ML20043E628
| ML20043E628 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/05/1990 |
| From: | Engle L Office of Nuclear Reactor Regulation |
| To: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| GL-83-28, TAC-54086, TAC-54087, NUDOCS 9006130211 | |
| Download: ML20043E628 (3) | |
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- a meauIo, UNITED STATES
- #T NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555 June 5, 1990 Docket Nos. 50-338 and 50-339 l
d Mr. W. L. Stewart Senior Vice President - Nuclear Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen. Virginia 23060 l
Dear Mr. Stewart:
SUBJECT:
SAFETY EVALUATION FOR GENERIC LETTER 83-28. ITEM 4.5.3, REACTOR TRIP RELIABILITY - ON-LINE FUNCTIONAL TESTING 0F THE REACTOR L
TRIP SYSTEM / NORTH ANNA POWER STATION, UNITS NO. 1 & NO. 2 (NA-1&2)(TACNOS, 54086 & 54087)
Generic Letter 83-28 Item 4.5.3, required confirmation from all licensees and..
applicantsthaton-1Inefunctionaltestingofthereactortripsystem, including independent testing of the diverse trip feature, was being performed.
By letters dated November 4, 1983 and November 18, 1985, you submitted responses to Generic Letter 83-28. Item 4.5.3.
Therein you indicated your endorsement of the Westinghouse Owners Group Report WCAP-10271, as being applicable to NA-1&2. The NRC staff, with the assistance of our contractor, Idaho National Engineering Laboratory, has reviewed your responses and the WCAP-10271 Owners Group report. We conclude that the existing intervals for i
on-line functional testing at NA-182 are consistent with achieving high reactor trip system availability.
1 Theenclosedcontractor'sreport(EGG-NTA-8341)providesthedetailsandbases for our conclusion. Therefore, we consider your response to Generic Letter 83-28, Item 4.5.3 to be complete.
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9006130211 900605 PDR ADOCK 05000338 P
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n' Mr; W. L. Stewart
-2 June 5, 1990 This completes all staff activities related to TAC Nos. 54086 and 54087.
Sincerely, Original Signed By Leon B. Engle Project Manager Project Directorate 11-2 Division of Reactor Projects.1/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page D' STRIBUTLON:
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- 6/fi/90 OFFICIAL RECORD COPY Document Name: SER FOR GL 8328 l
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Mr. W. L. Stewart North Anna Power Station Virginia Electric & Power Company Units 1 and 2 l
Cc:
Mr. William C. Porter, Jr.
C. M. G. Buttery, M.D., M.P.H.
County Administrator Department of Health Louisa County 109 Governor Street l
P.O. Box 160 Richmond, Virginia 23219 Louisa, Virginia 23093 Regional Administrator, Region II Michael W. Maupin, Esq.
U.S. Nuclear Regulatory Cemmission Hunton and Williams.
101 Mtrietta Street N.W., Suite 2900 P. O. Box 1535 Atlan:a, Georgia 30323 Richmond, Virginia 23212 Mr. W. T. Lough Virginia Corporation Commission Mr. f., E. Kane, Manager Division of Energy Regulation Nortn Anna Power Station P. O. Box 1197 P.O. Box 402 Richmond, Virginia 23209 Mineral, Virginia 23117 l
Old Dominion Electric Cooperative Mr. J. P. O'Hanlon I
c/o Executive Vice President Vice President - Nuclear Services Innsbrook Corporate Center Virginia Electric and Power Company 4222 Cox Road, Suite 102 5000 Old Dominion Blvd.
l Glen Allen, Virginia 23060 Glen Allen, Virginia 23060 Mr. E. Wayne Harrell Mr. R. F. Saunders Vice President - Nuclear Operations Manager - Nuclear Licensing Virginia Electric and Power Co.
Virginia Electric and Power Company 5000 Old Dominion Blvd.
5000 Old Dominion Blvd.
Glen Allen, Virginia 23060 Glen Allen, Virginia 23060 l
Mr. Patrick A. O' Hare Office of the Attorney General Supreme Court Building 101 North 8th Street Richmucd, Virginia 23219 Senior Resident Inspector North Anna Power Station U.S. Nuclear Regulatory Commission Route 2, Box 78 Mineral, Virginia 23117 l
ENCLOSURE 1 ENCLOSURE 3 EGG NTA-8341 March 1989 TECHNICAL EVALUATION REPORT
/daho National 4 REVIEW OF REACTOR TRIP SYSTEM AVAILA81LITY ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, Engineering RESOLUTION Laboratory Managed
. by the U.S.
David P. Mackowiak John A. Schroeder Department of Energy i
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Prepared for the U.S. NUCLEAR REGULATORY COMMISSION
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No, Of AC07.?lV00!DO h
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NOTICE l
l-This report was prepared as an account of work sponsored by an agency of i
the Unised States Government. Nether the United Sases Govermeest not any
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agency thereof. aor any of their employees. Inakes say warranty, empressed I
or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use. of any informanoa, apparatus, product or proc.
ess disclosed in this report, or represents that its use by such third party would l
not infringe privately owned nghts.
