ML18041A209

Review of Reactor Trip Sys Availability Analyses for Generic Ltr 83-28,Item 4.5.3,Resolution, Technical Evaluation Rept

ML18041A209
Person / Time
Site:	FitzPatrick
Issue date:	03/31/1989
From:	Mackowiak D, Schroeder J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML18041A210	List: ML18041A209
References
CON-FIN-D-6001, CON-IIT07-438-91, CON-IIT7-438-91 EGG-NTA-8341, EGG-NTA-8341-01, EGG-NTA-8341-1, GL-83-28, NUDOCS 8911060030
Download: ML18041A209 (56)

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ENCLOSURE 2 EGG-NTA-8341 March 1989 TECHNICAL EVALUATIONREPORT Idaho A REVIEW OF REACTOR TRIP SYSTEM AVAILABILITY National ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, Engineering RESOLUTION Laboratory Managed David P. Mackowiak by the U.S. John A. Schroeder Department of Energy

~~ E'RzLP i~ Prepared for the n

U.S. NUCLEAR REGULATORY COMMISSION Worl pen'onned vnder Dof Conuect No. D&AC07.78ID01570

4 ENCLOSURE 3 EGG-NTA-8341 March 1989 TECHNICAL EVALUATIONREPORT Idaho A REYIEW OF REACTOR TRIP SYSTEM AVAILABILITY National ANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, Engineering RESOLUTION Laboratory Managed Oavid P. Mackowiak by the U.S. John A. Schroeder Oeparrment of Energy n EGzGi~ Prepared for the

~~ U.S. NUCLEAR REGULATORY COMMISS/ON Worl pen'onned under

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DOE Conuect No. Df.AC07-76ID01570

s NOTICE

. This report was prepared as an account of work sponsored by an agency of the United States Government. neither the United Sates Government nor any agency thereof, nor any of their employees, makes any warranty. expressed or implied. or assumes any legal liabilityor responsibiTity for any third party' use. or the results of such use, of any information, apparatus. product or proc-ess disclosed in this report. or represents that its use by such third party would not infringe privately owned nghts.

s' EGG-NTA-8341 TECHNICAL EVALUATION REPORT: A REVIEW OF REACTOR TRIP SYSTEM AVAIIABILITYANALYSES FOR GENERIC LETTER 83-28, ITEM 4.5.3, RESOLUTION Oavid P. Mackowiak John A. Schroeder EGEG Idaho, Inc.

Idaho Falls, Idaho 83415 FIN 06001: Evaluation of Conformance to Generic Letter 83-28 for ORs (Project 2)

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ABSTRACT The Idaho National Engineering Laboratory (INEL) conducted a.

technical review of the commercial nuclear reactor licensees'esponses to the requirements of the Nuclear Regulatory Commission'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 provide the NRC staff with a basis to close out this issue with no further review. The licensees, as the four vendors'wners'roups, 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 to determine if Groups'eports 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 mee. the requirements of GL 83-28, Item 4.5.3.

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SUMMARY

The two anticipated transient without scram (ATWS) 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 83-28) 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 '.3, requirements as Owners Groups with reports either in direct response to Item 4.5.3, or with a technical basis for requesting extensions to the surveillance tes. intervals (STIs) that generally 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 further review.

For the review, the INEL defined th~ee 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'tem 4.5.3 compliance were, ( I) the five areas of concern of I>em 4.5.3, (2) the analyses'lant applicability, and ( 3) the NRC's RTS electrical unavailability base case estimates from the ATWS Rul,emaking Paper, F

SECY"83-293.

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Each Owners Groups'eports were reviewed to ensure that all five areas of concern from Item 4.5.3 were either included in th'e analyses or.

shown not to be significant with regard to RTS availability. The INEL review also ensured that the individual plants'ifferences from the analysis'odels were taken into account and their effects were shown not to significantly affect RTS unavailability. The Fort j St. Vrain responses to Item 4.5.3 were also reviewed'he Owners Groups'TS unavailability estimates were compared to the NRC's ATWS Rulemaking generic RTS unavailability estimates to determine the acceptability of the Owners Groups'onclusions 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.

l CONTENTS ABSTRACT

SUMMARY

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1. INTROOUCT ION 1.1 Historical Background .........................,...... 1 1.2 Review Purpose ~ ~ ~ ~ ~ ~ ~ ~ ~ 3

2. REVIEW CRITERIA .. 4 3, REVIEW METHOOOLOGY

4. REVIEW RESULTS ...

4.1 BLW Plants 8 4.2 CE Plants . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 4.3 ~

GE Plants . 9 4.4 Westinghouse Plants ..... ~ . 10 4.5 guantitative Review of Vendors'TS Unavailabilities 4.6 Fort St. Vrain 14

