ML20011F648

Forwards Prioritization Evaluation Re Generic Issue 117, Allowable Outage Times for Diverse Simultaneous Equipment Outages. Issue Addressed in Tech Spec Improvement Program & Will Be Dropped as Separate Issue

ML20011F648
Person / Time
Issue date:	02/26/1990
From:	Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Murley T Office of Nuclear Reactor Regulation
References
REF-GTECI-117, REF-GTECI-NI, TASK-117, TASK-OR NUDOCS 9003070079
Download: ML20011F648 (7)

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FEB 2 6 1990 MEMORANDUM FOR: Thomas E. Murley, Director, Office of Nuclear Reactor Regulation FROM:

Eric S. Beckjord', Director, Office of Nuclear Regulatory Research

SUBJECT:

GENERIC ISSUE 117, " ALLOWABLE OUTAGE TIMES FOR DIVERSE SIMULTANEOUS EQUIPMENT OUTAGES" The prioritization of Issue No. 117. " Allowable Outage Times for Diverse Simultaneous Equipment Outages," shows that the safety concern is being addressed in a broader context in the ongoing Technicel Specification Improvement Program in NRR.

It is our understanding that any significant findings resulting from this program will be factored into the Technical Specificatians for future plants. Thus, Issue 117 will be DROPPED from further consideration as a separate issue.

The enclosed prioritizacion evaluation will be incorporated into NUREG-0933, "A Prioritization'of Generic Safety Issues " and is being sent to the regions, L

'other offices, the ACR3, and the PDR, by copy of this memorandum and its L

enclosure, to allow o'.hers the opportunity to comment on the evaluation.

All l

comments should be sent to the Advanced Reactors and Generic Issues-Branch, l-DRA,RES(MailStopNL/S-169).

Should you have any questions pertaining to the l-contents of this memorandum, please contact Ronald Emrit (492-3731).

1 pD l

Eric S. Beckjord, Director ffice of Nuclear Regulatory Research l-L

Enclosure:

Prioritization Evaluation l

cc:

E. Jordan, AEOD l'

.W. Russell, Reg. I S. Ebneter, Reg. II A. Davis, Reg. III-E R. Martin, Reg. IV E

J -Martin, Reg. V GPDREA ACRS eg 9003070079 9002:26 iy I /

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J ENCLOSURE PRIORITIZATION EVALUATION ISSUE 117:

Allowable Outage Times for Diverse Simultaneous Equipment Outages I

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ISSUE 117:

ALLOWABLE OUTAGE TIMES FOR DIVERSE SIMULTANEOUS EQUIPMENT OUTAGES

-DESCRIPTION

. Historical Background-This issue was identified in 1985 when a concern was raised that, due to non-obvious interactions between diverse equipment, Technical Specifications (T/S) may permit a plant to be in a dangerous configuration by the simultaneous outage of equipment resulting from concurrent test and/or maintenance operations.2221 Safety Significance Of the several thousand safety-related components that make up a nuclear power plant, it is easily conceivable that there may be several components out of servTce for test or maintenance at the same time. Of concern are the potential combinations of diverse components which may appear in the same accident sequence (cut sets).

Such combinations, even though involving redundant divisions of the same system, can give rise to possible high risks which warrant immediate shutdown. The T/S principally address simultaneous outages of redundant components in the same system or redundant systems performing the same or similar functions. Operators are frequently called upon to pass judgment on the acceptability of taking components out of service for maintenance or surveillance as well as tracking compliance with T/S.

It is believed that operators need more help than current documentation or training provide for identifying dangerous configurations that are often a function of valve alignments and switch settings as well as components declared inoperable.

This issue is applicable to all operating and future plants.

Possible Solution A proposed resolution involves the development of mini-computer-based software for use by control room operators in tracking and evaluating prevailing plant

. conditions.