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TECHNICAL EVALUATION REPORT: A REVIEW 0F REACTOR TRIP SYSTEM l
AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28, i
l ITEM 4.5.3, RESOLUTION l
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t David P. Mackowiak l
John A. Schroeder l
i EG&G Idaho, Inc.
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FIN 06001: Evaluation of Conformance to Generic Letter 83-28 for ors (Project 2) 1*
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ABSTRACT The Idaho National Engineering Laboratory (INEL) conducted a technical review of the commercial nuclear reactor licensees' responses to the requirements of the Nuclear Regulatory Commission's (NRC's)
Generic Letter 83-28 (GL 83-23), Item 4.5.3.
The results of this review, if all plants are shown to be covered by an adequate analysis, will provide the NRC staff with a basis to close out this issue with no further review.
The licensees, as the four vendors' Owners' Groups, submitted analyses to the NRC either directly in response to GL 83-28, Item 4.5,3, or to provide a basis for requesting changes to the Technical Specifications (TS) that would extend the Reactor Protection System (RPS) surveillance test intervals (STIs). To conduct the review, the INEL defined three criteria to determine the adequacy, plant applicability, and acceptability of the results.
The INEL examined the Owners Groups' reports to determine if the analyses and results met the established criteria.
Fort St. Vrain's responses to item 4.5.3 were also reviewed.
The INEL review results show that all licensees of currently operating commercial nuclear reactors have adequately demonstrated that their current on-line RPS test intervals meet the requirements of GL 83-28, Item 4.5.3.
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SUMMARY
The two anticipated transient without scram (ATVS) events at the Salem Nuclear Power plant in February of 1983, focused the attention of the Nuclear Regulatory Commission (NRC) on the generic implications of ATWS events. The NRC then published Generic Letter 83-28 (GL 83-23) which listed the actions the NRC required of all licensees holding operating licenses and others with respect to assuring the reliability of the Reactor Protection System (RPS). GL 83-28. Item 4.5.3, required licensees to demonstrate by review that the current on-line functional testing intervals are consistent with achieving high reactor trip system (RTS) availability. The licensees responded to the GL 83-28, Item 4.5.3, requirements as Owners Groups with reports either in direct response to J
Item 4.5.3, or with a technical basis for requesting-extensions to the surveillance test intervals (ST!s) that genera 11v included the Item 4.5.3 required reviews.
The NRC's Instrumentation and Control Systems Branch (ICSB), Office of Nuclear Reactor Regulation (NRR), requested the Idaho National Engineering Laboratory (INEL) to review the licensee availability analyses and evaluate the overall adequacy of the existing test intervals.
INEL review results showing general compliance with Item 4.5.3 will provide the NRC with a basis to close out Item 4.5.3 without 3
further review.
For the review, ths INEL defined three acceptance criteria, reviewed the licensees topical reports, contractor review reports, and NRC safety evaluations, and determined the adequacy of the analyses and the RTS availability estimates with regard to the review criteria.
The INEL review criteria to determine the licensees' Item 4.5.3 compliance were, (1) the five areas of concern of Item 4.5.3, (2) the analyses' plant applicability, and (3) the NRC's RTS electrical unavailability base case estimates from the ATWS Rulemaking Paper, SECY-83-293.
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Each Owners Groups' reports were reviewed to ensure that all five areas of concern from Item 4.5.3 were either included in the analyses or shown not to be significant with regard to RTS availability.
The INEL review also ensured that the individual plants' differences from the analysis' models were taken into account and their effects were shown not to significantly affect RTS unavailability. The Fort St. Vrain responses i
to item 4.5.3 were also reviewed.
The Owners Groups' RTS unavailability estimates were compared to the NRC's ATWS Rulemaking generic RTS unavailability estimates to determine the acceptability of the Owners Groups' conclusions that high RTS availability was demonstrated in the analyses.
The results of the INEL review showed that all licensees of currently operating commercial nuclear reactors have adequately demonstrated that their current on-line surveillance test intervals are consistent with achieving high RTS availability.
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s ACRONYMS ATWS Anticipated Transient Without Scram B&W Babcock & Wilcox BNL Brookhaven National Laboratory CE Cembustion Engineering GE General Electric HTGR High-Temperature Gas-Cooled Reactor ICSB Instrumentation and Control Systems Branch INEL Idaho National Engineering Laboratory LWR Light Water Reactor NFSC Nuclear Facility Safety Committee NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation PORC Plant Operations Review Committee PSC Public Service Company of Colorado PWR Pressurized Water Reactor RSSMAP Reactor Safety Study Methodology Applications Program RPS Reactor Protection System RTS Reactor Trip System SER Safety Evaluation Report l
STI Surveillance Test Interval l
TER Technical Evaluation Report W
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CONTENTS A B S T RA C T..............................................................
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SUMMARY
iii ACRONYMS..............................................................-
y 1.
INTRODUCTION.....................................................
1 1.1 Historical Background......................................
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-- 1,2 Review Purpose.............................................
3 2.