5. REVIEW CONCLUSIONS ..... ~...........

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6. REFERENCES ...............,... ,.......,

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~ ~ ..............'. .., ~ 17 TABLES

1. Comparison of Vendor and NRC RTS Unavailability Estimates ..... 13

I ACRONYMS ATWS Anticipated Transient Without Scram B&W Babcock 5 Wilcox BNL Brookhaven National Laboratory CE Combustion 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 Evalua ion Report STI Surveillance Test Interval TER Technical Evaluation Report Westinghouse

l TECHNICAL EVALUATION REPORT: A REVIEW OF REACTOR TRIP SYSTEM AVAILABILITYANALYSES FOR GENERIC LETTER 83-'28 ITEM 4.5. 3 RESOLUTION

1. INTRODUCTION

l. 1 Historical Back round 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 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 operator 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.

To address (3) above an interoffice, interdisciplinary-group was formed includ ng members from the Office of Nuclear Reactor Regulation's

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(NRR's) Division of Licensing, Division of Systems Integration, Division of Human Factors Safety, Division of Engineering, Division of'afety Technology, the Office of Inspection and Enforcement, the Office for Analysis and Evaluation of Operational Data, and NRC's Region I Office.

This group published NUREG-1000 1 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 2

Letter 83-28 (GL 83-28). The actions described in GL 83-28 address 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 Interface, (3) Post"Maintenance Testing, and (4) Reactor Trip System Reliability Improvements.

Item 4, above, is a,imed 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 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) uncertainties 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-out" caused by testing.

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The Babcock 8 Wilcox (BEW), 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' 3,4 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'nalyses 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 summarized their findings in Safety Evaluation Reports 'SERs).

This report documents a review of the Owners Groups'opical reports, the NRC SERs, and other analyses done at the Idaho National Engineering Laboratory ( INEL) by personnel in the NRC Risk Analysis Unit of EGLG Idaho, Inc. The INEL conducted the review at the request of the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Instrumentation and Control Systems Branch (ICSB). The review was performed to determine if the Owners Groups'nalyses demonstrated high RTS availability for the current test intervals, if the analyses included the five areas of concern from GL 83-28, and tf 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 analyhes. The review 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. REVIEW CRITERIA To conduct a review, one must have criteria, or standards, on which a judgment or decisions may be based. In this section, the INEL availability analyses review criteria are presented.

GL 83-28 established the three criteria used in the INEL review.

GL 83-28 stated that: (1) all licensees et al., (2) must demonstrate high RTS avai'>ability for the current test intervals by documented review when (3) accounting for such considerations as the five areas of concern listed in Section 1. 1. While GL 83-28 established all three criteria, it only

defined two of them who had to do a review and what the review had to take into account. The third and most subjective criterion, "high availability", was not defined.

To establish a definition of high availability, the INEL used the electrical unavailability base case estimates presented in Table A-1 of 14 Appendix A to SECY-83"293. 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 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 value-impact evaluations. Second, as stated in Reference 14, this NRC analysis

"...bases the RTS unavailabilities on worldwide experience to date. It is believed'that this gives a reasonable estimate of RTS unavailability that includes the common cause contributions that are believed to dominate. The experience based values are distributed across the four vendor designs based on a comparative reliability analysis that evaluates the major Cif'erences among the designs."

The estimates from the NRC ATWS analysis provide a framework with 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 concern from GL 83-28, Item 4.5.3, form the criteria used for this review to determine if the vendors'nalyses and estimates met the requirements of Item 4.5.3.

3. REVIEW METHODOLOGY The INEL conducted this review by examining the vendors'opical 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 reasons. First, the reports were examined to find out whether or not the vendors'nalyses 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'nalyses. 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'nalyses or the NUREG/CR-5197 analysis, the appropriate analysis and availability were compared to the review criteria established in Section 2. If the analysis adequately 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 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'nalyses and required 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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4. REVIEW RESULTS This section summarizes the results of the INEL review of the vendors'nalyses with regard to the five areas of concern and plant applicability.

The vendors'stimates 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,

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are examined.

4.1 B&W Plants The issues of GL 83-28, Item 4.5.3, were addressed by the B&W Owners Group and the results were submitted to the NRC by the individual utilities in their responses to GL 83-28. Topical Report BAW-10167 (Reference 5) was submitted to the NRC to provide a technical basis for increasing the on-line STIs and allowed outage times (AOTs) for B&W RTS instrument strings. The analysis presented in BAW-10167 was bu'ilt upon the previous analysis done to address the GL 83-28, Item 4.5.3 issues. However, some information 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/AOT 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-RE/-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 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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. The INEL reviewed,BAW-10167, BAW-10167, Supplement 1, and the TER and determined that the iKW analyses adequately covered all five areas of concern and that all currently operating BKW reactors are included'.