This software would include plant-specific event trees and detailed fault trees. Operators would manually enter the prevailing plant configurations such as valve alignments, switch settings, and equipment out of service at the time. The mini-computer would calculate the core-melt accident minimum cut sets and a core-melt frequency estimate based upon the prevailing configuration.

The core-melt frequency estimate would be an instantaneous value and not a time-averaged value.

A comparable system has been developed by the British for use on their Advanced Gas Reactors.

After the instantaneous core-melt frequency has been calculated, the results are compared with a set _ of critical values to assist the operators in making their decisions regarding subsequent actions. The British have elected not to display the calculated core melt frequency, but use the computer to compare the calculated frequency to two previously established thresholds.

Above a criticd value, a fast shutdown is warranted, provided that the risk in the 3.117-1

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shutdown state is not greater than the risk which results from keeping the

- system in the operating mode.

Below a lower threshold frequency, the plant i'

can remain in the configuration indefinitely.

In between, there is a finite allowable time in which the plant can remain in the configuration.

In addition to providing input regarding plant shutdowns, the software can also be used to establish priorities for repairing components and to aid in decisions on maintenance or surveillance priorities.

PRIORITY DETERMINATION

' Frequency Estimate To determine the core-melt frequency reduction, the cut sets for Oconee 3 and l

Grand Gulf were reviewed to identify where dual elements with maintenance or test outages existed.64 We will define two cut set elements as A and B as 4

follows:

A=Af + A, and B=Bf + B, where A, and B, are those unavailabilities of A and B due to component failure and A hre the unavailabilities of A and B due to test and/or maintenance i

outagI.&nd B" likelihood that A and B would be simultaneously out of service is:

The A x B = (Af + A,)(Bf + B,)

=ABf f + A B,+ A,Bf + A,B, f

Three of the terms include either the unavailability of A due to failure, the unavailability of B due to failure, or the unavailability of both A and B simultaneously due to failure. The fourth or last term is the unavailability of both A and B simultaneously due either to test or maintenance.

Removal of this latter product from the expansion of (A x B) approximates the effect of the generic issue resolution.

'Three cut sets in the Oconee 3 analysis and eight cut sets in the Grand Gulf analysis were found to contain double maintenance or test terms.

Removing the double test or maintenance contributions reduced each cut set by the following amounts:

Reduction (1) Oconee 3 Cut Set (Core-Melt /RY)

S2D 4 x 10 8 S3D 9 x 10 8 T2MQD 1 x 10 7 c

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Reduction (2) Grand Gulf Cut Set-(Core-Melt /RY)

T1PQI 5.3 x 10 7 T23PQI 7.0 x 10 8 T1PQE 4.7 x 10 8 T23PQE 2.0 x 10 8 SI 8.0 x 10 8 T1QW 1.7 x 10 7 T23QW 3.0 x 10 7 T1QUV 1.6 x 10 7 Consequence Estimate The consequences of these accident sequences are obtained by usin the CRAC Code" for PWR and BWR release categories as given in WASH-1400.

The release doses are based on the fission product inventory of a 1120 MWe PWR, meteorology typical of the Byron site, and a surrounding uniform population density of 340 persons-per square mile over a 50-mile radius from the plant site, with an exclusion radius of one-half mile from the plant.

Based on the above, for the frequency reductions of the cut sets described, a dose reauction of 0.55 man-rem /RY for PWRs and 9.1 man-rem /RY for BWRs can be achieved by the resolution of this issue.

For the 90 PWRs with an average life of 28.8 years, this reduction amounts to a decrease of 1,400 man-rem.

For the 44 BWRs with an average life of 27.4 years, this reduction amounts to a decrease of 11,000 man-rem. The total estimated risk reduction associated with this issue is 12,400 man-rem.

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Cost Estimate The resolution of this issue assumes the development and implementation of a mini-computer-based PRA at each plant for performing analyses of the-

- " instantaneous" core-melt frequency dependent upon the plant configuration.