REVIEW CRITERIA..................................................
4 3.
REVIEW METHODOLOGY...............................................
6 4.
REVIEW RESULTS...................................................
7 4.1 B&W Plants.................................................
8 4.2 CE Plants..................................................
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4.3 GE Plants..................................................
9 4.4 Westinghouse Plants........................................
10 4.5 Quantitative Review of Vendors' RTS Unavailabilities......
11 4.6 Fort St. Vrain.............................................
14 5.
R E V I EW C O N C LU S I O N S............................................... 16 6.
REFERENCES.......................................................
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TABLES 1.
Comparison of Vendor and NRC RTS Unavailability Estimates........................................................
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4 TECHNICAL EVALVATION REPORT: A REVIEV 0F REACTOR TR!p SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28.
ITEM 4.5.3 RESOLUTION i
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INTRODUCTION 1.1 Historical Background In February of 1983, two events occurred at the Salem Nuclear Generating Station that focused Nuclear Regulatory Commission (NRC) attention on the generic implications of anticipated transient without scram (ATWS) events.
First, on February 22, during startup of Unit 1 an automatic trip
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signal generated as a result of a steam generator low-low level failed to cause a reactor scram. The reactor was tripped manually by an cperator i
almost coincidentally with the automatic trip signal, so the fact that the automatic trip had failed to cause a scram went unnoticed.
Three days later on February 25, both of the scram breakers at Unit 1 failed to open on an automatic reactor protection system (RPS) scram signal.
The operators took action to control this second ATWS and succeeded in terminating the incident in about 30 seconds. Subsequent investigation related the failure of the Unit 1 RPS to cause a scram to sticking of the undervoltage trip attachment in the scram circuit breakers.
As a result of these events the NRC Executive Director for Operations directed the staff to undertake three related activities: (1)an evaluation of when and under what conditions the Salem plants would be allowed to restart; (2) a fact finding report of the events at Salem 1 and the circumstances leading to them; and (3) a report on the generic implications of these events.
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To address (3) above an interoffice, interdisciplinary group was
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formed including members from the Office of Nuclear Reactor Regulation's I
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(NRR's) Divi.sion of Licensing Division cf Systems Integration, Division of Human Factors Safety Div'ision of Engineering, Division of Safety Technology,.the Office of Inspection and Enforcement, the Office for Analysis and Evaluation of Operational Data, and NRC's Region I Office.
1 This group published NUREG-1000 as a result of their efforts to resolve the following questions: (1) is there a need for prompt actions to address, similar equipment in other facilities; (2) are the NRC and its licensees learning the safety management lessons; and (3) how should the priority and content of the ATWS Rule be adjusted.
As a result of the NUREG-1000 findings, the NRC issued Generic Letter 83-282 (GL 83-28). The actions described in GL 83-28 address l
issues related to reactor trip system (RTS) reliability. The actions covered fall into the following four areas:
(1) Post-Trip Review, (2)
Equipment Classification and Vendor Interf ace, (3) Post-Maintenance Testing, and (4) Reactor Trip System Reliability Improvements.
Item 4, above, is aimed at assuring that vendor-recommended reactor trip breaker modifications and associated reactor protection system changes are completed in pressurized water reactors (PWRs), that a comprehensive program of preventive maintenance and surveillance testing is implemented for the reactor trip breakers in PWRs, that the shunt trip attachment I
activates automatically in all PWRs that use circuit breakers in their reactor trip systems, and to ensure that on-line functional testing of the reactor trip system is performed on all light water reactors (LWRs).
The specific requirements of GL 83-28, Item 4.5.3, are that existing intervals for on-line functional testing required by Technical Specifications shall be reviewed to determine if the intervals are consistent with achieving high RTS availability when accounting for considerations such as: (1) uncertaintier in component failure rates; (2) uncertainties in common mode failure rates; (3) reduced redundancy during testing; (4) operator errors during testing; and (5) component " wear-cut" caused by testing.
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The Babcock & Wilcox (B&W), Combustion Engineering (CE), General Electric (GE), and Westinghouse (W) Owners Groups have submitted topical reports either in response to GL 83 28, Item 4.5.3'3'" or to provide a basis for requesting RTS surveillance test interval (STI) extensions.5,6,7,8,9,10,11 In general, the owners groups' analyses were not done on a plant specific basis.
Instead, the analyses addressed a particular class of reactor trip system and then discussed the applicability of the analysis to specific product lines. The NRC reviewed these reports for, among other things, their applicability to GL 83-28 Item 4.5.3 and summarizac their findings in Safety Evaluation Reports 12'I3 (SERs).
1.2 Review Purpose
.This report documents a review of the Owners Groups' topical reports, the NRC SERs, and other analyses done at the Idaho National Engineering Laboratory (INEL) by personnel in the NRC Risk Analysis Unit of EG&G Idaho, Inc.
The INEL conducted the review at the request of the U.S. Nuclear Regulatory Commission, Office of Nucioar Reactor Regulation, Instrumentation and Control Systems Branch (ICSB).
