ML20245B350

Discusses NRR Plans for Approval of GE Topical Repts NEDC-30844, BWR Owners Group Response to NRC Generic Ltr 83-28 & NEDC-30851P, Tech Spec Improvement Analysis for BWR Rps. Draft Reply & SER Encl

ML20245B350
Person / Time
Issue date:	05/27/1987
From:	Murley T Office of Nuclear Reactor Regulation
To:	Jordan E NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
Shared Package
ML20245B353	List: ML20245B350 ML20245C200 ML20245C269 ML20246C992
References
GL-83-28, NUDOCS 8706100358
Download: ML20245B350 (36)

Text

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MEMORANDUM FOR: Edward Jordan, Director Office for Analysis and Evaluation of 1 Operational Data FROM: Thomas E. Murley, Director Office of Nuclear Reactor Regulation NRR PLANS FOR APPROVAL OF GE TOPICAL REPORTS NE0C-30844,  ;

SUBJECT:

"BWR OWNERS' GROUP RESPONSE TO NRC GENERIC LETTER 83-28" AND NEDC-30851P, " TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS FOR BWR RPS"

'By letter dated January 31, 1985, the BWR Owners' Group submitted NEDC-30844, "BWR Owners' Group Response to NRC Ganeric Letter 83-28," in response to Item 4.5.3 of NRR Generic Letter 83-28. NEDC-30844 analysis of the BWR reactor protection to system address (RPS) provides a gene reliability concerns identified in response to the Salem ATWS event. Subsequently, by letter dated March 31, 1985, the Owners' Group submitted NEDC-30851P, a generic analysis that supports extending the surveillance intervals and allowable outage times for the RPS. NEDC-30851P extends the methodology developed for the NEDC-30844 report to support the noted Technical Specification changes.

Because both reports are based on the same methodology, the staff has prepared a single Safety Evaluation Report addressing its review of them. The staff evaluation was perfonned with the assistance of INEL under a technical assistance contract. Attachment 1 is a draft reply to the Owners' Group providing the conclusions of the staff review. The SER enclosed with this reply provides guidance on acceptable alternatives to existing STS (i.e.,

requirements. The SER re technical specification) presents and is subject relaxation to CRGR review. Theof a current proposed change staff position is an extension from monthly to quarterly of the surveillance interval for RPS l channel functional testing, and an extension from I hour to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the allowable equipment out-of-service time for RPS instrumentation. Attachment 2 is the CRGR review package that addresses the responses to the information requirements included in the CRGR Charter.

j The staff position proposed herein has been coordinated with AE0D and RES.

OGC has reviewed this package and has no legal objection.

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I't g c a Edward Jordan - 2.-

It is. requested that a CRGR review of this proposal be scheduled at the earliest opportunity.

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Thomas E. Murley, Director Office of Nuclear Reactor Regulation

Enclosures:

1 As stated cc w/ enclosures:

V. Stello J. Grace A. Davis R. Martin J. Martin DISTRIBUTION Central Files SRXB R/F T. Murley-J. Sniezek R. Starostecki L. Shao A. Thadani' N. W. Hodges T. Collins T. Collins R/F-

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OFFICIAL RECORD COPY

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ATTAClil1ENT 1

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Mr. Terry A. Pickens, Chaiman BWR Owners' Group i;

Northern States Power Corporation 414 Nicollet Mall l Minneapolis, Minnesota 55401

Dear Mr. Pickens:

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SUBJECT:

GENERAL ELECTRIC COMPANY (GE) TOPICAL REPORTS NEDC-30844, "BWR OWNERS' GROUP RESPONSE TO NRC GENERIC LETTER 83-28,"

AND NEDC-30851P, " TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS FOR BWR RPS" We have completed our review of the subject topical reports submitted by the BWR Owners' Group by letters dated January 31, 1985 and May 31, 1985.

Enclosure 1 provides our Safety Evaluation Report (SER). Our evaluation applies only to the matters described in the report.

We find that NEDC-30844 provides an acceptable basis for resolving Item 4.5.3 of Generic Letter 83-28, subject to confinnation of the plant-specific applicability of the generic analysis. Therefore, upon receipt of a letter with this confirmation from each licensee or applicant Item 4.5.3 of Generic Letter 83-28 will be closed out for that licensee or applicant.

