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At least 415 records · Page 23

Transmutation abilities of the SFR low void effect core concept 'CFV' 3600 MWth

This paper presents an evaluation of the potential of minor actinide transmutation in a 3600 MWth SFR core designed with the low void effect core concept (namely 'CFV concept'). This concept is based upon an axially heterogeneous design with an internal fertile zone, and two radial fuel zones with different heights. Two modes of minor actinide transmutation are considered. The homogeneous mode where the minor actinides (MA) are diluted in the fuel is studied considering different options: - MA diluted in the whole core, - MA diluted in the internal and external fuel zone, - MA diluted in the internal fertile zone, for which different isotopic vectors and contents in fuel are analyzed. The heterogeneous mode is also studied with MA placed in external blanket bearings, with contents of 20%. The results are compared to those obtained with a traditional homogenous core concept (SFRV2B type) in terms of transmutation performances. Impacts of the transmutation assumptions on transmutation performances, on fuel cycle and safety parameters (void effect, Doppler) are also presented. (authors)

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Electric actuator modification on a turbine driven feed pump at Calvert Cliffs nuclear power plant - 296

Feedwater systems are a leading contributor to unit SCRAMs as indicated by the 2015 INPO data, where 9 of the 51 SCRAMs recorded were a result of feedwater issues. Digital feedwater control systems and enhanced feed pump control modifications can greatly improve performance and control of the plant feedwater systems and help drive down the potential for reactivity management events and unit SCRAMs. Westinghouse has a strong background in the implementation of digital feedwater control systems, and in planning for a recently completed feedwater control system upgrade at Calvert Cliffs, the Westinghouse team worked with Calvert Cliffs to design a solution to a number of common operational issues. Primarily, improving feedwater control over the entire operating range, eliminating single points of vulnerability (SPV) in the feedwater/feed pump systems and eliminating operational issues with the feed pump steam admission valve hydraulic control oil system. A robust first-of-a-kind design and implementation program was undertaken to split the existing hydraulic valve gear actuator with separate electric actuators. Westinghouse worked with Calvert Cliffs and Emerson Process Management (EPM) to install and commission General Electric (GE) feed pump turbines with separate electric actuators for the high-pressure (HP) and low-pressure (LP) steam admission valves. This successful mechanical and control modification was a first for both the nuclear and non-nuclear power generation industries. Westinghouse is working with a number of other utilities to implement similar modifications for PWR and BWR feed pump applications. (authors)

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Nuclear power plant risk-informed surveillance frequency control program implementation at the Duke Catawba nuclear plant with a focus on instrumentation and control systems - 333

This paper summarizes an updated process for risk-informed surveillance frequency control program (SFCP) implementation at the Catawba Nuclear Power Plant in the past year. This program includes selection and prioritization of specific target surveillance test interval extensions; and development, review, and implementation of surveillance test risk informed documented evaluation (STRIDE) packages designed to support extension of conventional surveillance requirement test intervals, in accordance with 'Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guideline,' NEI 04-10, Revision 1 . The scope of work associated with STRIDE development includes probabilistic risk assessment (PRA) case studies, deterministic assessment (DA) evaluations, and, where required, instrument drift evaluation (IDE). The STRIDE implementation efforts have also included support of independent decision-making panel (IDP) meetings at the implementing power stations and IDP member training. The purpose of this paper is to provide a presentation of a refined process for STRIDE development with a focus on instrumentation and control systems based on author experience, which includes support for the development of multiple STRIDE packages for the Catawba Nuclear Plant. This paper outlines a framework for practical implementation of an SFCP within the context of an integrated risk-informed performance-based regulation application program with emphasis on instrumentation and control systems. (authors)

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Diablo Canyon power plant digital process protection system replacement licensing experience using ISG-06 - 395