The review was perfortred to determine if the Owners Groups' analyses demonstrated high RTS availability for the current test intervals, if the analyses included the five areas of concern from GL 83-28, and if all of the plants were covered by the analyses. The results of the review, if all plants are shown to be covered by an adequate analysis, would provide the NRC with a basis for closing out GL 83-28, Item 4.5.3, for all U.S. commercial nuclear reactors without further review.
The body of this report presents the review and its findings with regard to the stated objectives.
Section 2 describes the criteria used in the review to determine the adequacy of the analyses. The review l
methodology is discussed in Section 3.
Section 4 presents the review results. The review conclusions are given in Section 5.
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2.
RCVIEW CRITERIA i
To conduct a review, one must have criteria, or standards, on which a i
judgment or decisions nay be based.
In this section, the INEL availability
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analyses review criteria are presented.
GL 83-28 established the three criteria used in the INEL review.
GL 83-28 stated chat: (1) all licensees et al., (2) must demonstrate high RTS availability for the current test intervals by documented review when (3) accounting for such considerations as the five areas of concern listed
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in Section 1.1 While GL 83-28 established all three criteri), it only l
defined two of them-who had to do a review and what the review had to take into account.
The third and most subje:tive criterion, "high cvailability", was not defined.
To establish a definition of high availability, the INEL used the
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electrical unavailability base case estimates presented in Table A-1 of l
Appendix A to SECY-83-293.14 Unavailability is defined as 1.0 minus availability. A low unavailability is equivalent to a high availability.
Most analyses calculate a system unavailability rather than an i
availability.
Therefore, our criteria for a "high availability" will be expressed in terms of low unavailability for compatibility.
These RTS unavailability estimates from Reference 14 were used for two reasons.
First, they were used because they were developed by the NRC's ATWS Task Force as a reevaluation of the bases for the RTS unavailabilities used in ATWS rule talue-impact evaluations. Second, as stated in Reference 14,
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this NRC analysis l
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".. bases the RTS unavailabilities on worldwide experience to date.
It is believed that this gives a reasonable estimate of RTS unavailability that inc'udes the common cause contributions that are believed to dominate. The experience based values are distributed across the four vendor designs based on a i
comparative reliability analysis that evaluates the rajor i
differences among the designs."
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The estimates from the NRC ATWS analysis provide a framework with 4
- which to consider the topical' report Analyses estimates. The numerical'
- estimates in the SECY-83-293 for the four vendors combined with the five
- areas of cor.cern from'GL 83-28, Item 4.5.3, form the criteria used for this review to determine if the vendors' analyses and estimates met the requirements of Item 4.5.3.
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3.
REVIEW METHODOLOGY The INEL conducted this review by examining the vendors' topical reports (References 3, 4, 5, 6, 7, 8, 9, 10, and 11), the technical evaluation reports 15,16,17,18 (TERs) done as a part of the NRC. topical report review process, the NRC's SERs (References 12 and 13), and NUREG/CR-5197, Evaluation of Generic Issue 115, " Enhancement of Westinghouse Solid State Protection System."19 This was done for three First, the reports were examined to find out whether or not the reasons.
vendors' enalyses addressed the areas of concern from Item 4.5.3 and reflected a high RTS availability.
Second, they were examined to determine what plants were covered by the vendors' analyses.
Third, the Generic Issue 115 report provided an independent, updated estimate of the availability of the W solid state RTS for comparison to the review criteria.
For the plants covered by' the vendors' analyses or the JUREG/CR-5197_
analysis, the appropriate analysis and availability were comptred to the _
review criteria established in Section 2.
If the analysis aF luately addressed the areas of concern and demonstrated a high RTS availability, the plant' was accepted as having met the requirements of GL 83-28, Item 4.5.3.
The results of the comparisons for plants covered by a vendor analysis are given by vendor in Section 4.
For plants not directly covered by a vendor's analysis, an acceptable means was found to extend the analyses to cover the plants. This was done for-two plants: Clinton 1 (GE) and Maine Yankee (CE).
The means by which i
the analyses were extended to cover these two plants are also discussed by vendor in Section 4.
One plant, Fort St. Vrain, a high temperature, gas-cooled reactor (HTGR), was not covered by any of the four vendors' analyses and requireo special consideration. The INEL examined the responses from Fort St. Vrain required by GL 83-28, Item 4.5.3 to determine if the responses demonstrated an acceptably high RTS availability. The review of the Fort St. Vrain responses is given in Section 4.6.
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REVIEW RESULTS This section summarizes the results of the 1NEL review of the vendors' analyses with regard.to the five arets of concern and plant applicability.
