA document, a review of Emergency Planning Hazards Assessments, a review of building construction, occupancy, fire protection features, dispatch records, LLNL alarm system records, fire department training records, and fire department policies and procedures.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Emergency Response Capability Baseline Needs Assessment - Compliance Assessment

This document was prepared by John A. Sharry, LLNL Fire Marshal and Division Leader for Fire Protection and was reviewed by LLNL Emergency Management Department Head, James Colson. This document is the second of a two-part analysis on Emergency Response Capabilities of Lawrence Livermore National Laboratory. The first part, 2016 Baseline Needs Assessment Requirements Document established the minimum performance criteria necessary to meet mandatory requirements. This second part analyses the performance of Lawrence Livermore Laboratory Emergency Management Department to the contents of the Requirements Document. The document was prepared based on an extensive review of information contained in the 2016 BNA, a review of Emergency Planning Hazards Assessments, a review of building construction, occupancy, fire protection features, dispatch records, LLNL alarm system records, fire department training records, and fire department policies and procedures. The 2013 BNA was approved by NNSA’s Livermore Field Office on January 22, 2014.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Functional Requirements for an Electronic Work Package System

This document provides a set of high level functional requirements for a generic electronic work package (eWP) system. The requirements have been identified by the U.S. nuclear industry as a part of the Nuclear Electronic Work Packages - Enterprise Requirements (NEWPER) initiative. The functional requirements are mainly applied to eWP system supporting Basic and Moderate types of smart documents, i.e., documents that have fields for recording input such as text, dates, numbers, and equipment status, and documents which incorporate additional functionalities such as form field data “type“ validation (e.g. date, text, number, and signature) of data entered and/or self-populate basic document information (usually from existing host application meta data) on the form when the user first opens it. All the requirements are categorized by the roles; Planner, Supervisor, Craft, Work Package Approval Reviewer, Operations, Scheduling/Work Control, and Supporting Functions. The categories Statistics, Records, Information Technology are also included used to group the requirements. All requirements are presented in Section 2 through Section 11. Examples of more detailed requirements are provided for the majority of high level requirements. These examples are meant as an inspiration to be used as each utility goes through the process of identifying their specific requirements. The report’s table of contents provides a summary of the high level requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Evaluation of Advanced Signal Processing Techniques to Improve Detection and Identification of Embedded Defects

By the end of 1996, 109 Nuclear Power Plants were operating in the United States, producing 22% of the Nation’s electricity [1]. At present, more than two thirds of these power plants are more than 40 years old. The purpose of the U.S. Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the operating lifetimes of nuclear power plants (NPPs) beyond 60 years [2]. The most important safety structures in an NPP are constructed of concrete. The structures generally do not allow for destructive evaluation and access is limited to one side of the concrete element. Therefore, there is a need for techniques and technologies that can assess the internal health of complex, reinforced concrete structures nondestructively. Previously, we documented the challenges associated with Non-Destructive Evaluation (NDE) of thick, reinforced concrete sections and prioritized conceptual designs of specimens that could be fabricated to represent NPP concrete structures [3]. Consequently, a 7 feet tall, by 7 feet wide, by 3 feet and 4-inch-thick concrete specimen was constructed with 2.257-inch-and 1-inch-diameter rebar every 6 to 12 inches. In addition, defects were embedded the specimen to assess the performance of existing and future NDE techniques. The defects were designed to give a mix of realistic and controlled defects for assessment of the necessary measures needed to overcome the challenges with more heavily reinforced concrete structures. Information on the embedded defects is documented in [4]. We also documented the superiority of Frequency Banded Decomposition (FBD) Synthetic Aperture Focusing Technique (SAFT) over conventional SAFT when probing defects under deep concrete cover. Improvements include seeing an intensity corresponding to a defect that is either not visible at all in regular, full frequency content SAFT, or an improvement in contrast over conventional SAFT reconstructed images. This report documents our efforts in four fronts: 1) Comparative study between traditional SAFT and FBD SAFT for concrete specimen with and without Alkali-Silica Reaction (ASR) damage, 2) improvement of our Model-Based Iterative Reconstruction (MBIR) for thick reinforced concrete [5], 3) development of a universal framework for sharing, reconstruction, and visualization of ultrasound NDE datasets, and 4) application of machine learning techniques for automated detection of ASR inside concrete. Our comparative study between FBD and traditional SAFT reconstruction images shows a clear difference between images of ASR and non-ASR specimens. In particular, the left first harmonic shows an increased contrast and sensitivity to ASR damage. For MBIR, we show the superiority of model-based techniques over delay and sum techniques such as SAFT. Improvements include elimination of artifacts caused by direct arrival signals, and increased contrast and Signal to Noise Ratio. For the universal framework, we document a format for data storage based on the HDF5 file format, and also propose a modular Graphic User Interface (GUI) for easy customization of data conversion, reconstruction, and visualization routines. Finally, two techniques for ASR automated detection are presented. The first technique is based on an analysis of the frequency content using Hilbert Transform Indicator (HTI) and the second technique employees Artificial Neural Network (ANN) techniques for training and classification of ultrasound data as ASR or non-ASR damaged classes. The ANN technique shows great potential with classification accuracy above 95%. These approaches are extensible to the detection of additional reinforced, thick concrete defects and damage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Roadmap for the international, accelerator-based neutrino programme

