ML18085A471

From kanterella
Jump to navigation Jump to search
M180405B: Scheduling Note and Slides - Meeting with the Advisory Committee on Reactor Safeguards (Public Meeting)
ML18085A471
Person / Time
Issue date: 04/05/2018
From:
Advisory Committee on Reactor Safeguards
To:
Shared Package
ML18085A083 List:
References
M180405
Download: ML18085A471 (58)
v • d • e


Text

SCHEDULING NOTE

Title:

MEETING WITH ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (Public)

Purpose:

Semi-annual meeting with the NRC's independent Advisory Committee on Reactor Safeguards (ACRS) for the ACRS to provide their views to the Commission on issues recently reviewed by the Committee.

Scheduled: April 5, 2018 10:00 a.m.

Duration: Approx. 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Location: Commissioners' Conference Room, 1st fl OWFN

Participants:

Presentation ACRS Members 50 mins.*

Mike Corradini, ACRS Chairman

  • Overview
  • Revision 3 to Regulatory Guide 1 .174
  • State-of-the-Art Reactor Consequence Analysis (SOARCA) Project:

Sequoyah Integrated Deterministic and Uncertainty Analyses Dana Powers, ACRS Member

  • Report on the Safety Aspects of the Construction Permit Application for Northwest Medical Isotopes, LLC, Radioisotope Production Facility Joy Rempe, ACRS Member
  • Biennial Review and Evaluation of the NRG Safety Research Program Commission Q & A 30 mins.

Discussion - Wrap-up 5mins.

  • For presentation only and does not include time for Commission Q & A's 1

\

(~~~u.S.NRC United States Nuclear Regulatory Commission Protecting People and the Environment ACRS MEETING WITH THE U.S. NUCLEAR REGULATORY COMMISSION April 5, 2018

c~~* U.S.NR C United States Nuclear Regulatory Commission Protecting People and the Environment Overview Mike Corradini

Accomplishments Since our last meetia,g with the Commission on October 6, 2017, we issued 11 Reports

Combined Reactivity Control Systems Capability

Reports

  • State-of-the-Art Reactor Consequence Analysis (SOARCA) Project: Sequoyah Integrated Deterministic and Uncertainty Analyses
  • Report on the Safety Aspects of the Construction Permit Application for Northwest Medical Isotopes, LLC, Radioisotope Production Facility
  • Biennial Review and Evaluation of the NRC Safety Research Program 4

Reports

  • Safety Evaluation for Topical Report ANP-10300P, Revision O, AURORA-B:

An Evaluation Model for Boiling Water Reactors; Application to Transient and Accident Scenarios

  • Safety Evaluation of the NuScale Power, LLC Topical Report TR-0116-20825-P, Applicability of AREVA Fuel Methodology for the NuScale Design

5

Reports

  • Safety Evaluation for Topical Report APR1400-F-M-TR-13001, Revision 1,

PLUS7 Fuel Design for the APR1400

  • Safety Evaluation for ANP-10333P, Revision O, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Ro_d Drop Accident (CRDA)

6

Reports

  • Regulatory Guide 1.232:

Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors

  • Assessment of the Quality of Selected NRC Research Projects 7

Ongoing / Future Reviews

  • Design Certification

-APR 1400

- NuScale

  • Early Site Permit

- Clinch River

  • Brunswick Units 1 & 2 MELLLA+

8

Ongoing / Future Reviews

- Seabrook

- Waterford Unit 3

- River Bend

  • AP1000

- WCAP assessing potential debris generation from AP1000 cables and non-metallic insulation (GSl-191) 9

Ongoing / Future Reviews.