The vendors' estimates of RTS availability are compared to the review availability criteria, Also, some insights concerning RTS availability, gained from an examination of RTS importance measures from selected PRAs, are examined.
j 4.1 B&W Plants 4
The issues of GL 83-28, Item 4.5.3, were addressed by the B&W Owners-Group and * % results were submitted to the NRC by the individual uti11 ties in their m mses to GL 83-28. Topical Report BAW-10167 (Reference 5) was submitted tc +.ne NRC to provide a technical basis for increasing the 1
on-line STIs and allowed outage times (A0Ts) for B&W RTS instrument I
strings. The analysis presented in BAW-10167 was built upon the previous analysis done to address the GL 83-28, Item 4.5.3 issues. However, some 1
~tnformation that was resolved in the generic letter analysis was not repeated in the subsequent Topical Report because it was not relevant to the proposed Technical Specification changes. To make BAW-10167 applicable to both GL 83-28, Item 4.5.3 and STI/A0T issues, the Owners Group submitted BAW-10167, Supplement 1 (Reference 6), to the NRC.. Supplement 1 completed the B&W analysis by addressing all remaining Item 4.5.3 issues. The BAW -10167 and Supplement 1 analyses included the implementation of the-automatic shunt trip on the reactor trip circuit breakers as required by GL 83-28, Item 4.3.
The INEL has previously reviewed the BAW-10167 and Supplement 1 analyses and documented the review in a TER, EGG-REQ-7718 (Reference 15).
For the TER, sensitivity studies which included all of the Item 4.5.3 areas of concern were conducted on the RTS models.
The sensitivity study results I
showed the models to be insensitive to variations in the failure rates associated with the Item 4.5.3 areas of concern.
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i The INEL reviewed BAW-10167, BAW-10167, Supplement'1, and the TER and determined that the B&W analyses adequately covered all five areas of' f
concern and that all currently operating B&W reactors are included.
4.2 CE Plants Licensees with CE reactors responded to the requirements of GL'83-28, Item 4.5.3, as the CE Owners Group by submitting CE NPSD-277 (Reference 3) to the NRC. The NPS0-277 RTS availability analysis specifically included all.five areas of concern and all currently operating CE reactors except Waterford 3, which was not in commercial operation until September 1985.
The CE Owners Group also submitted CEN-327 (Reference 7) to provide licensees with a basis for requesting RTS STI extensions. This 1 ster analysis expanded on the simplified models of NPSO-277 to include all RTS input parameters. All currently operating CE plants except Maine Yankee were covered in the CEN-327 analysis.
The CEN-327 STI analysis specifically included the NPSO-277 analyses of the Item 4.5.3 areas of concern except component wear-out" during testing.
The CEN-327 analysis showed that the major contributors to RTS unavailability for the four plant L
classes are common cause failures of the trip circuit breakers which are tested on a monthly basis, p
In both NPSD-277 and CEN-327, the CE RPS designs-are grouped into four i
ll classes by signal processing and trip device differences, otherwise the logic and physical layouts of the RTS are the same for all RTS plant p
classes.
In NPSD-2/7, Maine Yankee is included in RPS Plant' Class 2.
In CEN-327, Waterford 3 is included in RPS Plant Class 3.
Between NPS0-277 and CEN-327, all of the CE plants are included in plant classes analyzed in t
- CEN-327.
This review considers the analysis and results'in CEN-327 adequate for Item 4,5.3 resolution for all classes of CE plants.
The INEL has previously reviewed CEN-327 with regard to STI extension i
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-effects and documented the review in a TER,-EGG-REQ-7768 (Reference 16).
L The results of sensitivity studies cone for the TER show the models to be l
insensitive to an order of magnitude increase in the component independent J
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failure rates. The insensitivity to increased component failure rates V
along with the CE analysis results showing trip circuit breaker common I;
cause failures to be the major contributor to RTS unavailability provides a-
.a basis for this review to conclude that RTS test-induced component wear-out is not'an issue at CE reactors.
The INEL reviewed CEN-327 and the TER and determined that the CE analyses have adequately covered all five areas of concern or they have been shown not to contribute to RTS unavailability and that all currently operating CE reactors are included.
4.3 GE Plants Licensees with GE reactors responded to the GL 83-28, Item 4.5.3 requirements as the BWR Owners' Group by submitting NECD-30844 (Reference'4) to the NRC.
The RTS availability analysis specifically included the five areas of concern and covered both generic relay and solid-state RTS designs which includes all currently operating BWRs. GE stated that the relay RPS configurations for BWR plants have the same primary design features. Therefore, the generic relay RTS models used in NECD-30844 do not differ significantly from the specific BWR plants. GE used the Clinton 1 drawings for the solid-state RTS models. Since Clinton 1 is currently the only GE plant with a solid state RTS, no plant unique analysis is'necessary.
The BWR Owners' Group also submitted NEC0-30851P (Reference 8) to the
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NRC. The analysis in this second report used the base case results from NECD-30844 to establish a basis for requesting revisions to the current Technical Specifications for the RTS. The INEL had previously reviewed NECD-30844 and NECD-30851P with regard to both Item 4.5.3 and STI extension acceptability and documented the review in a TER, EGG-EA-7105 (Reference 17). Due to insufficient information, the INEL review could not complete the solid-state RTS review and accepted only the relay'RTS analysis results. The NRC reviewed the topical reports and the TER and 9
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.c issued an SER (Reference 12). ' The NRC accepted the analysis results as a reference for TS chang'es related to the RTS and as resalution to GL 83-28, Item 4.5.3, for GE relay plants only. The INEL later completed the solid state RTS analysis review and issued Rev-li o the TER (Reference 18), thus-t L
accepting the analyses for all. classes of GE plants, l
This review' examined both GE analyses and the Rev 1 TER and determined that all five areas of concern are. included in the analyses and that all currently operating GE reactors are included, i
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4.4: Westinghouse plants L
Licensees with Westinghouse reactors did not respond directly to the-requirements ~of GL 83-28, Item 4.5.3.