In line with its terms of reference the ICFA Neutrino Panel has developed a roadmap for the international, accelerator-based neutrino programme. A "roadmap discussion document" was presented in May 2016 taking into account the peer-group-consultation described in the Panel's initial report. The "roadmap discussion document" was used to solicit feedback from the neutrino community---and more broadly, the particle- and astroparticle-physics communities---and the various stakeholders in the programme. The roadmap, the conclusions and recommendations presented in this document take into account the comments received following the publication of the roadmap discussion document. With its roadmap the Panel documents the approved objectives and milestones of the experiments that are presently in operation or under construction. Approval, construction and exploitation milestones are presented for experiments that are being considered for approval. The timetable proposed by the proponents is presented for experiments that are not yet being considered formally for approval. Based on this information, the evolution of the precision with which the critical parameters governinger the neutrino are known has been evaluated. Branch or decision points have been identified based on the anticipated evolution in precision. The branch or decision points have in turn been used to identify desirable timelines for the neutrino-nucleus cross section and hadro-production measurements that are required to maximise the integrated scientific output of the programme. The branch points have also been used to identify the timeline for the R&D required to take the programme beyond the horizon of the next generation of experiments. The theory and phenomenology programme, including nuclear theory, required to ensure that maximum benefit is derived from the experimental programme is also discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Test Plan for the Wake Steering Experiment at the Scaled Wind Farm Technology (SWiFT) Facility.

This document is a test plan describing the objectives, configuration, procedures, reporting, roles, and responsibilities for conducting the joint Sandia National Laboratories and National Renewable Energy Laboratory Wake Steering Experiment at the Sandia Scaled Wind Farm Technology (SWiFT) facility near Lubbock, Texas in 2016 and 2017 . The purpose of this document is to ensure the test objectives and procedures are sufficiently detailed such that al l involved personnel are able to contribute to the technical success of the test. This document is not intended to address safety explicitly which is addressed in a separate document listed in the references titled Sandia SWiFT Facility Site Operations Manual . Both documents should be reviewed by all test personnel.