  • Guidance and Bases

- Draft Regulatory Guide DG-1327, Reactivity-Initiated Accidents

- NUREG on High Burnup Fuel Storage and Transportation

- NUREG/BR-0058

  • Advanced Reactors

- Licensing Modernization Framework

- Functional Containment Policy Paper 10

Ongoing / Future Reviews

- ISG-06 Revision

- Diversity and Defense-in-Depth against Common Cause Failure

- Integrated. Action Plan

  • Rulemaking

- Emergency Preparedness for SMRs

- Non-Power Production or Utilization Facility 11

Ongoing/ Future Reviews

.* Thermal-Hydraulic Phenomenology

- GSl-191

  • PWR Owners Group In-vessel Debris Test Results

- AREVA's AURORA-B Transient Code Suite: LOCA

  • Metallurgy and Reactor. Fuels

- Consolidation of Dry Cask and Dry Fuel Storage Standard Review Plans 12

Ongoing / Future Reviews

  • Reliability and PRA

- Level 3.PRA

- Human Reliability Analysis Method Development

  • Control Room Abandonment Risk 13

(-~~ U.S.NR C United States Nuclear Regulatory Commission

Background

  • The General Design Criteria are the minimum requirements for principle design criteria for water-cooled nuclear plants to provide reasonable assurance that the facilities can be operated safely
  • GDC's were based on the licensing of early commercial water-cooled reactor plant designs
  • Staff has acknowledged that fulfillment of some

.of the GDC may not be necessary or appropriate for some designs

  • NuScale reactor is a modular, passive, water-cooled reactor design with innovative design features 15

Background

GDC 27, Combined Reactivity Control Systems Capability

The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin

  • for stuck rods the capability to cool the core is maintained 16

Background

  • Staff has historically interpreted the intent of GDC 27 to require that the reactor:

- Be reliably controlled in normal operation

- Achieve and maintain a safe shutdown condition, including subcriticality beyond the short-term, using only safety-relat*ed equipment following a DBE with margin for stuck rods

  • Staff informed NuScale that an exemption vvould be required for its reactor design 17

Requested Exemption

  • NuScale submitted a request for an exemption to GDC 27
  • Staff plans to evaluate vvhether the NuScale design meets the underlying intent of the GDC and assures public health and safety are maintained based on tvvo criteria:

- Demonstrate sufficient core cooling

- DBE sequence of events is not expected to occur during the lifetime of a module 18

Maintain Long-Term Cooling

  • To ass~re long-term core cooling, \Ne expect that NuScale \Nill perform an evaluation to ensure SAFDLs are not exceeded for any of the DBE scenarios considered. Analyses \NOuld include:

- Consideration of operator actions

- Estimates of the return to power and associated strategies to return to a subcritical condition

~ Assurance that the margin does not degrade over the duration of the event 19

Low Probability of Return to Power

  • The staff evaluation criteria should be augmented to include:

- An assessment of the incremental risk to public health and safety from the hypothesized situation

- Whether that risk increase_ is acceptable, consider.ing the entire NuScale facility 20

Low Probability of Return to Power

  • Non-safety SSCs that provide boron addition should have certain characteristics

- They should not degrade during plant operations

- They should function reliably when called upon, including operator actions needed for their startup and alignment 21

ACRS Conclusion and Recommendation The proposed criteria are reasonable provided the follo\Ning recommendations and enhancements outlined in the letter report are addressed:

1. Evaluate the overall risk and not just the frequency of the challenge
2. Risk considerations should be based on the facility rather than an individual module 22

c~~u.S.NRC United States Nuclear Regulatory Commission Protecting People and the Environment Revision 3 to Regulatory Guide 1.174 John Stetkar

Background

  • Describes key. princi_ples and guidance for the use of risk information in regulatory decisions
  • Primary intent of Revision 3 to clarify guidance for considering defense-in-depth 24

ACRS Engagement

  • Four Subcommittee meetings from May 2016 to August 2017
  • ACRS previously reviewed evolution and interpretation of the defense-in- .

depth philosophy (NUREG/Kn,1-0009) during the* staff's evaluation of issues for implementation of a proposed Risk Management Regulatory Framework 25

ACRS Recommendation.

- Substantially expands and clarifies the guidance for consideration of ~efense-in-depth and its integration with the other risk-informed decision-making principles

- Clarifies the staff's intent for determining acceptability of a PRA for use in risk-informed decisions

- Enhances the guidance on evaluation and treatment of uncertainties

  • Issued in January 2018.