This paper provides a summary of the preparation of a License Amendment Request (LAR) for replacement of the existing Diablo Canyon Power Plant (DCPP) Eagle 21 digital Process Protection System (PPS) by Pacific Gas and Electric Company (PG and E) using US Nuclear Regulatory Commission (USNRC) Digital I and C (DI and C) Interim Staff Guidance (ISG) document ISG-06, Licensing Process . The USNRC has designated the LAR for the DCPP PPS replacement project as the pilot application for use of DI and C ISG-06. DI and C ISG-06, Revision 1, issued for use on January 19, 2011, describes the licensing process the NRC staff may use to review an LAR for a DI and C modification. The LAR for the replacement of the Eagle 21 PPS was submitted to the NRC on October 26, 2011, was accepted for review on January 13, 2012, and was approved on December 21, 2016. A more detailed review of the specific PPS replacement design including the Defense-In-Depth and Diversity approach is provided in a paper presented at the 2015 conference. (authors)

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A Case Study in the Application of TSUNAMI-3D - Part 1, Multigroup

The use of the sensitivity and uncertainty tool suite within SCALE, known as the TSUNAMI suite, has grown following its initial introduction more than a decade ago. Proper use of these tools provides a more complete understanding of system responses to perturbations, especially in the area of nuclear cross section uncertainties. Recent advances within TSUNAMI have added generalized perturbation theory to the k{sub eff} sensitivity capability that was included in TSUNAMI originally. The generation of sensitivity data is encouraged in the review process for new critical experiment requests sponsored by the US Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP) to demonstrate that the proposed experiment will satisfy the identified data need. Many individuals participate in this process, but many analysts have not received proper training in the use of the TSUNAMI tools. The adjoint perturbation theory used to calculate sensitivities is not typically included in nuclear engineering curricula, leaving inexperienced engineers particularly at risk for incorrect use of the tools. The first step in the use of the TSUNAMI tools is the generation of a sensitivity data file (SDF), which contains all of the energy-dependent sensitivity data generated for a model. Most often, the SDF is generated from a three-dimensional (3D) model using either the KENO V.a or KENO-VI code within the TSUNAMI-3D sequence. The sensitivity data generated in TSUNAMI-3D must be accurate for correct quantification of system similarity or any other use of the sensitivity data. The sensitivities calculated by TSUNAMI can be affected by many of the input parameters, cross section processing models, and geometry modeling techniques used within the TSUNAMI-3D input. Fortunately, the total sensitivity of the most important nuclides can be confirmed simply by using direct perturbation (DP) calculations. Confirmation of the total sensitivity of these important nuclides provides confidence that the flux and importance function calculated within the TSUNAMI sequence will yield accurate sensitivities for all nuclides, reactions, and energies of interest. This process is essentially identical for the multigroup (MG) methods and the continuous-energy (CE) methods introduced in SCALE 6.2. This paper discusses the use of DP calculations and provides a case study to confirm that accurate sensitivities are generated by TSUNAMI-3D for a particular benchmark experiment. A companion paper provides a similar case study applying the CE TSUNAMI capability to the same critical experiment. (authors)

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National criticality experiments research center: Status and update - 14575

Four critical assembly machines (Comet, Planet, Flat-Top, and Godiva-IV) constitute the critical experiment capabilities at the National Criticality Experiments Research Center. All four machines are operational. Between these machines, systems ranging from the thermal through the intermediate to the fast spectrum may be assembled. Steady-State, transient, and super-prompt critical conditions may be explored. In this paper, the status and capabilities of NCERC are discussed. (authors)

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SPC-3178 Rev 5 Beryllium Oxide Procurement Specification

TREAT microreactor Experiment Cell (T-REXC). This facility will provide a test bed for microreactor developers to perform demonstration projects and critical experiments in an expedited manner. In addition to other systems, structures, and components (SSCs), T-REXC will provide beryllium oxide (BeO) reflector components that T-REXC users can incorporate into their equipment installation. This specification details supplier requirements associated with these BeO components.

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ECAR-5367 Rev 0 Hydraulic Assessment of the Primary Coolant Test System Design

A MARVEL thermal-hydraulic analysis team has developed the computational fluid dynamicsmodel of the Primary Coolant Test (PCAT) system design based on the CAD drawing providedfrom INL PCAT design team. This ECAR addresses the hydraulic system characteristics(pressure loss and flow distribution) in the various sub-components of the PCAT design. Thisreport provides potential design optimization direction on the PCAT test. Most importantly, thisreport evaluates the feasibility of the natural circulation at the targeted heat balance between thecore and heat sink design.