4.2 CE Plants Licensees with CE reactors responded to the'equirements of GL 83-28, Item 4.5.3, as the CE Owners Group by submitting CE NPSD-277 (Reference 3) to the NRC. The NPSD-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 later analysis expanded on the simplified models of NPSD"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 NPSD-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 classes are common cause failures of the trip circuit breakers which are tested on a monthly basis.

In both NPSD"277 and CEN"327, the CE RPS designs are grouped into four classes by signal processing and trip device differences, otherwise the logic and physical layouts of the RTS are the same for all RTS plant classes. In NPS0-277, Maine Yankee is included in RPS Plant Class 2. In CEN-327, Waterford 3 is included in RPS Plant Class 3. Between NPSD-277 and CEN-327, all of the CE plants are included in plant classes analyzed in 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 effects and documented the review in a TER, EGG-REQ-7768 (Reference 16).

The results of sensitivity studies done for the TER show the models to be insensitive to an order of magnitude increase in the component independent

fai lure rates. The insensitivity to increased component failure. rates along with the CE analysis results showing trip circuit breaker common 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 RTSi 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'roup 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'roup also submitted NECD-30851P (Reference 8) to the 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

I issued'an SER (Reference 12). The NRC accepted the analysis results as' reference for TS changes related to the RTS and as resolution 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 1 to the TER (Reference 18), thus accepting the analyses for all classes of GE plants.

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, 4.4 Westin house Plants 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 was applied to a typical Westinghouse four loop reactor plant with a solid state RTS in WCAP-10271. The methodology was extended to cover RTSs for two, three, and four loop plants with either relay or solid state logic in WCAP-10271, Supplement 1 (Reference 10).

The NRC reviewed the Westinghouse topical reports with the assistance of Brookhaven National Laboratory (BNL) and issued an SER (Reference 13) limiting their acceptance to changes to only the analog channel STIs at Westinghouse plants.

The W methodology used fault trees to model the RTS. The models included the following five major contributors to RTS trip unavailability:

1. Unavailability of components'ue to random failures

2. Unavailability of components due to test 10

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3. Unavailability of components due to unscheduled maintenance

4. Unavailability of components due to human error

5. Unavailability of components due to common cause failure.

While the W 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 was taken for the diversity inherent in the W RTS design. These two conservatisms, had they been included in the model, would cause the increase in the RTS unavailability estimates to be smaller than the observed factors.

Test-induced component wear-out'as not addressed in any manner in the W RTS analysis. However, the RTS analyses done by the other vendors, References 3, 4 and 6, specifically investigated the effects of this issue on RTS unavailability. Despite the differences among the other vendors'TS designs, they all found the effects of test induced component wear-out on RTS unavailability to be insignificant. Based on the other the INEL concluded that the effects of test-induced 'component vendors'nalyses, wear-out on W RTS unavailability would also be insignificant. Therefore, the INEL considers all W plants to be covered by adequate analyses.

4.5 uantitative Review of Vendors'TS Avai labilities So. far, only the adequacy of the vendors'nalyses 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'TS availability es imates to determine if they are indeed indicative of "high availability."

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In Table I, the four vendors'TS unavailability estimates are compared to the rgview estimates of low unavailability as defined in Section 2. The B&W and GE vendors'stimates are given as an overall RTS unavailability per demand by plant model and RTS type, respectively. The CE and W vendors'stimates 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 I, the INEL used the unavailability estimates for high pressurizer pressure, the 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'TS unavailability estimates can be attributed to design differences among the RTSs. 8&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 8&W and CE RTS designs are inherently more reliable than the 2 of 3 analog channels for some parameters in the W design. Also the 2 of 4 trip logic in the B&W and CE RTSs is more reliable than the W I of 2 trip logic. The combination of these two design differences make the W RTS unreliability somewhat higher than the other vendors'TS unavailabilities.

