ML18337A268
| ML18337A268 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 12/03/2018 |
| From: | Tennessee Valley Authority |
| To: | NRC/NRO/DLSE |
| Sutton M, Fetter A | |
| References | |
| Download: ML18337A268 (23) | |
Text
TVA Clinch River SMR Project Early Site Permit Application December 6th, 2018 Advisory Committee on Reactor Safeguards Full Committee Meeting
Acknowledgement and Disclaimer Acknowledgment: "This material is based upon work supported by the Department of Energy under Award Number DE-NE0008336."
Disclaimer: "This presentation was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof."
Advisory Committee on Reactor Safeguards l 2
Presentation Outline Clinch River Nuclear Site - Overview
- Dan Stout Early Site Permit Application - Overview
- Ray Schiele Emergency Preparedness
- Archie Manoharan Advisory Committee on Reactor Safeguards l 3
Clinch River Nuclear Site - Overview Dan Stout Director, Nuclear Technology & Innovation Advisory Committee on Reactor Safeguards l4
TVAs Mission Advisory Committee on Reactor Safeguards l 5
TVA Clinch River Site Site Access to 500 KV and 161 KV transmission Neighbor to DOE, an interested customer Basic Infrastructure Abundant and skilled workforce Strong community support TVA owned/controlled Advisory Committee on Reactor Safeguards l 6
Early Site Permit Application (ESPA)
An Early Site Permit assesses site suitability for potential construction and operation of a nuclear power plant.
Application includes:
Site Safety Analysis Report to address impacts of the environment on the plant Environmental Report Emergency Plans (Part 5A and Part 5B)
Exemptions (Part 6)
ESPA based on a plant parameter envelope (PPE)
Composite of reactor and engineered parameters from four U.S. light-water SMR designs with unique design features that bound the safety and environmental impact of plant construction and operation Developed based on NEI 10-01 guidance with margin added to specific parameters Assumes two or more SMR units of a single design Up to 800MWt for a single unit with a combined nuclear generating capacity not exceeding 2420 MWt (800 MWe)
Advisory Committee on Reactor Safeguards l 7
2017 NRC Review of ESPA 2018 2019 2020 ESPA ESPA Rev. 1 ESPA Rev. 2 FEIS FSER NRC Issues Accepted Submitted Planned Submittal ESP 12-30-16 12-15-17 Dec 18 Safety Review PSER SER w/ no OIs 8/4 10/20 FSER 8/17 Full ACRS Audits & RAIs 5/15 8/22 10/17 11/14 12/5 ACRS Subcomm. Meetings Scoping Environmental Review Meeting 5/15 Comment Period FEIS 6/21 Notice of Intent Audits & RAIs DEIS 4/13 4/26 4 Contentions 2 Contentions Commission Hearing(s)
Filed Admitted Ruling 6/12 10/10 5/3 Notice of Hearing, Commission Hearing Opportunity TVA ASLB Contested 4/4 Appeals Ruling Hearing Advisory Committee on Reactor Safeguards l 8 11/6 7/31 Terminated
ESPA Summary NRC Commenced Review in FY 17 Contains more than 8000 Pages Supported by over 80,000 pages in referenced documents Efficient Use of Audits Few Requests for Additional Information (RAIs)
Frequent, Clear, and Candid Communication Advisory Committee on Reactor Safeguards l 9
Early Site Permit- Overview Ray Schiele Licensing Manager Advisory Committee on Reactor Safeguards l10
Application Organization Part 1 - Administrative Information Part 2 - Site Safety Analysis Report Chapter 1 - Introduction and General Description Chapter 2 - Site Characteristics Chapter 3 - Aircraft Hazards Chapter 11 - Radioactive Waste Management Chapter 13 - Emergency Planning Chapter 15 - Transient and Accident Analysis Chapter 17 - Quality Assurance Part 3 - Environmental Report Part 4 - Limited Work Authorization - Not Used Part 5 - Emergency Plan Part 6 - Exemptions and Departures Part 7 - Withheld Information Advisory Committee on Reactor Safeguards l 11 Part 8 - Enclosures
ESPA Development Regulatory bases for the SSAR: Regulatory bases for the ER:
NRC Regulations10 CFR 20, 10 CFR 50, National Environmental Policy Act, 10 CFR 52, and 10 CFR 100 NRC Regulations10 CFR 51 and 10 CFR 52, NUREG-0800, Standard Review Plan for NRC Regulatory Guide 4.2, Preparation of the Review of Safety Analysis Reports for Environmental Reports for Nuclear Power Nuclear Power Plants: LWR Edition Stations, NRC Regulatory Guide 1.206, Combined NRC Regulatory Guide 4.7, General Site License Applications for Nuclear Power Suitability Criteria for Nuclear Power Stations, Plants (LWR Edition)
NUREG-1555, Federal, regional, state and RS-002, Processing Applications for Early local environmental statutes, as applicable, and Site Permits RS-002, Processing Applications for Early Site Permits. Advisory Committee on Reactor Safeguards l 12
ESPA NRC Interactions Pre-Environmental Report Visit March 2013 PPE Development September 2014 Pre-application Site Visit October 2014 Alternative Sites Visit June 2015 ESPA Readiness Review August 2015 Hydrology and Health Physics Audit April 2017 Seismic/Geotechnical Audit May 2017 Environmental and Meteorology Audit May 2017 QA Inspection April 2018 Meteorology and Health Physics Audit May 2018 Advisory Committee on Reactor Safeguards l 13
ASER/ACRS Committee Timeline 1st Set ASERs Issued April 2018 ACRS Subcommittee Meeting May 2018
- SSAR Sections 2.1, 2.2, 3.5.1.6, 15.0.3 ASER - SSAR 13.3 Issued July 2018 ACRS Subcommittee Meeting August 2018
- SSAR Section 13.3 ASER - SSAR 2.5 Issued September 2018 ACRS Subcommittee Meeting October 2018
- SSAR Section 2.5 2nd Set ASERs Issued October 2018 ACRS Subcommittee Meeting November 2018
- SSAR Sections 2.3, 2.4, 11.2/11.3, 17.0 ACRS Full Committee Meeting December 2018 Advisory Committee on Reactor Safeguards l 14
Emergency Preparedness Archie Manoharan Licensing Engineer Advisory Committee on Reactor Safeguards l 15
ESPA - Emergency Preparedness Approach Emergency Planning (EP) Information Layout - 3 Areas Part 2, SSAR, Section 13.3, Emergency Preparedness Plume exposure pathway (PEP) emergency planning size (EPZ) sizing methodology Part 5, Emergency Plan Two major features (Onsite) Emergency Plans Part 5A - Site Boundary EPZ Emergency Plan Part 5B Mile EPZ Emergency Plan Part 6, Exemptions and Departures 2 sets of exemption requests Exemption requests for a PEP EPZ at Site Boundary Exemption requests for a 2-mile PEP EPZ The final EPZ size for the Clinch River Site will be determined at COLA stage Advisory Committee on Reactor Safeguardsl 16
PEP EPZ Sizing Methodology Takes SMR design and safety advancements into consideration Dose-based, consequence-oriented approach to determine an appropriate EPZ size Consistent with the NUREG-0396 sizing rationale - spectrum of accidents are addressed Approach has the same dose criteria as NUREG-0396 - 1 rem total effective dose equivalent (TEDE)
Technical Criteria - PEP EPZ should:
Criterion A - encompass those areas in which projected dose from design basis accidents (DBAs) could exceed the U.S. Environmental Protection Agency (EPA) early phase protective action guide (PAG)
Criterion B - encompass those areas in which consequences of less severe core melt accidents could exceed the EPA early phase PAG Criterion C - be of sufficient size to provide for substantial reduction in early health effects in the event of more severe core melt accidents Advisory Committee on Reactor Safeguards l 17
PEP EPZ Sizing Methodology Step 1 - Accident scenario selection DBA from Chapter 15 Design and site specific Probabilistic Risk Assessment (PRA) for severe accident scenarios Considers - all modes, internal & external events, applicable fuel handling, spent fuel pool, and multi-module accidents Sequences with mean core damage frequency (CDF) greater than 1E-8 per reactor-year (rx-yr)