Based on our review of NEDC-30851P, we find that it provides an acceptable generic basis for supporting plant-specific Technical Specification changes related to the reactor protection system (RPS) for plants using a relay RPS, subject to the conditions noted herein. We also have reviewed the topical report regarding Technical Specification (TS) changes on the solid-state RPS.

We needed and obtained additional information for our review. The results of our review of the solid-state RPS reliability analyses will be presented in a future staff SER.

As noted in the enclosed SER, applicants for proposed Technical Specification changes for individual plants must:

1. Confim the applicability of the generic analyses for NEDC-30851P to its plant.

2. Demonstrate by use of current drift infonnation provided by the equipment vendor or plant-specific data, that the drift characteristics for instrumentation used in the RPS channels in the plant are bounded by the assumptions used in NEDC-30851P when the functional test interval is extended from monthly to quarterly.

I Mr. Terry A. Pickens 3. Confirm that the differences between the RPS in the plant and the PPS of the generic analysis plant were included in the plant-specific analysis using the procedures of Appendix K of NEDC-30851P, or provide plant-specific analyses to demonstrate that there is no appreciable change in RPS availability or public risk.

Enclosure 2 provides an acceptable fonnat for proposed TS changes based on NEDC-30851P. Our review of plant-specific changes will consider only the applicability of the topical report to the specific plant. Because the proposed changes are voluntary and subject to the confinnatory actions noted above, we do not intend to issue a revision to the Standard Technical Specifications (STS) for BWR plants at this time. Eventually the STS will be revised to incorporate these changes as part of the NRC program for TS improvements.

In accordance with procedures established in NUREG-0390, " Topical Reports Review Status," we request that the GE/BWR Owners' Group publish accepted versions of NEDC-30851P, both proprietary and non-proprietary, within 3 months of receipt of this letter. The accepted versions should (1) incorporate this letter and the enclosed Safety Evaluation Report between the title page and the abstract and (2) include an -A (designated accepted) following the report identification symbol.

Should our acceptance criteria or regulations change so that our conclusions as to the acceptability of the reports are no longer valid, the BWR Owners' Group and/or the applicants referencing these topical reports will be expected to revise and resubmit their respective documentation, or submit justification for the continued applicability of the topical reports without revision of their respective documentation.

Sincerely, Ashok C. Thadani, Assistant Director for Systems Division of Engineering & Systems Technology Office of Nuclear Reactor Regulation

Enclosures:

1. Safety Evaluation Report

2. Fonnat for Technical Specification Changes

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ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF BWR OWNERS GROUP REPORTS PEDC-30844 AND 30851P ON JUSTIFICATION FOR AND EXTENTION OF ON-LINE TEST INTERVALS AND ALLOWABLE OUT-OF-SERVICE TIMES FOR BWR REACTOR PROTECTION SYSTEMS

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1.0

SUMMARY

The staff has reviewed the General Electric Company (GE) Topical Reports NEDC-30844, "BWR Owners' Group Response to NRC Generic Letter 83-28," and NEDC-30851P, " Technical Specifications Improvement Analysis for BWR RPS."

These reports were issued by the BWR Owners' Group to respond to Generic Letter 83-28, Item 4.5.3 and to support the proposed extension of reactor protection system (RPS) on-line test intervals and allowable out-of-service times (A0Ts) for RPS test and repair. The staff has concluded that the analyses presented 4

in the Owners' Group reports are acceptable for resolving these issues, subject to the limitations and conditions presented herein.

2.0 BACKGROUND

On February 25, 1983, both scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon receipt of an automatic RPS signal. The operator terminated the incident about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers in this incident ,

was related to sticking of the under-voltage trip atb 3 ment. Earlier on February 22, 1983, at the same unit, a steam generator low-low level during l

In this case, the reactor l plant startup resulted in an automatic trip signal.

was tripped manually by the operator almost coincidentally with the automatic trip. 1 l

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4 On February 28, 1983, the NRC Executive Director for Operations directed the staff to investigate and report on the generic implications of these occurrences. The results of the staff's assessment are in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant" (Ref.1).

Subsequently, the staff issued Generic letter 83-28 (Ref. 2) reouesting that j all licensees of operating reactors, applicants for an operating license, and j holders of construction permits respond to generic issues raised by the analyses of these two anticipated transient without scram (ATWS) events. Item 4.5.3 of this generic letter requested that licensees and applicants review the existing RPS on-line functional test intervals required by technical specifications (TS). They are to ensure that current and proposed intervals j for such testing are consistent with a goal of achieving high RPS availability l considering uncertainties in component failure rates, uncertainties in common f mode failure rates, reduced redundancy during tMting, operator errors during testing, and component wear caused by the testing.