The costs assume that the NRC will demonstrate the capability of this resolution on a selected plant, develop a rule and_ implement the requirement, and review and evaluate submitted licensee responses. The licensee is assumed to develop the computer software for each plant, verify and validate the software package, to train a group of licensed reactor operators to sufficiently understand the fundamentals of PRA and on the use of software, perform a plant-specific PRA (assumed to occur to fulfill the requirements of the IPE resulting in no_ cost for the PRA),1222 purchase the necessary computer equipment to run the program, and to maintain and update the software to reflect changes in plant equipment.

Industry Cost:

It was estimated that 2 man years for software development and 0.8 man years-for verification and validation of the plant software would be. required.

The hardware costs are estimated to be approximately $5,250 for the computer and associated hardware, i.e. printer and color display.

Over the life of the plant, it is further estimated that the computer equipment will be updated (replaced) one time for a total equipment cost of $10,500/ plant.

Training will be provided for two one-week courses for seven operators for a total of 14 man-weeks with an instructor at a cost of $5,000 per course.

At a

. cost of $125,000/ year for labor, the industry costs are estimated to be

$40,000/ plant.

For 134 affected plants, the total industry cost is $53.6M.

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.'t NRC Cost:

It is' estimated that -it would. require 2 man years for validation

'and verification of the resolution and 0.5 man year for the development and implementation of the' rule.

In addition, one man-week will be required to.

evaluate and approve licensee change submittals. The total NRC cost is estimated to be $0.65M.

Value/ Impact Assessment i

Based upon a total public dose reduction of 12,400 man-rem and a total cost of $54.25M, the value/ impact score is given by:

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S 12,400 man-rem

. $54. 25M

= 230 man rem /$M l

Other Considerations.

JThe above score'.is based on the assumption that the PRA performed to satisfy the IPE would be adequate to develop the computer software.

However, if it is assumed that 2 man years / plant would be required to develop the PRA to a depth of detail necessary to develop the software, the value/ impact score would then be 106 man-rem /$M.'

CONCLUSION Based upon the total man-rem reduction and the value/ impact score, this issue would normally be given a medium priority ranking. Given the further consideration that ongoing studies indicate that generic maintenance outage times ~used in PRAs may be low by as much as a factor of 5, the issue could possibly be rated as high.122s However, this issue is being addressed in a much broader investigation in the Technical Specification Improvement Program L

(TSIP) 2224 Any significant findings resulting from the staff's studies will be j'

factored into the T/S for" future plants through the TSIP. Therefore, it is recommended that this issue be DROPPED as a separate issue.

REFERENCES i

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16. WASH-1400 (NUREG-75/014), " Reactor Safety Study, An Assessment of l~

Accident Risks'in U.S. Commercial Nuclear Power Plants," U.S. Nuclear L

Regulatory Commission, October 1975.

64.

NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development," U.S. Nuclear Regulatory Commission, February 1983.(Supplement 1) May 1983, (Supplement 2)

December 1983, (Supplement 3) September 1985, (Supplement 4) July

-i 1986.

l 1221. Memorandum for W. Minners from F. Rowsome, " Candidate Generic l

Safety Issue:

Allowable Outage Times for Diverse, Simultaneous l

Equipment Outages," May 9, 1985.

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- 1222. NRC Letter to' All Licensees Holding Operating Licenses and Construction Permits for Nuclear Power Reactor Facilities, " Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 6 50.54(f),"

(Generic Letter No. 88-20), NovraMvn-23, 1980.

1223. Proceedings of the International Topical Meeting on Probability, Reliability, and Safety Assessment, PSA '89, p. 48, " Potential Underestimation of Test and Maintenance Unavailabilities in Probabilistic Risk Assessments," American Nuclear Society, April 2-7, 1989.

1224. Memorandum for B. Morris from F. Gillespie, "Prioritization of GI-117, ' Allowable Outage Times for Diverse Simultaneous Equipment Outages,'" August 4, 1989.

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