17 WIND ENERGY↗
DOE OSTI2017

Fermilab HEPCloud Facility Decision Engine Design

The Decision Engine is a critical component of the HEP Cloud Facility. It provides the functionality of resource scheduling for disparate resource providers, including those which may have a cost or a restricted allocation of cycles. Along with the architecture, design, and requirements for the Decision Engine, this document will provide the rationale and explanations for various design decisions. In some cases, requirements and interfaces for a limited subset of external services will be included in this document. This document is intended to be a high level design. The design represented in this document is not complete and does not break everything down in detail. The class structures and pseudo-code exist for example purposes to illustrate desired behaviors, and as such, should not be taken literally. The protocols and behaviors are the important items to take from this document. This project is still in prototyping mode so flaws and inconsistencies may exist and should be noted and treated as failures.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site (Rev.3)

U.S. Department of Energy (DOE) Order 435.1, Radioactive Waste Management, requires that performance assessments (PAs) and composite analyses (CAs) for low-level radioactive waste (LLW) disposal facilities be maintained by the responsible field office. This plan describes the activities performed to maintain the PA and the CA for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada National Security Site (NNSS). The plan is based on DOE Order 435.1 Radioactive Waste Management (DOE 2007), DOE Manual 435.1-1, Radioactive Waste Management Manual (DOE 2011), and DOE Standard DOE-STD-5002-2017, Disposal Authorization Statement and Tank Closure Documentation (DOE 2017). The facility’s Disposal Authorization Statement (DAS) specifies conditions for the continuing operation of a LLW facility. The DAS is issued based on a review of the facility's PA, CA, the maintenance plan, preliminary closure plan, and preliminary monitoring plan. The DAS specifies the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. The Maintenance Plan uses an Unreviewed Disposal Question (UDQ)/Unreviewed Composite Analysis Question (UCAQ) change control process to identify, evaluate, and control planned or discovered changes affecting the PA, CA, DAS, or other documents composing the Radioactive Waste Management Basis (RWMB). The UDQ/UCAQ process uses the PA and CA GoldSim models to calculate updated results, assess the significance of changes, and to assess the need to revise the PA, CA, or DAS. In addition to the change control process, the Maintenance Plan requires an annual review of all changes and an annual assessment of the need to revise RWMB documents, plans, and programs. The cumulative effect of all changes is assessed annually by preparing a revised PA and CA baseline model that is used to update PA and CA results. The updated results are used to assess compliance with the PA performance objectives and the CA dose constraint, and to assess the need to revise the PAs, CAs, and other documents, plans, and programs supporting the RWMB. An Annual Summary Report documenting the results and conclusions of the annual review is prepared and submitted to DOE.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

FY2019 parameters for Gold ions in Booster, AGS, and RHIC

Sections 1 through 6 of this report give parameters for low-energy gold ions circulating in RHIC at injection and in AGS at extraction. Section 7 gives parameters for Au79+ ions circulating in RHIC at the standard injection magnetic rigidity of 81.11378003 Tm . Section 8 gives parameters for medium-energy Au79+ ions circulating in RHIC at 31.2 GeV per nucleon. The standard parameters for gold ions circulating in Booster, AGS, and RHIC are documented in. For convenience they are listed here in Sections 9 through 11 . Gold ions normally come from EBIS but can be provided by Tandem if EBIS is down for maintenance or repairs. Section 12 gives parameters for Tandem Au31+ ions circulating in Booster. In order to achieve a single bunch of Tandem ions at Booster extraction, a 6 to 3 to 1 bunch merge is required. This has been developed by Keith Zeno and is documented in. Different bunch merges and ll patterns in AGS are used depending on the desired bunch intensity in RHIC and the desired number of bunches to be injected into RHIC per AGS cycle. These are illustrated in Sections 13 through 17 . The standard bunch merges done in Booster and AGS are documented in. Bucket and bunch parameters for clean injection of low-energy gold ions into RHIC are documented in. Parameters of past medium and low energy gold ion setups are documented in.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Reactor Physics Experiment NEA Benchmark Possibilities for Measurements at ORCEF and Other DOE Facilities