3.0 APPROACH The BWR Owners' Group decided to attempt to resolve these issues generically.

It connissioned GE to perform generic analyses and apply the generic results to the individual boiling water reactor (BWR) plants. (The generic analyses are applicable to a vast majority of plants that have a relay RPS as well as to the rest of the plants that have solid-state RPS.)

Two GE topical reports were issued: (1) NEDC-30844 (Ref. 3), which analyzed a representative BWR plant and provided a technical basis for ensuring that the current RPS on-line test intervals meet the recommendations of Generic Letter 83-28, Item 4.5.3, and (2) NEDC-30851P (Ref. 4), which used the base case results from NEDC-30844 to establish a basis for extending the current RPS on-line test intervals and A0Ts. These reports used reliability analyses with fault tree modelling to estimate RPS failure frequency. Sensitivity analyses 1 l

were used to vary the factors that represented the five areas of concern l delineated in Item 4.5.3 so that their impact was considered appropriately. f The acceptance guideline used by GE for the TS changes is based on net increase l l

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in risk, which is the difference between the increase in risk that would result

.from the TS changes and the decrease in risk that would result from the reduced likelihood of inadvertent scrams or exceeding the Limiting Conditions of Operations. If the net change in risk is detemined to be insignificant, the TS changes would be acceptable. To apply generic plant analyses to specific plants, GE collected necessary information on the RPS for each BWR, detemined the differences for each plant, and analyzed the effect of each identified difference on the RPS failure frequency (this procedure is described in

-- Appendix K to NEDC-30851P).

4.0 NRC ACTION

- The staff engaged the services of Idaho Nuclear Engineering Laboratory (INEL) te review the data and met'hodology used in the two GE reports. This review was to determine the validity of using the RPS failure frequency as a risk measure; to assess the adequacy of the fault tree analyses and the supporting data; and to detemine reliability calculations adequacy using the WAMCUT (Ref. 5) and FRANTIC III (Ref. 6) computer codes. INEL notes that, based on conservative assumptions, the estimated increase in RPS unavailability due to the proposed TS changes would contribute a very small increase to estimated core-melt frequency. However, if the benefits due to the TS changes are taken into account such as the reduction in the number of inadvertent test-induced scrams, the net change in risk resulting from the TS changes would be considered insignificant. Using this reasoning, the staff agrees with INEL on the basis for acceptance. INEL issued a technical evaluation report (TER) (Ref. 7) \

presenting the details and results of its review. I J

On the basis of its review of the TER, the staff endorses the conclusion that the methodology used and results obtained in the two GE reports were verified The staff, for the relay RPS, but were not verified for the solid-state RPS.

therefore, finds the analyses presented for the relay RPS in NEDC-30844 and NEDC-30851P acceptable to support a determination that the current on-line RPS test intervals are consistent with the high RPS availability required by The staff also finds the use of the analyses Generic Letter 83-28. Item 4.5.3.

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acceptable for supporting the proposed extensions to TS test intervals and increases in A0Ts.

However, these findings are limited to plants using GE BWR relay RPSs. The staff review of additional infonnation and analysis must be completed before a position on the solid-state RPS can be provided. The staff's finding are discussed in the following sections.

5.0 SPECIFIC COMMENTS ON THE TOPICAL REPORTS AND THEIR RESULTS 5.1 Methodology and Data The GE analysis evaluates the RPS TS using the RPS failure frequency as the risk measure. In essence, the GE analysis evaluates the changes in the RPS failure frequency as a result of the changes in the RPS TS (for example, the changes in the RPS functional testing intervals and in the A0Ts for repairing or testing RPS channels). A brief discussion of the methodology used to calculate the RPS failure frequency follows. j The calculations of RPS failure frequency depend on two sets of parameters.

The first set consists of initiating events that eventually lead to actuation of the RPS. The second set consists of "RPS unavailabilities," which are the probabilities that the RPS is unavailable given the demands for RPS actuation.

Depending on each initiating event, the number of sensors that could actuate the RPS would vary. Therefore, the RPS unavailability for one initiating event may differ from that for another.