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ECAR-6573 Rev 0 PGS Stirling Engine Radiation Lifetime Estimate

This evaluation determines the expected service lifetime for the Stirling Engines in the Power Generation Subsystem (PGS) based on the radiation resistance of the organic materials, electrical components, and magnets, along with the projected radiation dose rate. Components that are not projected to last for a six-month operational period are discussed in more detail, including their associated risk due to failure and possible remedial actions.

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ECAR-6574 Rev 1 MARVEL Guard Vessel System FEA and ASME Analysis

This ECAR encompasses the design-by-equation requirements in ASME BPVC Section III Division 5 Subsection HCB, in addition to the service level and design-by-analysis requirements of Subsection HBB, as they apply to the Guard Vessel.

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ECAR-6576 Rev 0 MARVEL Secondary Support Structure Analysis

The Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor generates electricity using four Stirling cycle engines. Each Stirling engine, along with its mounting system, is called a Secondary Support Structure (SSS). An individual SSS consists of a Secondary Output Structure (SOS), a Secondary Support Structure Base, an insulation block, and assorted hardware. The SOS consists of the Stirling engine, the Secondary Output Structure Ring, a vibration isolator, and hardware. The purpose of this analysis is to show that the SSS is appropriately designed to meet structural, mechanical, and maintenance requirements for specified operating conditions.

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ECAR-6589 Rev 0 Reflector Support Structure Analysis

The purpose of this ECAR is to document compliance with best engineering practices, ANSI/AISC N690 and the seismic requirements of IBC-2015 for the Reflector Support Structure of the MARVEL microreactor.

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ECAR-7300 Rev 1 Verification and Validation of MCNP6.2 for MARVEL Neutronic Analysis

This report documents the verification and validation (V&V) efforts of the Monte Carlo N-Particle transport code (MCNP) version 6.2 on the Sawtooth supercomputer for the Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor required for the preliminary documented safety analysis. This document records V&V for a safety, hazards, analysis, and design software used for design and analysis of safety class structures, system and components (SSC)s.

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ECAR-7514 Rev 0 Analysis of the MARVEL Central Insurance Absorber Actuator Components

This ECAR analyzes the MARVEL Central Insurance Absorber (CIA) rod actuator design. Functions of interest include general vertical position, velocity, and SCRAM performance when carrying the CIA rod payload. Assessed components include the motor, drive train, electromagnet, bearings used for alignment and SCRAM velocity control, limit switches, 2nd position indicator, and resolver.

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ECAR-7517 Rev 1 MARVEL I&C Failure Modes and Effects Analysis

The purpose of this document is to perform a single-failure analysis through the use of a Failure Modes and Effects Analysis (FMEA) for the Safety Related components of the MARVEL Instrumentation and Control (I&C) System. The intent is that this document will meet the requirements for a single-failure analysis described in IEEE-379, “IEEE Standard for Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems”, to verify that this design does indeed meet the single failure criterion. Principles of IEEE-352, “IEEE Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Safety Systems” are followed to ensure the analysis is consistent with industry standards.

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ECAR-7547 Rev 0 MARVEL Intermediate Heat Exchanger Liner ASME Analysis

This ECAR analyzes the Secondary Coolant System liner, otherwise known as the Intermediate Heat Exchanger (IHX) Liner, as an ASME Section VIII, Division 1 pressure vessel. This analysis is only necessary in the event of a failure of the primary coolant boundary, in which case the IHX liner acts as a secondary confinement boundary to prevent primary coolant leakage. The liner main body or vessel is analyzed, as well as the liner flange ring.

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ECAR-8055 Rev 0 Verification and Validation of Star-CCM+ for Computational Fluid Dynamics Analyses for the MARVEL Microreactor

The objective of this Engineering Calculations and Analysis Report (ECAR) is to provide documentation and highlight relevant information regarding the verification and validation (V&V) of the commercial computational fluid dynamics (CFD) code STAR-CCM+ for the thermal and fluids analyses performed for the MARVEL microreactor.

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