26

Future Revisions

  • Plans to expand the guidance on integrated decision-making and the use of uncertainty as an input to the decision process
  • Encourage staff to also consider extending the guidance to address applications of risk information for new reactors, which may have much different risk profiles and lower overall levels of risk than currently operating reactors 27

(--~-u.S.NRC United States Nuclear Regulatory Con1mission

.Protecting People and the Environment STATE-OF-THE-ART REACTOR CONSEQUENCE ANALYSES (SOARCA) PROJECT SEQUOYAH INTEGRATED DETERMINISTIC AND UNCERTAINTY ANALYSES John Stetkar 28

Background

  • Original Peach Bottom and Surry SOARCA studies reported only

point estimate results, without an evaluation of the uncertainties in those estimates

  • Subsequent to the original studies, focused uncertainty analyses were performed for selected scenarios at Peach Bottom and Surry 29

Background

  • In most cases, the uncertainties were retrofit around the point estimate values used in the original studies
  • These* focused studies provided important insights about how consideration of the uncertainties affects understanding and interpretation of the results 30

Background

  • Sequoyah study extends the scope of the SOARCA analyses to include a focused evaluation of severe accident response for a PWR with an ice condenser containment
  • It is intended specifically to examine the effects from hydrogen generation
  • and release, timing and locations of ignition, and containment vulnerability to failure caused by a highly energetic deflagration 31

Background

  • Sequoyah study evaluates responses to one short-term and one* long-term station blackout scenario
  • Assumes each scenario is caused by a severe earthquake
  • Offsite emergency response models account for infrastructure damage
  • Integrated evaluation of uncertainties for thermal-hydraulic response and offsite consequences for only the short-term blackout scenario .

32

ACRS Engagement

  • Three Subcommittee meetings on Sequoyah SOARCA study between May 2016 and October 2017
  • April 2017 joint Subcommittee meeting on changes to thermal-hydraulic models and analyses in MELCOR 33

. ACRS Conclusions and Recommendations 1

  • Sequoyah SOARCA study has significantly advanced the understanding of severe accident progression in a PWR with an ice condenser containment
  • It demonstrates the importance of an integrated assessment of uncertainties about equipment performance, thermal-hydraulic

.phenomena, and emergency planning 34

ACRS Conclusions and Recommendations 2

  • Study evaluates site-specific conditional consequences from two station blackout scenarios, tailored to examine the effects from hydrogen generation, ignition, and containment failure vulnerability
  • It does not examine other scenarios that may be important for containment failure or bypass 35

ACRS Conclusions and Recommendations 2

  • Study does not account for accident mitigation strategies that have been implemented at Sequoyah
  • Results from the study should not be extrapolated to other PWRs with ice condenser containments at other sites 36

ACRS Conclusions and Recommendations 3

  • Sequoyah SOARCA report should be published after the staff more clearly documents the following issues and research needed for their resolution:

- Potentially important modeling uncertainties

- Justification for safety valve failure rates

-. Failures to complete some MELCOR simulations involving an early stuck-open pressurizer safety valve 37

ACRS Conclusions and Recommendations 4

  • Staff should examine and resolve the issues regarding safety valve failure rates and MELCOR performance before further enhancements are made to the SOARCA studies 38
  • Staff Responses
  • Report emphasizes that study is specific to Sequoyah
  • Evaluation of model uncertainty is outside the study scope
  • Report contains enhanced discussions .

of safety valve failure rates and insights from incomplete MELCOR runs

  • Will address safety valve failure rates in updated Surry uncertainty analyses 39

("~U.S.NRC United States Nuclear Regulatory Commission Protecting People and the Environment Report on the Safety Aspects of the Construction Permit Application for Northwest Medical Isotopes, LLC, Radioisotope .