For each initiating event, GE developed a fault tree to quantify PPS unavailability per demand. The fault tree models the logical relationship of the faults that may contribute to RPS unavailability. The logical representation of the fault tree was used with the computer code WAMCUT to i

obtain the dominant cutsets which are the combinations of faults that cause the RPS to be unavailable. The dominant cutsets, together with information on

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i testing and repairing the RPS, are then used with the computer code FRANTIC III, which calculates the unavailability of RPS per demand.

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l To detemine the adequacy of fault tree analyses for estimating RPS unavailabilities, INEL used the computer code COMCAN III (Ref. 8) to generate cutsets. INEL then compared these cutsets with the cutsets generated by WAMCUT, which GE used. In general, INEL detemined that the cutsets obtained In addition, l by GE are adequate for estimating the relay RPS unavailabilities.

- INEL determined the overall data used in the GE relay RPS analyses are valid, f

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The initiating event frequencies used in the GE analyses are mostly in agreement with those from other reputable data sources (Ref. 9,10,11). INEL found that a higher estimate for pressure regulator failure frequency should be used. However, the use of this higher value would result in only a very small increase in overa'll RPS failure frequency and is not considered significant.

With respect to component failure rates used in the GE relay RPS analysis, INEL detemined that the failure rates are acceptable for estimating RPS unavailabilities. However, INEL could not verify the random failure rates for the solid-state RPS the basis of the infomation and reference in the GE reports. In addition, INEL found that GE used a simplified fault tree model for the solid-state RPS to do WAMCUT computer runs instead of the fault tree provided in Appendix I of the NEDC-30851P report. GE did not provide any l

documentation for simplifying the fault tree. The staff agrees with the INEL findings. On this basis, the staff asked and obtained additional infomation regarding the solid-state RPS reliability analyses from GE to complete our {

review.

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5.2 Uncertainties in Component Failure Rates GE has performed sensitivity analyses to determine the sensitivity of RPS To do this, GE unavailability to di. certainties in component failure rates.

multiplied the component failure rates with an error factor, which is the ratio i of the upper uncertainty bound and the median value. The results indicate that f uncertainties in the component failure rates have a negligible impact on RPS l

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unavailability. Therefore, GE concluded that the RPS unavailability is not sensitive to the uncertainties in component failure rates.

l To verify GE's conclusion, INEL perfomed a sensitivity run on the main steam isolation valve (MSIV) closure event. INEL detemined that RPS unavailability was about the same as the RPS unavailability for the base case value.

Therefore, the staff concludes that uncertainties in component failure rates do i not significantly affect RPS unavailability and, hence, the failure frequency.

5.3 Comon Cause Failure Rates GE detemined that the comon cause failure rates of the scram contactors do contribute significantly to the RPS unavailability for each of the initiating events. GE contends that even when these conson cause failure rates are considered, the results are still lower than other published results.

The staff notes that the proposed changes to the technical specifications would increase to weekly the frequency of testing of the scram contactors by means of the manual scram / test switches in the control room. The staff believes that this -is an efficient way of detecting the comon cause failures of scram contactors.

5.4 Component Wear Out Caused by Testing The GE analysis indicates that among the components in the RPS, the scram contactors are de-energized whenever an individual sensor and its associated relay are tested. Because 11 different types of sensors are tested while the reactor is at full power, the scram contactors would be challenged more often than other components in the RPS. For this reason, GE examined the effect of scram contactor wear out caused by testing. The GE analysis indicates that scram contactor wear created by the number of tests required by current TS does not cause any significant increase in RPS failure frequency.

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7 To verify the GE analysis, INEL performed a sensitivity run using the computer code FRANTIC III. The data input to the code included a factor that increased the failure rate after each test. The INEL results correspond to the GE results and indicate that the RPS unavailability is not significantly impacted by scram contactor wear created by testing.

l l According to GE, no scram contactors have failed to date, and there is no indication that wear out is becoming a potential problem. With RPS sensor

- channel testing reduced from monthly to quarterly, the frequency of RPS component actuation would be reduced and RPS components would be less likely to fail due to wear.

Therefore, the staff concludes that RPS unavailability is not significantly

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impacted by wear of component due to the proposed test interval.

5.5 Sensitivity to Reduced System Redundancy During Testing GE examined the impact of reduced system redundancy during testing on the RPS unavailabilities by comparing two cases. In the first case, a sensor channel is "jumpered" during a test and is unable to provide an RPS signal upon actual demand. In the second case, a sensor channel is placed in trip during a test and thus does provide an RPS signal.

