Use of Low Power and Shutdown Risk in Plant Specific Reactor Regulatory Activities

Listed below are topics on which discussion and feedback are sought at the workshop: 1. Are LPSD core damage frequency (CDF) and large early release frequency (LERF) comparable to full power CDF and LERF? What methods and assumptions should be used to answer this question? 2. Are the LPSD CDF and LERF contributors comparable to the contributors from full power? What are the methods and assumptions should be used to answer this question?

3. How many plant operational states (POS) are needed to adequately represent the risk associated with LPSD operations? 4. Should the scope of LPSD analyses include fuel handling and storage, e.g., full core offloading? What methods and assumptions should be used to answer this question? 5. Is there a sufficient technical basis (knowledge of core melt phenomena, source terms, varying containment configurations, etc.) available to support LERF analysis for LPSD? If not, what issues require additional study? If a sufficient technical basis exists, what information sources can be cited to support the assertion? 6. Is the CDF and LERF associated with the transition from one operational state to another important? What methods and assumptions should be used to answer this question? 7. Is a traditional PRA approach needed to provide an understanding of LPSD for risk-informed regulatory decision-making? If not, what other approaches are available? What are their strengths and limitations? 8. Currently, the staff is supporting efforts to produce a nation consensus standard on full power PRA to support risk-informed decision- making. Is a standard on LPSD needed or desirable? Should it be a national consensus standard? 9. Draft NUREG-1602 provides reference material on the scope and quality of a LPSD PRA. Is the information in this draft complete and correct? Is it useful as reference material in making assessments on an application specific basis on the scope and quality of a LPSD risk assessment to support that particular application? How could it be improved? 10. Would draft NUREG-1602 be useful as a starting point to develop a standard on LPSD PRA? What would be needed? Should it specify acceptable LPSD PRA methods? 11. Given the lack of experience in performing LPSD PRAs, should a standard for LPSD PRA provide both (1) requirements for what activities should be performed and (2) detailed information/instructions on how those activities should be performed? 12. Is LERF an appropriate metric for meeting the Safety Goal Policy Statement for all POS? If not, what metrics should be used? For example, should there be a metric on long term release frequency to supplement LERF? What should it be based upon? 13. Can NUREG/CR-6595 be used to calculate LERF for LPSD conditions? If not, what additional guidance should be added to the report to support LERF calculations for LPSD conditions? 14. Are average equipment unavailabilities during LPSD conditions (resulting in average CDF and LERF estimates) sufficient to support risk-informed decision-making? 15. Is the following definition of an initiating event during LPSD adequate: ``An event that causes loss of the function(s) necessary to maintain the plant in its existing operating state?'' If not, then what changes should be made to enhance the definition? 16. Are there generic data sources for the identification and quantification of LPSD initiating events? If so, are the data sources publicly available? Are these generic data sources consistent? 17. Do certain LPSD operational states have the potential to have more human failures than full power operation? If event trees and fault trees are used to model the response of a plant to LPSD initiating events, where is the more appropriate place to model these human failures? What is the basis for this choice? 18. Are the human reliability analysis methods used in full power analyses sufficient to characterize the unique characteristics and conditions under which humans operate during LPSD? If not, what improvements are required to ensure an adequate representation of human actions during LPSD conditions? If so, how are these methods being used to identify errors of commission? 19. What are the important uncertainties (parameter, model, and completeness) that should be considered in LPSD analyses? How should these uncertainties be evaluated in LPSD analyses? 20. Are there any other issues related to Level 1 and 2 analyses that are important to the development of LPSD risk (CDF and LERF)? Reference material (available for inspection and copying for a fee a the NRC Public Document Room, 2120 L Street N.W. (Lower Level), Washington D.C. 20555-0001; a free single copy of each document, to the extent of supply, may be requested by writing to Distribution Series, Printing and Mail Services, Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington D.C. 20555-0001) includes: • RG 1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis''. • NUREG/CR 6143, ``Evaluation of Potential Severe Accidents During Low Power and Shutdown Operation at Grand Gulf, Unit 1,'' 1995. • NUREG/CR-6144, ``Evaluation of Potential Severe Accidents During Low Power and Shutdown Operation at Surry, Unit 1,'' 1995. • NUREG-1602, ``The Use of PRA in Risk-Informed Applications,'' Draft, June 1997. • NUREG/CR-6595, ``An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events,'' January 1999. In addition (available via the ASME web site, or contact Jess Moon at ASME, email moonj@asme.org): • ASME RA-s-1999, Draft #10, ``Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications,'' Draft for public review and comment.

WORKSHOP MEETING INFORMATION: The Commission intends to conduct a workshop to solicit information related to the risk associated with low power and shutdown conditions sufficient to support decision-making for risk-informed regulation. Persons other than NRC staff and NRC contractors interested in making a presentation at the workshop should notify Erasmia Lois, Office of Nuclear Regulatory Research, MS: T10- E50, U.S. Nuclear Regulatory Commission, Washington D.C., 20555-0001, (301) 415-6560, email: exl1@nrc.gov