ML091390631

ACRS Meeting with the U.S. Nuclear Regulatory Commission, - June 4, 2009, Slides

ML091390631
Person / Time
Site:	Beaver Valley, Browns Ferry, Indian Point, Three Mile Island, Susquehanna, Framatome, Vogtle, 05200017, 05200006, 05200021, 05200011
Issue date:	06/04/2009
From:	Bonaca M V Advisory Committee on Reactor Safeguards
To:
Duraiswamy S, 415-7364
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ML091390564	List: ML091390631
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ACRS MEETING WITHTHEUSNUCLEAR THE U.S. NUCLEAR REGULATORY COMMISSION June 4, 2009 OVERVIEWMARIO V. BONACA Accomplishments*Since our last meeting with the Commission on November 7, 2008, we issued 16 Reports*Topics included:Ctitidt 3-C on t a i nmen t acc id en t pressure credit issue-Selected Chapters of the ESBWR design certification

application

-Vogtle early site permitapplication and limited work authorization -Technical basis for revising 10CFR50.46(b)loss

-of-coolant 4 10 CFR 50.46(b) loss of coolant embrittlement criteria for fuel

cladding materials -Pressurized thermal shock rule

-Regulatory Guide on managing the safety/security interface-Regulatory Guide on cyber security programs for nuclear

facilities 5-Options to revise NRC regulations based on ICRP recommendations License Renewal Since November 2008:*Completed review of the Vogtle license renewal application *Performed interim review of four applications (Beaver Valley, IndianPoint,ThreeMileIsland 6 Indian Point, Three Mile Island Unit 1, and Susquehanna) *Performed interim review of the NIST research reactor

• Discussed with the staff the status of license renewal activities, interim staff guidance, and implementation of the recommendations from the self

assessment 7

• Will perform final review of six applications, including NIST

research reactor, during CY2009*Will review updates to the GALL Report and license renewal guidancedocuments 8 guidance documents Extended Power Uprates*We have expressed concerns with credit for containment accident pressure associated with EPUs in our February 16, 2007, and March 18, 2009, reports 9 reports*We will review the Browns Ferry Unit 1 EPU after receiving the

complete safety evaluation

report

• Browns Ferry Units 2 and 3 EPU application review has been

deferred by the staff at the request of TVA. ACRS will review this application after receivingthecompletesafety 10 receiving the complete safety evaluation report.

New Plant Activities*Completed review of the SER Chapters for the ESBWR design

certification application-Provided six interim letters on20Chapters 11 on 20 Chapters-Will review the resolution of open items and the ACRS issues and the final SER

• Completed review of the early site permit application and limited work authorization for the Vogtle plant*Reviewing topical reports associatedwiththeUS

-APWR 12 associated with the US APWR design*Reviewing revisions to the AP1000 Design Control

Document

• Review of the SER on the EPR design certification application will start in July 2009*Review of the SER on North Anna COL application, referencing 13ESBWR design will begin in June

2009

• Will continue to interact with the NRO staff to establish schedule for review of design certification and COL applications to ensure timely completion of ACRS review 14 Ongoing/Future Activities*Advanced reactor research plan *Combined license a pp lications 15 pp*Design certification applications*Digital instrumentation and control systems

• Extended power uprates*Fire protection

• High-burnup fuel and cladding issues*Human reliability analysis 16*License renewal applications

• New fuel designs and materials

• Next generation nuclear plant (NGNP) project*Pellet clad interaction failure under EPU conditions

• Research quality assessment*Revisions to regulatory guides and SRPs*Risk-Informing the regulations

• Safeguards and security matters

• Safetyculture 17*Safety culture*Safety research program report

• Seismic issues

• State-of-the-Art Reactor Consequence Analyses (SOARCA) Project*Sump strainer issues

• TRACE code applicability to newreactors 18 new reactors*Waste management, radiation protection, decommissioning, and materials issues*Watts Bar Unit 2 operating license Crediting Containment Accident Pressure in the 19NPSH CalculationsWilliam J. Shack NPSH MarginSatisfactory performance of the ECCS and containment heat removal system pumps requires adequate NPSH marginRG 1.1: Emer g enc y core coolin g 20gygand containment heat removal systems should be designed so that adequate NPSH is provided to system pumps assuming no increase in containment pressure from an accident To maintain defense in depth, it is desirable that ECCS function not depend on containment integrity, so that an unexpected loss of containmentintegritynotleadDefense in Depth/Additional Safety Margin 21 containment integrity not lead automatically to core meltIntent of RG 1.1 is to ensure that this independence of the ECCS function be maintained consistently for all reactors Sump strainer blockage is a complex issue. Difficult to provide a demonstrably "conservative" answer. Desirable to maintain margin to address uncertainties 22 Extended Power Uprates