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'stimates and the NRC's estimates are influenced by a number of factors. These factors include, (I) 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'nalyses and between different Vendors'nalyses. These factors 12

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TABLE 1. COMPARISON OF VENDOR AND NRC RTS UNAVAILABILITYESTIMATES Vendor RTS NRC RTS Unavailability Estimates Unavai 1 abi l i ty Estimates Vendor Failures/Demand Fai lures/Demand B8W Davis Bessie Model 1E-10 3E-5 Oconee Class Model 1E-6 3E-5 CE Plant Class 1 2E"7 2E-5 Plant Class 2 3E-6 2E-5 Plant Class 3 3E-6 2E-5 Plant Class 4 2E-6 2E-5 GE Relay Plants 3E-6f 2E-5 Solid-state Plants 3E-6f 2E-5 Relay Plants 5E-5g 5E-5d Solid-state'Plants 5E"5g 5E"5

a. All estimates are rounded off to one significant digit.

b. From Reference 14, Table A-l, 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).

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help explain the differences between the Vendors'nd the NRC's point estimates of RTS availability.

4.6 Fort St. Vrain 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 Nuclear Regulatory Commission Region IV staff. The current testin fre uenc at Fort St. Vrain has been dictated b the Nuclear Re viator 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 following reply to the NRC in a letter 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 Limiting Conditions for Operation 4.4. 1 (Plant Protective System) and its associated surveillance 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 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 testing requirements were reviewed based on past experience.

Possible changes to the testing intervals in certain cases where available test data may support such changes has (sic) been discussed at length with the Nuclear Regulatory Commission staff. The Nuclear Regulatory Commission staff has informed Public Service Company of Colorado that no such changes would be acceptable at this time."

The INEL review interpreted these responses from Fort St. Vrain to mean the NRC has established Fort St Vrain's RTS current test intervals,

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the current test intervals have been evaluated by PSC, and the NRC will not allow changes to the test intervals at this time.

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I'0 From these 'responses, the. INEL concluded that Fort St. Vrain has conducted the reviey 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 requirements.

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5 REVIEW CONCLUSIONS 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 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 avai 1 abi 1 i ty.

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6. REFERENCES U.S. Nuclear Regulatory Commission, Generic Im lications of ATWS Events at the Salem Nuclear Power Plant, NUREG-1000, April 1983.

U.S. Nuclear Regulatory Commission Letter, 0. G. Eisenhut to All Licensees et al., Re uired Actions Based on Generic Im lications of Salem ATWS Events, Generic Letter 83-28, July 8, 1983.

Combustion Engineering, Reactor Protection S stem Test Interval Evaluation Task 486, CE NPS0-277, Oecember 1984.

S. Visweswaran et al., BWR Owners'rou Res onse to NRC Generic Letter 83-28 Item 4.5.3, NECD"30844, January 1985.

R. S. Enzinna et al., Justification for Increasin the Reactor Tri S stem On-line Test Interval, BAW-10167, May 1986.

R. S. Enzinna et al., Justification for Increasin the Reactor Tri S stem On-line Test Interval Su lement Number 1, BAW-)0167, Supplement Number 1, February 1988.

Combustion Engineering, RPS/ESFAS Extended Test Interval Evaluation, CEN-327, May 1986.

W. P. Sullivan et al., Technical S ecification Im rovement Anal ses for BWR Reactor Protection S stem, NEC0-30851P, May 1985.

R. L. Jansen et al., Evaluation of Surveillance Fre uencies and Out of Service Times for the Reactor Protection Instrumentation S stem, WCAP-10271, January 1983.

R. L. Jansen et al., Evaluation of Surveillance Fre uencies and Out of Service Times for the Reactor Protection Instrumentation S stem R. L. Jansen et al., Evaluation of Surveillance Fre uencies and Out of Service Times for the Reactor Protection Instrumentation S stem Su lement 1-P-A, WCAP-10271, Supplement 1-P-A, May 1986.

U.S. Nuclear Regulatory Commission Memorandum, G. C, Lainas to E. J.

Butcher, Acce tance for Referencin of General Electric Com an GE To ical Re orts NECO-30844 BWR Owners'rou Res onse to NRC Generic Letter 83-28 and NECD"30851P Technical S ecification Im rovement Anal ses for BWR Reactor Protection S stem April 28, 1986.

U.S. Nuclear Regulatory Commission Letter, C. 0 Thomas to J. J.

Sheppard, Acceptance for Referencin of Licensin To ical Report WCAP-10271 Evaluation of Surveillance Fre uencies and. Out of Service Times for the Reactor Protection Instrumentation S stems February 21, 1985.

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The licensees, as the four vendors'wners'roups, 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 (TSs) that would extend the Reactor Protection System (RPS) surveillance test intervals (STIs). To conduct the review, the INEL defined criteria to determine the adequacy, the plant applicability, and the acceptability of three the resu'its. The INEL examined the Owners Groups'eports'o determine and results met the established criteria. Fort St. Vrain's responses to Item 4.5.3 if the analyses 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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