Prior to the Salem ATWS, they had submitted WCAP-10271 (Reference 9) to the NRC to provide a basis for requesting changes to the Technical Specifications regarding the RTS. The Westinghouse methodology attempted to' balance safety and operability and l
was applied to a' typical Westinghouse four loop reacter plant with a solid state RTS in WCAP-10271.
The methodology was extended to cover RTSs for 1
two, three, and four loop plants with either relay or solid state logic in l
WCAP-10271, Supplement 1-(Reference 10).
t l-The NRC reviewed the Westinghouse topical reports with the assistance of Brookhaven National Laboratory (BNL) and issued an SER (Reference 13) l; limiting their acceptance to changes to only the-analog channel STIs at l
Westinghouse plants.
The W methodology used fault trees to model the RTS. The models
' ncluded the following five major contributors to RTS trip unavailability:
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Unavailability of components due to random failures 2.
Unavailability of components due to test p
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Unavailability of components due to unscheduled maintenance i
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Unavailability of components due to human error i
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Unavailability of components due to common cause failure.
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.While the y analysis did not directly include any sensitivity studies concerning these five areas, the component unavailabilities were increased as the test interval length increased. The STI analysis results showed a factor of'3 to 5 increase in the RTS unavailability estimates-for the longer test interval.
Two conservatisms exist in the models that are relevant:
first, no credit was taken for early failures that would be detected and, second, no credit wa: taken for the diversity inherent in the W RTS design. These two conservatisns, had they been included in the model, would cause the increase in the RTS unavailability estimates to be smaller than the observed factors.
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Test-induced component wear-out was not addressed in any manner in the l,
y RTS analysis. However, the RTS analyses done by the other vendors, j
L References 3, 4 and 6, specifically investigated the effects of this issue on RTS unavailability. Despite the differences among the other vendors' RTS designs,.they all found the effects of test induced component wear-cut L
on RTS unavailability to be insignificant. Based on the other vendors' L
analyses, the INEL concluded that the effects of test-induced component wear-out on y RTS' unavailability would also be insignificant.
Therefore, the INEL considers all y plants to be covered by adequate analyses.
4.5 Quantitative Review of Vendors' RTS Availabilities i
So. far, only the adequacy of the vendors' analyses has been discussed. No determination has been made of the acceptability of the numerical estimates from the various RTS availability analyses.
In this section, the INEL review considers the four Owners Groups' RTS availability-estimates to determine if they are indeed indicative of "high availability."
-l 11 g
4 i
1 In' Table 1, the four vendors' RTS unavailability estimates are compared to the review estimates of low unavailability as defined-in Section 2.
The B&W and GE vendors' estimates are given as an overall RTS unavailability per demand by plant model and RTS type, respectively. The CE and W vendors' estimates are given on a similar basis with an additional consideration that was not necessary for the B&W and GE analyses.
In the CE and W analyses, RTS unavailability was estimated for all input parameters.
For the CE and W unavailability estimates _in Table 1, the INEL-used the unavailability estimates for high pressurizer pressure, the 4
parameter analyzed in Reference 19 as the limiting parameter for an ATWS in terms of the number of input channels and diversity of trip signal.
The differences in the relative values of the three PWR vendors' RTS-1 unavailability estimates can be attributed to design differences among the t
RTSs. - B&W and CE RTSs have four analog channel inputs for each monitored parameter with four trip logic channels while W RTSs have three or four analog channel inputs for each parameter with only two-trip logic
]
channels. The 2 of 4 analog channels for the B&W and CE RTS designs are inherently more re N ie than the 2 of'3 analog channels for some parame'ters in the W design. Also the 2 of 4 trip logic in the B&W and
. CE RTSs is more reliable ~than the W 1 of 2 trip logic.
The combination of these two design differences make the W RTS unreliability'somewhat higher than the other vendors' RTS unavailabilities.
1 The comparison shows the b&W, CE, and GE RTS unavailability estimates are: lower than the NRC's estimates while the W estimates are.the same as
~ the NRC's.