Criterion B: Less severe core melt scenarios - Mean CDF greater than 1E-6 per rx-yr, intact containment Criterion C: More severe core melt scenarios - Mean CDF greater than 1E-7 per rx-yr, containment bypass or failure Step 2 - Determine source term releases from selected accidents Step 3 - Calculate dose consequences at distance Step 4 - Compare the dose at distance to EPA early phase PAG COL applicant would perform an analysis using the PEP EPZ size methodology, with site- and design-specific input, to justify the PEP EPZ size for the COLA Advisory Committee on Reactor Safeguards l 18
PEP EPZ Sizing Methodology - Example Analysis Criteria A & B: DBA and less severe accidents Dose consequences do not exceed the early phase EPA PAG - 1 rem total effective dose equivalent (TEDE)
Criterion C: More severe accidents Calculate distance at which conditional probability to exceed 200 rem whole body exceeds 1E-3 per rx-yr Verify the PEP EPZ is of sufficient size to provide for substantial reduction in early health effects Design-Specific Example Analysis - Evaluates NuScale Power Plant at Clinch River Site Criteria Site Boundary Dose TEDE EPA Early Phase PAG (rem) Limit TEDE (rem)
A: Design Basis Accidents 0.104 1 B: Less Severe Core Melt Accidents 0.158 1 C: Reduction in Early No accident scenarios met the required screening criteria.
Severe Health Effects Advisory Committee on Reactor Safeguards l 19
Part 5 - Emergency Plan Part 5 of the ESPA contains the major features of two distinct Emergency Plans for Clinch River Site in accordance with 10 CFR 52.17(b)(2)(i).
Part 5A Describes major features of an Emergency Plan for a PEP EPZ consisting of the area encompassed by the Site Boundary.
Part 5B Describes major features of an Emergency Plan for a PEP EPZ consisting of an area approximately two miles in radius surrounding the Clinch River Site.
Both plans address the 16 planning standards in NUREG-0654,Section II, which reflects the requirements in 10 CFR 50.47(b)(1) through 10 CFR 50.47(b)(16) and Appendix E to 10 CFR Part 50 considering the requested exemptions described in Part 6 of the ESPA Advisory Committee on Reactor Safeguards l 20
Part 6 - Exemptions and Departures Pursuant to 10 CFR 52.7, Specific Exemptions, which is governed by 10 CFR 50.12, Specific Exemptions, TVA requested exemptions from the following emergency preparedness requirements for the Clinch River Site:
Certain standards in 10 CFR 50.47(b) regarding onsite and offsite emergency response plans for nuclear power reactor Certain requirements of 10 CFR 50.33(g) and 10 CFR 50.47(c)(2) to establish PEP EPZ for nuclear power plants Certain requirements of 10 CFR Part 50, Appendix E, which establish the elements that make up the content of emergency plans Two Sets of Exemptions Exemptions for a PEP EPZ established at the Site Boundary Deviate from 10-mile PEP EPZ Various elements of a formal offsite emergency plan Evacuation time estimates Certain elements of offsite notifications and exercises Exemptions for an approximate 2-mile PEP EPZ Deviate from 10-mile PEP EPZ Advisory Committee on Reactor Safeguards l 21
Emergency Preparedness Approach - Summary ESPA COLA Approval of the dose-based, Approval of design specific PEP EPZ Methodology consequence oriented methodology implementation of the methodology (Part 2, SSAR, Section 13.3) for determining the PEP EPZ size approved in the ESPA Approval to deviate from the current Approval of design specific PEP EPZ EPZ Size 10-mile PEP EPZ requirements size based on design specific (Part 6) based on the methodology to implementation of the methodology determine PEP EPZ size Approval of the remaining elements Approval of the major features of the of either the Site Boundary or 2-mile Emergency Plan Site Boundary and 2-mile emergency emergency plans OR a new plan (Part 5) plans presented in Part 5 based on design specific PEP EPZ size using methodology Advisory Committee on Reactor Safeguards l 22