ML20082V282

Safety Evaluation Supporting Amend 49 to License NPF-42

ML20082V282
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Site:	Wolf Creek
Issue date:	09/12/1991
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9109230191
Download: ML20082V282 (5)

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UNITED STA1ES

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATFD TO AMENDMENT NO. 4910 FACILITY OPERATING LICENSE _NO. NPF.et2_

WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482

1. 0 INTRODUCTION By letter dated April 22, 1988 (Ref. 1), the Westinghouse Owners' Group (WOG) submitted topical report WCAP-11736 entitled " Residual P.at Removal System Autoclosure Interlock Removal Report for the Westinghouse Owners' Group," for NRC review.

WCAP-11736 documents the analyses performed to justify deletion of the autoclosure interlock (ACl) on the Residual Heat Removal System (RHRS) suction / isolation valves at four reference plants:

Salem Unit 1 Callaway Unit 1, North Anna Unit 1, and Shearon Harris Unit 1.

The reference plants represent the lead plant in each of four groups into w'iich WOG participating plants were cate.gorized based on similarity of RHRS configuration and design characteristics.

The proposed ACI deletion addresses NRC concerns regarding potential failure of ACI circuitry resulting in isolation of the RHRS with attendar.t loss of decay heat removal capability during cold shutdown and refueling.

A Safety Evaluation Report (SER) documenting the NRC review of WCAP-11736 was issued on August 8, 1989 (Ref. 2).

The SER concluded that a net safety benefit would result from removal of the RHRS ACI provided that five plant improvements delineated in the SER are implemented.

In addition, the SER concluded that the information contained in WCAP-11736 may be referenced to supplement licensees' plant-specific submittals requesting removal of the RHRS ACI.

However, such reference only would be used to show compliance with those items that are generic to the WOG plants.

A plant-specific submittal would be I

required of each licensee seeking approval to remove the RHR ACI.

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The above ref erenced plant improvements are listed below:

1 (1) An alarm will be added to each RHR suction valve which will actuat.e if l

the valve is open~and the pressure is greater than the open permissive setpoint and less than the RHR system design pressure minus the RHR pump l

head pressure.

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(2) Valve position indication to the alarm must be provided from stem-mounted limit switches and power to these switches must not be affected by power lockout of the valve.

(3) The procedural improvements described in WCAp-11736 should be implemented.

(Procedures themselves are plant-specific).

(4) Where feasible, power should be removed from the RHR suction valves prior to their being leak-checked (plant-specific).

(S) The RHR suction valve operators should be sized so that the valves cannot be opened against full system pressure (plant-specific).

2.0 EVALUATION By letter dated March 5, 1991, Wolf Creek Nuclear Operating Corporation, the licensee for the Wolf Creek Generating Station (WCGS), submitted an applica-tion to revise Technical Specifications (TS) 4.5.2.d.1, 4.4.9.3.2 (a and b),

and the associated Bases (Ref. 3).

Supplementary information was provided by letter dated July 24, 1991 (Ref. 4).

These TS revisions have been proposed in support of the licensee's plans to remove the RHRS ACI during their 1991 refueling outage.

The proposed revision to TS 4.4.9.3.2 deletes the surveil-lance requirement to verify once every 31 days that one of the two in-series suction valves in each RHRS train is in the open position with its power removed, and to verify once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the second suction valve in each train is open.

This is replaced by the requirement to verify once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that both suction valves in each train are opea.

The IS applies when the RHR suction relief valves are being used for cold overpressure protection.

Regarding TS 4.5.2.d.1, the proposed revision deletes the surveillance require-ment for verifying ACI operability (the open permissive interlock surveillance remains unchanged).

The July 24, 1991, submittal provided additional clarifying information and did not change the initial no significant hazards consideration determination.

As noted above, the NRC approved report WCAP-11736 provides the underlying basis for justifying the licensee's planned action.

The WCAP-11736 reference plant for WCGS is Callaway Unit 1.

The licensee's submittals (Ref. 3 and 4) include a plant-specific analysis of the planned ACI deletion as a supplement to WCAP 11736.

Both WCGS and Callaway are Standardized Nuclear Unit Power Plant Systems (SNUPPS) plants.

As such, the WCGS RHRS design is essentially identical to that of Callaway.

The licensee, after examination of WCAP-11736, concluded on this basis that the results of the Callaway evaluation are directly applicable to WCGS.

In addition, the licensee has addressed each of the five plant improvements set forth in Reference 2 and iisted above.

Where deviations from these improvements are proposed by the licensee, analyses are presented to demonstrate that equivalent levels of safety exist.

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'1.. With regard to the above mentioned five plant improvements, the licensee's March 5, 1991 and July 24, 1991, submittals provided the following responses:

Improvement 1 - the licensee plans to modify the alarm cir:uit to provide a control room alarm which will actuate if any of the four RHRS suction /

isolation valves

' not fully closed and RCS pressure exceeds the alarm setpoint.

The re

. int (360 psig) is consistent with the WCAP-11736 guidance.

Also, in accordance with WCAP-11736, the open permissive interlock (OD1) for each RHPS suction valve will remain intact and unchanged.

Improvement 2 - the licensee proposes to use existing limit switches located in the valve motor operator rather than installing stem-mounted limit switches for valve position indication to the new alarms, lhese contacts are aif ferent from the limit switch contacts which presently provide valve position to the main control board.