• For some plants, demonstrating adequate NPSH for EPU operation would require:-Credit for all of the predicted 23containment accident pressure-Reliance on operator action to maintain NPSH

-Reliance on COP credit for long duration*In some cases, pump cavitation is expected even after crediting all of the predicted accident 24 ppressure ACRS Position on COP Credit*NRC should seek to maintain independence of containment function and accident mitigation and additional margin for NPSH 25 ACRS MARCH 18, 2009 LETTER

• Intended primarily to address voluntary requests for a change in the licensing basis

• SRP should be revised to state that,ifCOPcreditisgranted 26that, if COP credit is granted based on risk information, all subsequent licensing applications involving COP credit should also include risk information

• Demonstrate that it is not practical to reduce or eliminate the need for COP credit by hardware changes or requalification of equipment 27*If credit for COP is granted, it should be limited in amount and duration

• If operator actions are required to maintain overpressure, it must be demonstrated they can be performed reliably, and that any increase in risk is acceptably

small 28

• Continue to use guidance in RG-1.82 Rev. 3 and the licensing-basis analyses assumptions and methods to show that the Recommendation on Analyses and Revision of RG-1.82 29available NPSH exceeds that needed for the ECCS and containment heat removal system pumps

• If COP credit based on the licensing-basis analyses is not small and limited in duration, RG-1.82 should be revised to request additional anal y ses and 30 yinformation that demonstrate the COP credit needed is small and limited in duration on a more realistic basis

• Such information could include thermal-hydraulic analyses that reduce conservatism but account for uncertainties and PRA results that show that large COP credit is needed only for verylowprobabilityevents 31very low-probability events*If operator actions are required, it should be shown they can be implemented in procedures and performed reliably and that any resulting increases in risk are

small ACRS Position on Decisionmaking*Granting COP credit should depend on integrated decisionmaking that considers less conservative estimates of 32the COP credit; the likelihood of scenarios that require COP credit; and the operator actions required to maintain NPSH Conclusion*Our March 18, 2009 letter is consistent with long-standing ACRS position*Expect to provide technical input to the development of Revision 4 33to RG-1.82

• Had a briefing on a draft of the staff's White Paper. While comprehensive, it did not resolve the ACRS concerns*In the review of an y particular 34 ypapplication for credit, the fidelity of containment and core calculations need to be taken into

account

• BWROG submitted and staff reviewed a more realistic methodology for evaluating COP credit*ACRS awaits the staff's safety evaluationoftheBWROG 35evaluation of the BWROG methodology Pressurized Thermal Shock RuleJ. Sam Armijo Rule Requirements*This rule requires plant-specific evaluations of vessel embrittlement and flaw distributions. It also requires evaluationofnewsurveillance 37evaluation of new surveillance data to ensure detection of unexpected embrittlement trends Three Plant Study*The screening limits are based upon a detailed study of the PTS challenges at three plants*Medium and large LOCAs were themajorcontributorstothe 38 the major contributors to the through-wall cracking frequency (TWCF), which is the risk metric Generalization*A generalization study evaluated the variability of PTS challenges from internal events in plants not included in

the detailed study 39*The likelihood and severity of the important PTS challenges were determined to be representative of those for the entire fleet of PWRs

• A bounding analysis on the effects of external events showed that their contribution to TWCF was less than that of internal events

• Togetherwiththe 40Together with the generalization study on internal events, this finding provides assurance that plant-specific analyses of PTS challenges are not needed

• The Committee concurs with the staff's conclusion that plant-specific evaluations of PTS challenges are not needed andthatthescreeningcriteria 41 and that the screening criteria in 10 CFR 50.61a may be applied to the entire fleet of

PWRs Recommendations*To aid in the implementation of the rule, the staff should undertake an effort to verify and document the capability of NDEproceduresthatwillbe 42 NDE procedures that will be used to characterize the flaw distributions in reactor vessels

• An effort is needed to plan for the most effective use of surveillance samples to ensure that any deviations from the cu rr e n t u n de r sta n d in g o f 43cuetudestadgoembrittlement trends in reactor vessels will be identified in a timely manner DigitalI&CMatters 44 Digital I&C MattersGeorge E. Apostolakis

• Reviewed Regulatory Guide 5.71, "Cyber Security Programs

For Nuclear Facilities" *Reviewed Digital I&C Interim Staff Guidance 5, "Highly-IntegratedControlRoom