Production Facility Dana Powers

Background

  • Northwest Medical Isotopes, LLC (NWMI) submitted a preliminary design for a facility that addresses hazards associated with the extraction of 99 Mo from irradiated targets and the fabrication of targets for irradiation 41

Background

  • ACRS reviewed the preliminary safety analysis report submitted by NWMI and the draft final safety evaluation report prepared by the NRC staff 42

Conclusions and .

Recommendations

  • Once the design* is finalized, the proposed facility can be
  • constructed and licensed for operation with adequate protection of the- public health and safety and no undue risk to the environment 43

Conclusions and Recommendations

  • A construction permit for the proposed radioisotope production facility can be *issued to NWMI 44

~U.S.NRC United States Nuclear Regulatory Commission Protecting People and the Environment Biennial Review and Evaluation of the NRC Safety Research Program Joy Rempe

Background

  • 2018 biennial review addresses 1997 Com-mission guidance:*

- Need, scope, and balance of reactor safety research program

- Progress of ongoing activities

- How well RES anticipates research needs and how it is positioned for changing environment 46

Background

  • 2018 biennial review also emphasizes:

- Prioritization and identification of new research needs

- Long-term planning

- More succinct ACRS report 47

I .

Background

  • 2018 biennial report developed using insights from:

- Initial meeting with RES Director to obtain overview of program, plans, priorities, and areas of interest

- Three working group meetings to discuss research conducted by each RES division:

Division of Risk Analysis {DRA), Division of System Analysis {DSA), and Division of Engineering {DE)

- Other ACRS activities 48

DRA Review Findings

  • Division-specific:

- Level 3 PRA

- IDHEAS 49

DRA Review Findings

  • General:

- It is not clear ho\N *research priorities account for integrated consideration of 'enterprise risk', \Nhich addresses factors such as safety and security, en,erging issues, innovative technologies and associated

~ncertainties, preservation of core con,petencies, and development and rn_aintenance of analysis methods and

  • tools 50

DSA Review Findings

  • Division-specific:

- The agency must have an independent reactor safety analysis capability

- Consequence analysis should be a core competency 51

DSA Review Findings

  • General:

- Difficult strategic choices needed to maintain current computational capabilities and core competencies

  • and to anticipate and adapt to future regulatory needs 52

DE Review Findings

  • Division-specific:

- Material performance computer code development

- Codes and standards review

- Risk evaluation needed prior to embarking on spent fuel dry storage cask research 53

DE Review Findings

  • General:

- Rather than developing data independently, efforts should focus on identifying data that licensees or applicants must provide

- An effective process should be -

developed for terminating ongoing research that ceases to be high priority 54

Conclusion and Recommendations

  • NRC safety research program appears to be meeting near-term agency needs satisfactorily
  • Current process to prioritize agency

-research could be improved by performing a systematic assessment that emphasizes

'enterprise risk' in research project selection, evaluation, and./

termination 55

Conclusion and Recommendations

  • RES should develop long-term strategies to address emerging technical issues, support development and maintenance of needed analytical tools and data bases, emphasize activities that improve regulatory efficiency, and identify and preserve needed core competencies 56

Abbreviations ACRS Advisory Committee on Reactor NWMI Northwest Medical Isotopes, LLC Safeguards PRA Probabilistic Risk Assessment CFR Code of Federal Regulations PWR Pressurized-Water Reactor CRDA Control Rod Drop Accident RES Office of Nu~lear Regulatory Research DBE Design Basis Event SAFDL Specified Acceptable Final Design DG Draft Regulatory Guide Limit GDC General.Design Criteria SECY Secretary of the Commission GSI Generic Safety Issue SER Safety Evaluation Report l&C Instrumentation and Control SMR Small Modular Reactor IDHEAS Integrated Human Event Analysis SOARCA State-of-the-Art Reactor Consequence System Analyses ISG Interim Staff Guidance SRM Staff Requirement Memorandum LOCA Loss-of-Coolant Accident SSC Structure, System, and Component MELLLA+ Maximum Extended Load Line Limit Analysis Plus 99Mo Molybdenum 99 57 L

Retrieved from "https://kanterella.com/w/index.php?title=ML18085A471&oldid=2411855"