The INEL review recognizes the Vendors' estimates and the NRC's
- estimates are influenced by a number of' factors. These factors include, j
('1) the data uncertainties for both the NRC and Vendors analyses, (2) the scarcity of actual RTS failures world wide, (3) the modeling assumptions and simplifications used by both the,NRC and the Vendors, and (4) the differing levels of model development between the _NRC analysis and the Vendors' analyses and between different Vendors' analyses. These factors 12
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TABLE 1.. COMPARISON OF VENDOR AND NRC RTS UNAVAILABILITY ESTIMATES" r
Vendor.R'TS NRC RTS e
b Unavailability Estimates Unavailability Estimates Vendor (Failures / Demand)
(Failures / Demand)
B&W l
Davis Bessie Model 1E-10" 3E-5d Oconee Class Model 1E-6" 3E-5d CE Plant Class'l 2E-7' 2E-5 Plant-C1 ass.2 3E-6' 2E-5 Plant Class 3 3E-6' 2E-5 i
Relay Plants 3E-6 2E-5 i
I Solid-state Plants 3E-6 2E-5 h
M Relay Plants SE-59 d
SE-5 Solid-state Plants SE-59 d
SE-5 a.
All estimates are rounded off to one'significant digit.
b From Reference 14, Table A-1, base case RTS_ electrical unavailability estimates.
c.
From Reference 5, base case.
d.
Includes automatic shunt trip on the reactor trip circuit breakers.
e.
From Reference 7, Tables 4.1-1, 4.2-2, 4.1-3, and 4.1-4, respectively; base case test interval, high pressurizer pressure unavailability estimate.
f.
From Reference 4.
g.
From Reference 19, solid state RTS base case. Applied to relay plants based on similarity of design (see Reference 11, Section 3.2.2 and 3.2.3).
13 i
help explain the differences between the Vendors' and the NRC's point estimates of'RTS availability.
4.6 Fort St. Vrain X
L Fort St. Vrain responded to GL 83-28, Item 4.5.3 in a letter to 20 Eisenhut dated November 4, 1983
, stating:
" Existing intervals for on-line functional testing required by the Technical Specifications are-currently under.
review by Public Service Company of Colorado (PSC) and the r
L Nuclear Regulatory Commission Region IV staff. The current J
testing frequency at Fort St. Vrain has been dictated by the 1
Nuclear Regulatory Commission staff." (Underline added)
In response to a request for information from_the NRC concerning the Fort St. Vrain responses to GL 83-28 previously sent, PSC sent the 21 following reply to the NRC in a lett'er to Johnson, dated June 12, 1985
" Existing intervals for the on-line testing required by the Technical Specifications were reviewed by Public Service Company-of Colorado. A Technical Specification change to Limitin Conditions for Operation 4.4.1 (Plant Protective System) gand its associated survo111ance requirements (SR 5.4.1) are currently being reviewed by the Plant Operations Review Committee (PORC).
This Technical Specification change is expected to be approved by i
the PORC and the Nuclear Facility Safety Committee (NSFC) by June 30, 1985.. As part of the development process for these proposed changes to the Technical Specifications, on-line functional I
testing requirements were_ reviewed based on past experience.
l Possible changes' to the testing intervals in certain cases where l
available test data may support such changes has (sic) been p
discussed at length with the Nuclear Regulatory Commission E
staff. The Nuclear Regulatory Commission staff has informed Public Service Company of Colorado that no-such changes would be acceptable at this time."
t The INEL review. interpreted these responses ~from Fort St. Vrain to l
.mean the NRC has established Fort St. Vrain's RTS current test intervals, l.
the current' test intervals have been evaluated by PSC, and the NRC will not L
allow changes.to the test intervals at this time.
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'From these responses, the INEL concluded that Fort St. Vrain has.
conducted the review required by GL 83-28, Item 4.5.3, and that the NRC-considers the PSC and NRC reviews. adequate-to meet the Item 4'5.3
~~
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{L' it-5 REVIEW CONCLUSIONS-1 All four LWR' vendors have. submitted topical; reports either in response to GL 83-28, Item 4.5.3, or to provide a basis for RTS STI extensions, or both.
For the most part, these reports have addressed all of the issues in Item 4.5.3.
Licensees not covered by the topical reports have submitted individual. responses to Item 4.5.3.
The analyses in the topical report have shown the currently configured RTSs to be highly reliable with the' current test intervals and prior.to implementing some of the~ requirements 'of GL 83-28.
Implementation of these additional requirements will. reduce the ATWS risk even further.
The INEL has ' reviewed the relevant topical reports, TERs, SERs,-
additional. analyses, and the individual licensee'submittals with regard to 7
GL 83-28, Item 4.5.3, requirements and the review criteria. Based on that review, the INEL concludes that all licensees of currently operating commercial nuclear power plants have adequately demonstrated that their current RTS test intervals are consistent with achieving high RTS availability.
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REFERENCES Lj
_1. :
-U.S.- Nuclear Regulatory Commission, Generic Implications of ATWS Events l!
at the Salem Nuclear Power Plant, NUEEG-1000, April 1983.
.2.
U.S. Nuclear Regulatory Commission Letter. D. G. Eisenhut to All i
-Licensees et al., Required Actions Based on Generic Imt,lications of Salem ATWS Events, Generic Letter 83-28, July 8,1983.
~
3.
Combustion Engineering, Reactor Protection System Test Interval Evaluation. Task 486, CE NPSD-277, December 1984.
u 4.
S. Visweswaran et al., BWR Owners'-Group Response to NRC Generic-Letter 83-28, Item 4.5.3, NECD-30844, January 1985.
5.