The RPS unavailability for the first case is higher than that for the second case. However, the difference between these two RPS unavailabilities was found to be small and indicates that reduced redundancy during testing has no significant impact on RPS unavailability. As an audit of the GE analyses., INEL performed a sensitivity analysis that showed that a sensor can be unavailable for 3 months without having a significant impact on RPS unavailability.

Therefore, the staff concludes that reduced system redundancy as a result of testing does not have a significant effect on RPS unavailabilities.

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8 5.6 Sensitivity to Human Error Rates During Testing The GE analysis considered two types of human errors during a test: an operator disabling components randomly during c test and an operator' causing a comon cause failure of all similar components during a test. The GE analysis determined that the fit st type of operator error does not have significant impact on RPS failure frequency, but the second type of operator error does.

The GE analysis determined that operator error disabling all scram contactors was the biggest contributor to the RPS failure frequency. Under the proposed TS requirements, the operators will perfonn channel functional testing of the r.anual scram on a weekly basis by actuating the channel manual scram / test switches in the control room. Therefore the staff believes the* the weekly testing of manual scram is an efficient way of detecting the common cause failures of scram contactors due to operator errors.

5.7 Changes in RPS Surveillance Testing Intervals GE calculated RPS failure frequency by varying the surveillance testing intervals of the average power range monitor (APRM) and other sensors from monthly to quarterly. The results showed a small change in RPS failure frequency.

To verify the GE analyses, INEL used the MSIV closure event and varied the testing intervals as above; the INEL results were the same as the GE results.

In view of the small change in RPS failure frequency as a result of the change in test intervals, the staff concludes that the proposed changes in testing intervals are acceptable.

5.8 Changes in Allowable Outage Times for Test and Repair i The current TS for the relay RPS allow A0Ts of I hour for repairing and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for testing a single sensor channel without placing the channel in a tripped l

9 state. GE believes that the short times allowed by the current TS may cause operator error as a result of stress during repair and testing. GE also conter.ds that placing an individual channel in a tripped condition when repairs i and tests cannot be completed within the allowable outage time may increase the l likelihood of an inadvertent scram. Therefore, GE proposes to extend the A0Ts for repair and for testing. The proposed A0Ts, which are based on the average times needed to complete tests and repairs, include sufficient time margins so l that the operators would not be placed under undue stress. To support these

- above changes, GE perfonned sensitivity analyses and concluded that changing of the A0Ts had a negligible impact on RPS failure frequency.

INEL audited these GE calculations, using FRANTIC III, to analyze the MSIV closure event to account for the extended test and repair times. The INEL results verify GE's findings. On this basis, the staff concludes that the proposed A0Ts for repair and test times of I hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for individual relay RPS sensor channels can be extended to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> respectively.

5.9 Plant-Specific Application of Generic Results The GE analyses considered the impact on RPS unavailability resulting from the differences in systems / components among the various BWR plants from BWR/3 to BWR/6. GE detennined that if the proposed RPS TS changes are implemented, there would be no significant increase of RPS failure frequency for the reviewed BWR plants that use the relay RPS. This detennination is based on use ,

l of the GE procedure given in Appendix K of NEDC-30851P for evaluating specific plants against the generic RPS design and analyses.

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In general, the staff finds the GE procedure valid for making such plant-specific comparisons. However, the staff cannot verify that the generic results apply to a specific plant without performing (1) detailed comparisons of plant-specific designs with the generic design and (2) sensitivity analyses.

The staff also notes that there is diversity among the analog trip units used in BWRs. The GE reports do not confirm that calibration of the analog trip 1

4 10 units can be extended from monthly to quarterly without creating excessive drift.

In addition, the reports do not supply drift information for other instrumentation used in the RPS. The staff has given a list of. plant-specific conditions licensees must meet to close out Itemlisted These conditions, 4.5.3in of Generic L and to make the plant-specific RPS TS changes.

Table 1 include a demonstration of adequate drift characteristics.

As for applying the generic results of the solid-state RPS analyses, the staff

- asked for and obtained additional information so that it could complete the assessment of the solid-state RPS reliability analyses.