-45 Integrated Control RoomHuman Factors Issues," and 6, "Licensing Process" ACRS March 19, 2009 Report*RG-5.71 on cyber security should not be published until it is revised

to:ProvideareferenceDI&C 46-Provide a reference DI&C computer, communication, and

network security framework that

identifies assets, associated plant functions, vulnerabilities, interaction, and access

pathways

-Include examples and more specific guidance on how the requirements of 10 CFR 73.54

can be met-Ensure that the guidance distinguishesbetweenDI&C 47 distinguishes between DI&C system and non-real-time

information technology system

architectures

-Address the issues of threat assessment, dependency analysis, and the use of Probabilistic risk assessment 48 ACRS April 21, 2009 Report on Digital I&C Interim Staff Guidance 5 and 6*Section 3, "Crediting Manual Operator Actions in Diversity and 49Defense-in-Depth (D3) Analyses,"

of ISG-5 should be revised to

incorporate additional guidance

on the estimation methods of the time required for operator action

• Increased rigor in the supporting analyses should be required as

the difference between the time available and the time required

for operator action decreases 50

• Draft ISG-6 should not be issued until Sections C and D are revised to specify that sufficient

design detail be provided to ensure deterministic behavior

and independence of each DI&C safety train 51 Options to Revise NRC RegulationsBasedonICRP Regulations Based on ICRP RecommendationsMichael T. Ryan Staff Options*No changes to existing framework *Update parts of regulations, not previously revised, to conform to

existin g 10 CFR Part 20 conce p ts 53 gpand quantities based on ICRP Publications 26 and 30 *Begin to further align NRC's regulatory framework with ICRP

Publication 103

• ACRS endorses the staff's preferred option 3, which would

begin to move toward greater alignment between 10 CFR Parts 20and50andAppendixIofPartFebruary 18, 2009 ACRS Report 54 20 and 50 and Appendix I of Part 50 with recommendations in ICRP Publication 103

• ACRS concurs with the staff position that NRC's current regulatory framework continues to provide adequate protection

for the health and safety of 55workers, the public, and the

environment

• The staff should continue its participation in ICRP and other national and international committees and standards

organizations

• TheNRCshouldnotdevelop 56*The NRC should not develop separate radiation protection regulations for plant and animal

species Progress on Recommendations of the Independent External Review Panel on the Materials Licensing ProgramMichael T. Ryan

• The staff has addressed each of the recommendations of the Independent External Review

Panel*The staff has developed Interim StaffGuidanceforreviewingnew 58 58 Staff Guidance for reviewing new license applications*Includes more detailed information gathering and on-site applicant visits

• Staff is developing a process to integrate the National Source Tracking and the Web-Based Licensing Systems as part of the

License Verification System*Efforts are under way to ittll37At 59 59 i n t egra t e a ll 37 A greemen t States into this system *This integration will take time and resources to complete and

implement

• Staff is pursuing ways to add more detail to the physical security requirements as

recommended by the Panel and will be addressed in currently plannedrulemakingsforlarger 60 60 planned rulemakings for larger sealed sources

• Adding Security with equal emphasis as Health, Safety, and Environment for materials licensees will require a change in the culture of the Agency ThAdthAt 61 61*Th e A gency an d th e A greemen t States share this responsibility

• The staff has plans to accomplish the objectives developed from all of the Panel's

recommendations*Some short term goals have alread y been accom p lished 62 62 yp*Additional progress will take time and resources AbbreviationsACRS Advisory Committee on Reactor SafeguardsBWRBoiling water reactor BWROGBoiling Water Reactor Owners Group CFRCode of Federal Regulations COLCombined license COPContainment overpressure CYCalendar year D3Diversity and defense in depth DI&CDigital Instrumentation and Control ECCSEmergency core cooling systemEPREvolutionary Power Reactor EPUExtended power uprate ESBWREconomic Simplified Boiling Water Reactor GALLGeneric Aging Lessons Learned Report ICRPInternational Commission on Radiological Protection ESFEngineered safety features I&CInstrumentation and control 63ISGInterim staff guidance LOCALoss-of-coolant accident NDENon-destructive examination NGNPNext Generation Nuclear PlantNISTNational Institute of Standards and Technology NPSHNet positive suction head NRCNuclear Regulatory Commission NROOffice of New Reactors PRAProbabilistic risk assessment PTSPressurized thermal shock PWRPressurized water reactor RGRegulatory Guide SERSafety evaluation report SRPStandard Review Plan SOARCAState-of-the-Art Reactor Consequence Analyses TVATennessee Valley Authority TWCFThrough-wall cracking frequency US-APWRUnited States -Advanced Pressurized Water Reactor