[Federal Register Volume 64, Number 57 (Thursday, March 25, 1999)]
[Notices]
[Pages 14471-14473]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-7275]


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NUCLEAR REGULATORY COMMISSION


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of public workshop.

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SUMMARY: The Nuclear Regulatory Commission has issued guidance for 
power reactor licensees on acceptable methods for using probabilistic 
risk assessment (PRA) information and insights in support of plant-
specific applications to change the current licensing basis. The use of 
such PRA information and guidance is voluntary. This guidance is 
documented in Regulatory Guide (RG) 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis.'' RG 1.174 states that a risk-
informed regulatory process must consider risk associated with all 
operating modes (full power, low power and shutdown). The staff is 
developing (as necessary) acceptable methods to provide an 
understanding of the risk associated with low power and shutdown (LPSD) 
operations sufficient to support decision-making for risk-informed 
regulation.

SUPPLEMENTARY INFORMATION: 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?

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    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 [email protected]):
     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: [email protected]

DATES: April 27, 1999.

AGENDA: Preliminary agenda is as follows (a final agenda will be 
available at the workshop):

Tuesday, April 27, 1999
    7:45 a.m. to 8:00 a.m. Introduction, opening remarks
    8:00 a.m. to 8:45 a.m. NRC Presentations plus open discussion

    --Purpose
    --Status of Activities
    --Plans
    --Understanding of LPSD risk

    8:45 a.m. to 9:15 a.m. Industry Presentations
    9:15 a.m. to 9:30 a.m. BREAK
    9:30 a.m. to 11:30 a.m. Industry Presentations
    11:30 a.m. to 12:45 p.m. LUNCH
    12:45 p.m. to 2:15 p.m. General Discussion of Issues/Topics
    2:15 p.m. to 2:30 p.m. BREAK
    2:30 p.m. to 4:15 p.m. General Discussion of Issues/Topics
    4:15 p.m. to 4:45 p.m. Wrapup

LOCATION: DoubleTree Hotel, 1750 Rockville Pike, Rockville, Maryland.
REGISTRATION: No registration fee for workshop; however, notification 
of attendance is requested so that adequate space, etc. for the 
workshop can be arranged. Notification of attendance should be directed 
to Erasmia Lois,

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Office of Nuclear Regulatory Research, MS: T10-E50, U.S. Nuclear 
Regulatory Commission, Washington D.C., 20555-0001, (301) 415-6560, 
email: [email protected]

FOR FURTHER INFORMATION CONTACT: Mary Drouin, Office of Nuclear 
Regulatory Research, MS: T10-E50, U.S. Nuclear Regulatory Commission, 
Washington D.C., 20555-0001, (301) 415-6675, email: [email protected]

    Dated this 18 day of March, 1999.

    For the Nuclear Regulatory Commission.
Mary Drouin,
Acting Chief, Probabilistic Risk Analysis Branch, Division of Systems 
Technology, Office of Nuclear Regulatory Research.
[FR Doc. 99-7275 Filed 3-24-99; 8:45 am]
BILLING CODE 7590-01-P