This report documents a wide variety of critical and subcritical reactor physics measurements that could be included in the Nuclear Energy Agency (NEA) International Handbook of Evaluated Reactor Physics Benchmark Experiments. The experiments were performed by John T. Mihalczo and coworkers at many (8) DOE facilities including the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF between 1958 and 1975), Los Alamos National Laboratory, Pacific Northwest National Laboratory, and Babcock and Wilcox at Lynchburg, Virginia, and the Y-12 National Security Complex (the latter between 1975 and up to 2006). This report lists the various measurement programs and a limited description of the reactor physics measurement that were performed in each that could be benchmarks. Some criticality safety measurements at these facilities have already been documented in the NEA International Handbook of evaluated Criticality Safety Benchmark Experiments. Some of these reactor physics measurements have also been documented in the literature and this document give a complete list of references for which John T. Mihalczo was an author or co-author. The experiments are divided into four classes: critical experiments only, critical and subsequent subcritical experiments, subcritical experiments at critical facilities, and subcritical measurements performed in material balance areas of the Y-12 National Security Complex (NSC) and other facilities. Because of the high cost and large amount of facility time related to these measurements, it is very cost-effective to mine the existing data to produce NEA reactor physics benchmarks. The information for many of these measurements is documented in critical experiment and subcritical experiment logbooks at the International Criticality Safety Benchmark Program at Idaho National Laboratory and in Laboratory Records at Oak Ridge National laboratory.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

How To Manual - 4.56

The "how to" document is designed to help walk the analyst through difficult aspects of software usage. It should supplement both the User's manual and the Theory document, by providing examples and detailed discussion that reduce learning time for complex set ups. These documents are intended to be used together. We will not formally list all parameters for an input here — see the User's manual for this. All the examples in the "How To" document are part of the Sierra/SD test suite, and each will run with no modification. The nature of this document casts together a number of rather unrelated procedures. Grouping them is difficult. Please try to use the table of contents and the index as a guide in finding the analyses of interest.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Work Plan for the Site-wide SSHAC Level 3 Probabilistic Seismic Hazard Analysis at INL

This work plan provides the overall framework, key activities, key tasks, and schedule for the execution of a probabilistic seismic hazard analysis at Idaho National Laboratory (INL). The probabilistic seismic hazard analysis will follow processes and procedures appropriate for the Senior Seismic Hazard Analysis Committee (SSHAC) Level 3 study provided in U. S. Nuclear Regulatory Commission guidance, NUREG-2213. The SSHAC Level 3 probabilistic seismic hazard analysis (PSHA) is being conducted as an INL site-wide effort. The work plan was developed and revised by participants who will perform the SSHAC Level 3 PSHA. The work plan also includes the quality assurance requirements commensurate with American Society of Mechanical Engineers Nuclear Quality Assurance NQA 1 2008/2009 and 2011 Addendum for design nuclear facilities under the U.S. Department of Energy and U. S. Nuclear Regulatory Commission. The scope of the SSHAC Level 3 PSHA is designed to produce the mean-centered definition of the seismic hazard with the appropriate treatment of uncertainties supported by the associated technical justifications. The SSHAC Level 3 study will properly implement and completely document the activities of evaluation, integration, and documentation. The evaluation process will consider data, models, and methods proposed by the larger technical community that are relevant to the PSHA at INL. The integration process will build seismic-source and ground-motion characterization models that represent the center, body, and range of technically defensible interpretations, taking into consideration the outcomes of the evaluation process. Key tasks and activities—such as a project kickoff meeting, three workshops, four formal working meetings, and a briefing meeting with the Participatory Peer Review Panel—will occur for the seven-member technical integration teams to construct seismic-source and ground-motion-characterization models. Both preliminary and final hazard-input documents will be developed to perform sensitivity analyses and the final PSHA, respectively. The entire study will be thoroughly documented in draft and final reports. The Participatory Peer Review Panel will review the SSHAC Level 3 study from start to finish, including the draft and final INL Site-wide PSHA reports, so they can provide a closure letter documenting the study’s adherence to SSHAC processes and technical adequacy of the final products.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