R. S. Enzinna et al., Justification for increasing the Reactor Trip
[
System On-line Test Interval, BAW-10167, May 1986.
i 6.
R. S. Enzinna et al., Justification for increasing the Reactor Trio System On-line Test Interval, Su))lement Numoer 1, BAW-10167, Supplement Number 1, February 19M.
l:
7.
Combustion Engineering, RPS/ESFAS Extended Test Interval Evaluation,
~
CEN-327, May 1986.
L 8.
W.'
P.- Sullivan et al., Technical Specification Improvement Analyses for BWR Reactor Protection System, NEC)-30851P, May 1985.
9.
R. L. Jansen.et al., Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor-Protection Instrumentation System, WCAP-10271, January 1983, b
10.
R. L. Jansen et al., Evaluation of Surveillance Frecuencies and Out of l'
Service Times for the Reactor Protection Instrumentation System.
Supplement 1, WCAP-10271, Supplement 1, July 1983.
.11.
R. L. Jansen et al., Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System.
i l
supplement 1-P-A,.WCAP-10271, Supplement 1-P-A, May 1986, 12.
U.S. Nuclear Regulatory Commission Memorandum, G. C. Lainas to E. J.
t L
Butcher, Acceptance for Referencing of General Electric Company (GE)
Topical Reports NECD-30844. "BWR Owners' Group Response to NRC Generic Letter 83-28," and NECD-30851P. " Technical Specification Improvement
' Analyses for BWR Reactor Protection System." April 28, 1986.
13.
U.S. Nuclear Regulatory Commission. Letter, C. O Thomas to J. J.
'Sheppard, Acceptance for Referencing of Licensing Topical Report 1
WCAP-10271. " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation Systems." February 21, 1985.
n l.
i 17
J u:
14.
U.S. Nuclear Regulatory Commission, Amendmerts to 10 CFR 50 Related to Antietaated Transients Without Scram (ATWS) Tvi'ts, SECY-83-293, July 19, 1933.
15.
J. P. Poloski and S. O. Matthews, Review of B&W Owner's Group Analyses for-Increasing The Reactor Trip Syste..n On-line Test Interval, EGG-REQ-7718, September 1988.
16.
O. P. Mackowiak and B. L. Collins, A Review of the Combustion Engineering Evaluation For Extending the RPS and ESFAS Test Intervals, EGG-REQ-7768, September 1988.
i 17, R. E. Wright and B. L. Collins, A Review of the BWR Owners' Group Technical Specification Improvement Analyses for the BWR Reactor Protection System, EGG-EA-7105, January 1986.
~
18.
R. E. Wright and B. L. Collins, A Review of the BWR Owners' Group Technical Specification Improvement Analyses for-the 8WR Reactor Protection System, EGG-EA-7105, Rev 1. March 1987.
l 19.-
D. A. Reny et al., Evaluation of Generic Issue 115, Enhancement of the l
Reliability of Westinghouse Solid State Protection Systems, NUREG/CR-5197, January 1989.
j ji 20.
Public Service Company of Colorado Letter, O. R. Lee to D. G.
)
Eisenhut,' Response to Generic Letter 83-28, November 4, 1983.
21 ~.
Public Service Company of Colorado Letter, J. W. Gham to E. H.
Johnson, Response to Generic Letter 83-28. June-12, 1985.
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TECHNICAL EVALUATION REPORT: A REVIEW 0F REACTOR TRIP SYSTEM AVAILABILITY ANALYSES FOR GENERIC LETTER 83-28.
j ITEM 4.5.3, RESOLUTION
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Regulatory and Technical Assistance I
EG&G Idaho, Inc.
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P. O. Box 1625 Idaho Falls, ID 83415 06001
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Division of Engineering and System Technology Office of Nuclear Reactor Regulation
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U.S. Nuclear Regulatory Comission Washington DC 20555
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u sv.cs m The Idaho National Engineering Laboratory (INEL) conducted a technical review of
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the commercial nuclear reactor licensees' responses to the requirements of the-Nuclear Regulatory Comission's (NRC's) Generic Letter 83-28 (GL 83-28) Item 4.5.3.
The results of this review, if all plants are shown to be covered by an adequate analysis, will i
provide the NRC staff with a basis to close out this issue with no further review.
The licensees, as the four vendors' Owners' Groups, submitted analyses to the NRC either i
directly in response to GL 83-28. Item 4.5.3, or to provide a basis for requesting changes !
to the Technical Specifications (TSs) that would extend ths hactor Protection System (RPS)' surveillance test intervals (STIs). To conduct the review, the INEL defined three L
criteria to determine the adequacy, the plant applicability, and the acceptability of the results.
The INEL examined the Owners Groups' reports to determine if the analyses and results met the established criteria.
Fort St. Vrain's responses to Item 4.5.3 l
were also reviewed.
The INEL review results show that all licensees of currently opera-ting commercial nuclear reactors have adequately demonstrated that their current on-line.
RPS test intervals meet the requirements of GL 83-28, Item 4.5.3.
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