6.0 CONCLUSION

Based on the INEL findings, the staff concludes that the GE analyses demonstrate general compliance with Item 4.5.3 of Generic Letter 83-28 for the In addition, the staff finds that the proposed facilities using the relay RPS.

changes in the TS for the relay RPS are generally acceptable. With respect to plant-specific implementation of changes in the RPS TS for a plant with a RPS, Table 1 lists plant-specific conditions that each licensee or applicant must meet to complete the resolution of Item 4.5.3 and niake any proposed TS Further, the results of solid 7 state RPS review will changes fully acceptable.

be provided in a future staff Safety Evaluation Report.

7.0 REFERENCES

1. US Nuclear Regulatory Comission, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant, Vols. I ard 2, NUREG-1000, April 1983.

2. Eisenhut, 6. G., NRC letter to All Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Pennits,

" Required Actions Based on Generic Implications of Salem ATWS Events" (Generic letter 83-28), July 8, 1983.

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- 3. S. Visweswaren et al., "BWR Owners' Group Response to NRC Generic Letter 83-28, Item 4.5.3," General Electric Company, NEDC-30844, January 1985.

.. 4 W. P. Sullivan et al., " Technical Specification Improvement Analyses for

. BWR Reactor Protection System," General Electric Company, NEDC-30851P, May 1985.

5. R. C Erdmann, F. L. Leverenz, and H. Kirch, "WAMCUT; A Computer Code

- -for Fault Tree Evaluations," EPRI NP-807, Science Applications, Inc.,

June 1978.

6. T. Ginzburg et al., " FRANTIC III; A Computer Code for Time-Dependent

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Reliability Analysis," Brookhaven National Laboratory and Science Applications, Inc., April 1984.

7. B. Collins et al., "A Review of the BWR Owners' Group Technical Specification Improvement Analyses for the BWR Reactor Protection System. " EGG-EA-7105, corporate publisher January 1986.

B. D. M. Rasmuson et al., "CONCAN III; Use of COMCAN III in System Design and Reliability Analysis, " EGG-2187 corporate publisher, October 1982.

9. A. S. McClymont et al., "ATWS: A Reappraisal, Part 3: Frequency of Anticipated, Transients," EPRI NP-2230, Science Applications, Inc.,

January 1982.

10. M. McCann et al., "Probabilistic Safety Analysis Procedures Guide,"

NUREG/CR-2815, Vol. 1 Rev. 1, August 1985.

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11. D. P. Mackowiak et al., " Development of Transient Initiating Event Frequencies For Use in Probabilistic Risk Assessment," NUREG/CR-3862 (Draft), June 1904.

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CONDITIONS TO CLOSE OUT RELAY PLANTS TABLE 1

.For plant-specific application of the TS changes proposed, and fer plant-specific closecut of Item 4.5.3 of Generic Letter 83-28, an individual

- licensee for a plant using a relay RPS must:

(1) Confirm the appl.icability of the generic analyses to its plant.

(2) Demonstrate, by use of current drift infortnation provided by the equipment vendor or plant-specific data, that the drift cha acteristics for instrumentation used in RPS channels in the plant are bcunded by the assumption used i:. NEDC-30851P when the functional test interval is extended from monthly to quarterly.

(3) Confirm that the differences between the parts of the RPS that perform the trip functions in the plant and those of the base case plant were included in the analysis for its plant done using the procedures of Appendix K of NEDC-30851P (and the results presented in Enclosure 1 to letter OG5-491-12 from L. Rash (GE) to T. Collins (NRC) dated November 25,1985), or present plant-specific analyses to demonstrate no appreciable change in RPS availability or public risk.

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V h' ENCLOSURE 2 CEANGES TO RI1AY R.PS TECHNICAL SPECIFICATION 3/4.3 INSTRUMENTATION'

~3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 1.1K1 TING CONDITION FOR OPERATION, 3.3.1 As a minisus, the reactor protection sytem instrumentation channels

.. shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM R.ESPONSE TIME as .shwn in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Systes requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition

• within twelve hours. The provisions of Specification 3.0.4 are not applicable.

b. With the number of OPERABLE channels less than ' required by the Minimum OPERABLE Channels per Trip Systes requirement for both trip systems, place at least one trip system ** in the tripped condition vitain nue hour and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the perforance of the CHANNEL CHICK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channela shall be performed at least on~ce per 18 months.

'An inopersble channel need not be placed in the tripped condition whkre'this i would cauise the Trip Function te occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the ACTION required by Table 3.31-1 for that Trip Function shall be taken.

