[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 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
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: exl1@nrc.gov
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: mxd@nrc.gov
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
