ML20214W882

Safety Evaluation Supporting Util 850524,1104 & 861105 Submittals Re Process Used to Identify Deviations from Emergency Response Guidelines.Justification Provided for Identifying Deviations Acceptable

ML20214W882
Person / Time
Site:	Wolf Creek
Issue date:	12/02/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20214W869	List: ML20214W865 ML20214W882
References
TAC-57368, NUDOCS 8612100444
Download: ML20214W882 (4)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION VANSAS GAS AND ELECTRIC C0l1PANY KANSAS CIT,Y POWER AND LIGHT C0f'PANY KANSAS ELECTRIC POWER COOPERATIVE, INC.

WOLF CREEK GENERATING STATION DOCKET NO. 50 482 BACKGROUND The Safety Evaluation Report related to the Operation of Wolf Creek Generating Station (NUREG-0881, Supplement 5) published in March 1985 pointed out that Kansas Gas and Electric Company (KG&E) needed to identify the deviations from the generic Emergency Response Guidelines (ERGS) that exist in the Wolf Creek Ertergency Procedures. On August 30, 1985 the staff sent KG&E a request with regard to certain specific, potentially safety significant deviations and re-quested KG&E to review its methods for identifying technical deviations from

_ the ERGS.

KG&E's responses of November 4, 1985 and November 5, 1986 provided adequate justification for the specific deviations identified by the staff. They also described how KGAE identified potentially safety-significant deviations from the ERGS and listed the deviations together with an explanation of each deviation.

We have reviewed all of the deviations from the ERGS identified by the licensee and by the staff and have found the deviations to be acceptable as discussec below.

1 EVALUATION The deviations from the ERGS identified by the staff were discovered as a result of the Detailed Control Room Design Review (DCRDR). On May 24, 19PF the licensee reported its findings resulting from the verification of the l task analysis for the SNUPPS DCRDR. In reviewing these findings, the NRC staff questioned whether certain of these findings with regard to control room design wculd also necessitate deviations frera the ERGS in order to provide the reactor operator with adequate guidance for ccping with emergencies. The staff questioned findings 1, 6, 8, 9 and 10 discussed below:

Finding 1: Boron Injection Tank (BIT) Ten.perature 1

ERG ECA 0.0 Loss of All AC Power, Step 13a(2) calls for BIT temperature I indication but none is nrnvided at the Wolf Creek Generating Station (hCGS).

8612100444 861202 2 PDR ADOCK 0500 P

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This ERG step calls for a check of BIT temperature and is based on the Westing-house ERG reference plert that raaintains the boron concentration in the BIT at 21,000 ppr... The bli is maintained at an elevated temperature to keep the boror in solution. At the WCGS, the boron cercentration has been reduced from 21,000 to 2,000 ppm. At the lower concentration, all of the boron will remain in solution at ar.ibient temperatures. Therefore the cperator does not need to know the BIT temperature arid should not be concerned with it in this Lriegency Orerating Procedure (EOP).

The licensee has deleted the UCGS-EOF step that required the ocerater to check BIT temperature. This is ecceptable to the staff.

Finding 6:

WCGS Emergency Procedure E-1 (corresponding to ERG Procedure E-1: Loss of Reactor Or Secondary Ceolant), step 13A, required operator action at greater than 535 GPl; (Resioual Heat Removal water flow to the RCS). Since control room indicators are graduated in increments of 100 GPM, the value of 535 GPi' cannot be read accurately.

The licensee has reevaluated this matter and found that the value 500 GPM, as read on the flow indicators in the control room, provides sufficient indication of PPR flow into the RCS. This WCGS Emergency Procedure was changed to require operator action after verifying RPR ficw to the RCS to be greater than 500 GPM. We find this change to be acceptable.

Findings 8 and 9:

Findings eight and nine alleged that no instrumentation was provided in the control room for the verification of flow of Component Cooling Water (CCW) to the RHR heat exchanger and to the (Reactor Coolant Purrp) Seal Water Heat

. Exchanger. -

The licensee reviewed this matter and found nurrerous ways of verifying CCW flow including the Balance of Plant Cemruter Display in the control room, service loop flow indicators ard beat exchanger temperature indications. In addition, the control room annunciators signal conditions of high flow and low flow in the CCW System and the reactor coolant pump Seal Return Heat Exchanger.

Cased cr tFis information, we find that this llCCS instrumentation is adequatc for implementation of the generic Festinghouse ERGS and therefore re deviction from the ERGS is necessary. The direct application of the pertirient LPGs is therefore appropriate at the VCCS.

Finding 10:

Finding 10 asserts that the generic EEG L-5, Steam Generator Tube Rupture, Step P, calls for steamline radiation tronitors t.hereas none are provided in the WCGS.