• • 1f more channels are inoperable in one trip system than in the other, place  ;

the trip system with more inoperable channels in the tripped condition, ascept when this would cause the Trip Function to occur. .

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CHANGES TO RI1AY RPS TECHNICAL SPECIFICATION 1

NOTES TO TABLE (s) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IM and SM channels shall be determined to overlap for at least (1/2) decades during asch startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlsp for at least '

. (1/2) decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a hast balance during OPIAATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWIR. Adjust the APRM channel if the absoTute difference is greater than 21 of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall act be included in determining the absolute difference.

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (ETPH) using the 71P system.

(g) Calibrate trip unit at least once per 92 days. ,

(h) Verify acasured core flow to be greater than or equal to established core flow at the existing flow control valve position.

l (i) This calibration shall consist of (verifying) (adjustment, as required  !

of) the 6+1 second simulated thermal power time constant.

(j) This function is not required to be OPIRABLE when the resctor pressure vessel head is removed per Specification 3.10.1. j l

(k) With any control rod withdrawn. Not applicable to control rode rasaved f per Specification 3.9.10.1 or 3.9.10.2.

l

---

m_____m_-___ __ m.__ _ __ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ _ _

3/4.3 IKSTpWI m T10W BL5!5 3/4.3.1 tract 0e PROTECT 10N SY$itw INSTRWImTION The reacter protection systet autentica11y initiates a reactor scrae to:

a. Freserve the integrity of the fuel $1 adding. J

b. Preserve the integrity of tne reactor coolant systee.

s. Mintaire the energy which must be absorbed following a loss.cf-coolant accident, and

d. Prevent inadvertent teltica11V.

This specification provides the limiting conditions for operation i etcessary to preserve the ability of the system to perfem its intended I fcettien even during periods when instivment thannels my be out of service j because of nintenance. Ifhen Meessary, one thannel may be made (noperable 1 for brief intarvals to conduct required surveillance.

The reactor protection system is made vp of two independent trip systees. There are usually fourThe channels to een14cr each parameter with two outputs of the channels in a trip system channels in each trip system.

are ccebined in a logic so that either channel will trip that tripThe system.

system The tripping of both trip systees will produce a reactor scram.

sseets the intent of 1EEE479 for nuclear power plant protection systems.

Specified surveillance intervals and surveillance and wintenance cutage tirnes have been determined in accordance with NEDC-30251 P ' Technical Specification Improvement Analyses for BWR Reactor Protection System.' as i approved by the NRC and documented inThe the $ER bases(letter to T.

for the tripA.setting Pickens of from dated .

the RPS are discussed in the bases for Specification 2.2.1.

The sensuremet of response tire at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time Ifrit assured in the safety onelysts. We stadit I i

was taken for these channels erith response tiets indicated as met applicable.

Pesponse time eay 64 demonstrated by any settes of seguential. overlapping er total channel test measurement Sensor provided such tests response timedemennrate verification nythebe total thannel response time 45 defined.

demonstrated by either (1) inplace, ensite or effsitt test sensurements, or (2) utilising replacement sensors with certified response times.

e d

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.. .'.? i i

ATTACHMENT 2 l RESPONSE TO REQUIREMENTS FOR CONTENT OF PACKAGE I SUBMITTED FOR CRGR REVIEW l

(i) The proposed generic requirement or staff position as it is proposed l to be sent out to licensees.

Staff Position is:

• Surveillance Testing Intervals for BWR RPS instrumentation may be extended from monthly to quarterly.

• Allowable repair and test times of I hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for relay RPS sensor channels may be extended to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> respectively-(ii) Draft staff papers or other underlying staff documents supporting the requirements or staff positions.

• SER on NEDC-30851P, " Technical Specification Improvement Analysis for BWR RPS Instrumentation."

• INEL Technical Evaluation Report on NEDC-30851P.

• NUREG-1024, " Technical Specifications - Enhancing the Safety Impact."

" 'SECY 86-10. " Recommendations fur Improving Technical Specifications."

(TSIP and AIF reports).

• SECY 86-310, " Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors."

• The staff acceptance of Licensing Topical Report WCAP-10271, " Evaluation of Surveillance Frequencies and Out-of-Service Times for the Reactor Protection Instrumentation Systems."

(iii) Each proposed requirement or staff position shall contain the sponsoring office's position as to whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce existing requirements or staff positions.

u , .- .