As a result, some diversity in valve position indication is achieved.

In addit'on, the alarm circuit will be powered by a supply which is separate from the supply that powers the valve control and position indication circuits.

This ensures that the alarms remain operational during valve power lockout.

Improvement 3 - the licensee has reviewed the WCGS plant procedures to identify any necessary revisions resulting from the proposed modifications.

Included in this review were WCGS procedures corresponding to the general procedures listed in Section 9.0 of WCAP-11736 If during startup, an open kHR suction valve cannot be closed upon receipt of an alarm, the alarm response procedure will di-ect operators to halt pressurization and return the plant to a safe shutdown condition.

Additionally, to further ensure alarm operability, the appropriate surveillance testing procedures will be revised.

Improvement 4 - the licensee does not plan to remove power from the RHR5 suction / isolation valves prior to leak testing.

The original intent of this recommended improvement was to ensure that the valves remained in the tested configuration during testing.

Leak testing of the RHRS surtion/ isolation valves at WCGS is normally performed in Mode 4 or 5.

Closure and power removal from these valves is required only prior to entering Mode 3.

One advantage of performing leak testing prior to entry into Mode 3 is that the amount of cooldown required to perform valve maintenance in the event the valve exhibits greater than allowable leakage is minimized.

The TS require that at least one RHR train be in service at all times during Modes 4 and 5.

The two RHR trains are, therefore, leak tested consecutively.

When testing is completed on one train, thet train is returned to service.

Removal of valve power prior to testing in Modes 4 and 5 (and subsequent restoration of power) would increase procedural complexity and time without offering a safety benefit.

Increased testing time would decreese the availability of the RHRS to remove decay heat and for the RHRS suction relief valves to assist in low-temperature overpressure protection.

It also should be noted that,

.' since the operable loop as well as the inoperable loop would be subjected to any unlikely pressure transients occurring during Modes 4 or S, removal of power to the suction valve being tested (on the inoperable loop) would not alter the impact of a pressure transient in RHRS piping inside or outside of containment.

Improvement S - the licensee cannot confirm that the RHRS suction velves would not have the capability of opening against full RCS pressure.

However, these valves are provided with an OPI feature whic5 prevents opening when RCS pressures exceed 425 psig.

The OPI is tested in accordance with TS 4.5.2.d.1 once every 18 months.

Additionally, these valves are deenergized during power operation.

Therefore, the likelihood of an ISLOCA scenario due to an inadvertent open signal when the RCS is at full pressure is extremely low.

On this basis, the licensee does not plan to downsi2e the motor actuators.

We have completed our evaluation of the licensee's March 5, 1991 and July 24, 1991 submittals and have concluded the following:

The licensee has adequately identified any differences in RHRS configura-tion and design / operational characteristics that exist between WCGS and the reference plant (Callaway) addressed in WCAP-11736.

Both WCGS and Callaway are SNUPPS plants.

As a result, these differences are insig-nificant and the analyses and conclusions presented in WCAP-11736 for Callaway are directly applicable to WCGS.

The licensee has adequately addressed the five plant improvements delineated in Refersnce 2.

Where deviations between these improvements and the licensee's proposed actions were identified, the licensee has adequately demonstrated that the proposed actions provided at least an equivalent level of safety.

The proposed change to TS 4.4.9.3.2 (a and b) includes several revisions.

The current TS requirement to perirtically verify that power is removed from one of the two in-series RHRS suction / isolation valves in each train is intended to ensure that a single failure of either of the two common pressure transmitters (which provide the ACI signal to these. valves) does not result in both RHRS trains becomirg isolated from the RCS.

With the planned removal of the ACI circuitry, however, the only mechanism that can cause an isolation of both RHRS trains is now eliminated.

Therefore, valve power removal and its associated surveillance requirement become unnecessary.

The revised TS requires only that these valves be verified open at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Additionally, for the second of the two in-series isolation valves in each RHRS train, verification of the open position is changed from once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The 72-hour surveillance frequency for the isolation valves is consistent with that j

specified for the reference plant (Callaway) in the already approved i

WCAP-11736.

This frequency now becomes identical to the existing surveillance frequency for verifying the open position of the power operated relief valves (PORVs).

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.' On the basis of the above evaluation, we find the proposed TS changes and the proposed plan for RHRS ACI removal to be acceptable.

3.0 STATE CONSULTATION

In accordance with the Con 7ission's regulations, the Kansas State official wa:

notified of the proposed issuance of the amendment.

The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be release 1 offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that the amendment involves no significant-hazards considerattan, and there has been no public comment on such finding (56 FR 33673).

According'.y, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statetent or environ.nental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 REFERENCES

1.

R. A. Newton, Chairman WOG, letter to NRC, dated April 22, 1988.

2.

A. Ihadani- (NRC) letter to R. A. Newton, Chairman WOG, " Acceptance for Reference WCAP-11736, Rev. O, ' Residual Heat Removal System, Autoclosure Interlock (ACI) Removal Report' in Plant Specific Submittals," dated August 8, 1989.

3.

F. T. Rhodes (WCNOC), letter to NRC, dated March 5, 1991.

4.

F. T. Rhodes (WCNOC), letter to NRC, dated July 24, 1991.

Principal Contributor:

Harvey I. Abelson, SRX8/NRR Date:

September 12, 1991

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