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g The purpose of this EPG Step is to identify which stean, generator may have a leaking tott. The operator observes stean generator water level and indications of radioactivity in the secondary system. The k' CGS E0Ps direct the operator to utilize steam generator water level, the turbine driven

, euxiliary feedwater pump steam eybaust radiation monitor, the steant generator power operated relief valve p1tre radiation monitors, the steen generator blowcowr. radiation monitors, and steam cererator sampling to identify the steam generator with the leaking tube. Based on the availability of these alternative neans for identifying the leakino steem generator by the use of radiation monitors, we have concluded that these UCGS instrumentation and procedures will provide the reactor operators with adequate inferration and guidance for identifying the leaking steam generator.

As a result of its review of the ERGS and background material for potentially safety significant deviations from the ERGS, the licensee identified additional deviations that required further censideration.

Five of these deviations from the ERGS are based on differences in the design of the kCGS from that of the Westinghouse ERG refererce plant. The design differences are:

1. WGS has the added feature of automatic closure of the pressurizer PORY block valve at "2185 psig oecreasing". Thus, where the ERGS call for verification of operability and preliminary opening of one block valyc, the WCGS E0Ps call for verification of operability, but a block valve n.ay or may not remain open and the block valves would be opened intnediately before opening a PCRV.

2. Circuitry has been added at WCGS to ensure that undervoltaoe (UV) relays do not actuate when a reactor coolant pump (RCP) is started. The circuitry delays operation of the UV relays and is activated by resetting

. the Safety injection (SI) Actuation System. Thus a step was added in the procedure for natural circulation cooldown to reset SI prior to starting an RCP.

3. The UCGS does not have steam line radiction monitors as in the reference plant, but it does have stear generator PORV plume monitors that are not in the reference plant. Appropriate devietions from the ERGS were incorporated into the PCCS procedures as discusseo under finding 10 above.

4 It the LC65, the control rod position indicators, including the red bottom lights, are rot cperational during a loss of all AC power.

Operator instructions to check the rod bottcr lights and the rod position indicators have not been put into the KCCS E0P for loss of All AC Power. The operator mil observe neutron flux and the position of reactor trip and bypass breukers to verify reactor trip.

5. At the WCGS, in order to mininize ECC5 flow in the event of loss of en,ergency coolant recirculotten, certain valves may be coerated from the control rocm to modulate or throttle the flew c1 emergency coolant,

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whereas in the reference plantothe valves are either fully open or fully closed. Since modulation or throttling is the preferable method of controlling flow, the WCGS E0P instructs the operator to throttle flow rather than fully open or close valves and turn purros on or off.

We have reviewed these five deviations that are baseo on design differences between the WCGS and the ERG reference plant. We have concluded that these are safety significant deviations from the ERGS that are necessary to achieve the cbjectives of the ERGS. Furthennore, the procedural changes adopted et the WCGS provide appropriate instructions to the operator to cope with the pertinent emergencies.

The licensee's procedures also deviate from the ERGS involving reactor trip or safety injection. The ERGS have a guideline, E-0, for Reactor Trip or Safety Injection (SI) and a separate guideline ES-0.1 for Reactor Trip Response (without SI) whereas at the WCGS there is a procedure EMG-0 for Safety Injection and a procedure EPG-ES-02 for Reactor Trip. The licensee has taken the first four steps of E-0, that is the verification or check of Reactor Trip, Turbine Trip, AC Power and SI actuation and placed them in both EMG-0 and ENG-ES-02. Thus the WCGS operator can start with either the Reactor Trip Guideline or the Safety Injection Guideline and then transfer to the one which is more appropriate. The ERG Guidelines require the operator to start with E-0 and then, if appropriate, transfer to ES-0.1, Reactor Trip.

The licensee has indicated that simulator exercises have demonstrated the benefits of giving the operator direct access to the Reactor Trip Procedure EMG-ES-02. lle have reviewed the licensee's description of this deviation from the ERGS and found that the licensee has properly accommodated the transitions involved and that the KCGS procedures will accomplish the objectives of the corresponding ERGS. Therefore, this deviation from the ERGS is acceptable to the staff.

The licensee also provided additional guidance in two areas that was not provided explicitly in the ERGS. Additional guidance was pruvided throughout the EOFs in the event that the reactor vessel liquid inventory instrurrentation is not operational. Also, additional precautions are taken in EPG C-II, Un-contrclied Depressurization of All Stecm Generators, that will provide further assurance that cold overpressurization of the reactor vessel will not occur.

These additions to the E0Ps do not depart from the strategy cr irtent of the ERGS and do rot irhibit the implementation of other operator actions. These additions are therefore acceptable to the staff.

CONCLUSION We have reviewed the dif ferences between the 1.'estinghouse Owners' Group ERGS and the Knif Creek E0Ps that were identified from the staff's examination of the DCRDR and the licensee's revieu to identify deviations from the ERGS. Based on this review, we have concluded that these procedural changes adopteo at the WCGS provide adequate guidance and information to the operator to cope with emergencies and achieve the pertinent objectives of the bestinghouse generic ERGS.

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