2 The RPS channel functional test surveillance interval can be extended from monthly to quarterly for relay plants, and the allowable outage time for relay RPS sensor channels may be extended from I hour to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This is a relaxation of a staff position.

(iv) The proposed method of implementation with the concurrence (and any coments) of OELD on the method proposed.

0GC concurred in this package. 0GC concurrence was subject to the resolution of certain comments which have been incorporated in this package.

(v) Regulatory analyses generally conforming to the directives and guidance of NUREG/BR-0058 and NUREG/CR-3568.

The topical report indicated a negligible change in core melt frequency as a consequence of changing the surveillance intervals and allowed outage times. The benefits include a decrease in manpower needed for surveillance testing, a decrease in manpower exposure, and the potential for a slight decrease in inadvertent scrams and challenges to safety systems.

(vi) Identification of the category of reactor punts to which the generic requirement or staff position is to apply.

The staff position is applicable to all Boiling Water Reactors designed by General Electric Company on a voluntary basis.

(vii) For each such category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The evaluation shall document for consideration information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed action:

(a) statement of the specific objectives that the proposed action is i designed to achieve; The objective is to advise licensees of the acceptability of modifying current Technical Specification requirements to extend RPS surveillance intervals and equipment out-of-service times in accordance with the BWR Owners' Group analysis. Licensee

{

response is voluntary and must be considered by the staff as an operational enhancement. Such review priority is low relative to safety related issues, i

.. e.e ,,

3 (b) . General' description of the activity.that would be' required by the

. licensee or applicant in order to complete the action;

-Licensees' choosing to adopt the new surveillance test intervals-must submit a plant-specific Technical Specification request which modifies the surveillance test intervals and allowed outage times. Surveillance. procedures would need to be modified

-in the same way.

(c) Potential change in the risk to t' he public from the accidental offsite release of radioactive material; NEDC-30851P results indicate negligible change in overall public risk from this action. The staff SER concurs with this assessment.

(d) Potential impact on radiological exposure of facility employees and other onsite workers; Reduction in surveillance requirements should slightly reduce occupational exposure.

(e) Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction delay; Not Applicable.

(f) The potential safety impact of changes in plant or operational y complexity, including the relationship to proposed and existing regulatory requirements and staff positions; Reduction in surveillance requirements will minimize human errors and associated challenges to scram systems. Based'on the results of the analysis, plant safety is not expected to be adversely impacted by this action.

(g) The estimated resource burden on the NRC associated with the proposed action and the availability of such resources; If the generic analysis in the GE reports can be readily applied to the plant-specific submittals, the NRC resource burden should be minimal. A license amendment request would be in response to l' matters for which an acceptable format for the TS changes has been established by the Enclosure in the SER. The processing of the license amendments would be carried out by NRR staff and should require approximately one man week of NRC technical specialist staff resources.

4 (h) The potential impact of differences in facility' type, design or age on the relevancy and practicality of the proposed action; Not applicable since the staff position is that implementation is voluntary.

(1) Whether the proposed action is interim or final, and if interim, the justification for imposing the proposed action on an interim basis.

This is the final staff position.

(viii) For each evaluation conducted pursuant to 10 CFR 50.109, the proposing-Office Director's determination, together with the rationale for the determination based on the consideration of paragraph (i) through (vii)above,that:

(a) There is a substantial increase in the overall protection of public health and safety or the common defense and security to be derived from the proposal; and (b) The direct and indirect rosts of implementation, for the facilities affected, are justified in view of this increased protection.

Not applicable.

(1x) For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing Office Director's determination, together with the rationale for the determination based on the considerations of paragraph (1) through (vii)above,that:

The public health and safety and the common defense and security j (a) would be adequately protected if the proposed reduction in requirements or positions were implemented, and (b) The cost savings attributed to the action would be substantial enough to justify taking the action.

The Safety Evaluation Report provides the rationale to relax the existing requirements. Reduction in surveillance requirements will minimize human $

errors and challenges to scram systems. heither a reduction in plant safety nor an increase in risk is expected as a result of this action. Therefore,  !

the reduction in manpower requirements for surveillance and the potential for decreased safety systems challenges warrants the action. Furthermore, licensing topical report acceptance will be cost effective to our resources.

Routine licensing amendments with the established TS changes format will be processed by NRR Staff and should require ro mere than 1 man week of technical specialist staff resources.

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