SNRC-1348, Forwards Addl Info Re Low Power Operation at 25%,per 870616 Request.Info Should Close Out Correspondence in View of NRC Cancellation of Util 25% Power Request

Forwards Addl Info Re Low Power Operation at 25%,per 870616 Request.Info Should Close Out Correspondence in View of NRC Cancellation of Util 25% Power Request

ML20236J465
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	08/04/1987
From:	Leonard J LONG ISLAND LIGHTING CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
SNRC-1348, NUDOCS 8708060181
Download: ML20236J465 (10)

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LONG ISL_AND LIGHTING COM PANY Duwamm mamme 4 SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 JOHN D. LEONARD, JR.

VICE PRESIDENT . NUCLE AR OPE RATIONS sNRC-1348 AUG 04 887 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Request for Additional Information 25% Low Power Shoreham Nuclear Power Station - Unit #1  !

Docket No. 50-322 Gentlemen:

In accordance with your June 16, 1987 request for additional information concernino low power operation at 25% of full power, enclosed please' find ur response. We trust that this ,

information is respor sive to your request and should close out  !

correspondence in view of the Commissions cancellation of LILCO's 25% power request. However, LILCO has refiled this request in accordance with 3 0 CTR ,50.57 (c) on July 14, 1987. Should you require any additi_ mil information, please do not hesitate to contact my office.

Sincerely yours,

,m

_hy' I

/J. . eonard, Jr.

Vic President - Nuc ear Operations

-BEG /ap Attachment cc: R. Lo C. C. Warren W. Russell

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ATTACHMENT Question 1. Identify those safety-related pumps, valves or other mechanical equipment that would experience significantly different operating conditions, l particularly flow conditions, at low power J operation as opposed to full power operation.

Response 1:

Methodology All plant safety-related systems and equipment were investigated for the resultant effects of continued operation at 25% power. The primary consideration in this investigation was to determine those systems with flow characteristics that were power dependent. The systems were broken down into the following categories and reviewed appropriately:

o Systems Not Required to Operate During Normal Operation These systems (e.g. ECCS, Emergency Power),

which are required solely to mitigate the consequences of an accident, are unaffected by 25% power operation.

o Atmospheric Control and System Heat Removal These systems (e.g. RBSVS, RBCLCW, RBSW) have j been sized for continued plant operation at l 100% power but are designed to be operated ,

over the entire range of power operation. 1 Their sole function is to remove heat gains associated with plant equipment and personnel.

These systems are unaffected by prolonged operation at 25% power.

o Reactor Support Functions These systems (e.g. RWCU, CRD hydraulics) operate in their normal design configuration E lows and pressures) and will see the same service during full or partial power operation, o Steam and Power Generation Functions These systems (e.g. main steam, feedwater) are power dependent and have been designed to be operated over the entire range of power operation.

. _ _a

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Results Using the categories described above, a system operational l review of the safety related systems was performed to assess j the impact of prolonged plant operation at 25% power. The '

attached. table summarizes the results of this review. Eight

. safety related systems.have been identified as power dependent. These systems include the following:

o Neutron Monitoring System o' Reactor Building Service Water System o Reactor Building Closed Loop Cooling Water System o RBSVS/CRAC Chilled Water System o Drywell Air Cooling System o Reactor Recirculation System o Main Steam System o Feedwater System o Neutron Monitoring System Although power dependent, this system is designed to operate over the entire range of power operation. As a result of this review, it has been determined that operation at 25% power will induce no significant diverse operational conditions which would affect neutron monitoring system operation.

o Reactor Building Service Water System This system is considered power dependent in that the total service water flow requirements are dependent on the heat gains requiring removal from the various heat exchangers served by the service water system. This system is designed to operate with one or more pumps operating during all normal  ;

and upset conditions. As a result of this review, '

it has been determined that operation at 25% power will induce no significant diverse operating conditions which would adversely affect the reactor building service water system operation.

o Reactor Building Closed Loop Cooling Water, RBSVS/CRAC Chilled Water and Drywell Air Cooling Systems These systems are all power dependent as they remove heat gains associated with power operation.

All three systems are designed to operate over the entire range of reactor power levels and are designed for both normal and upset conditions. As a result of this review, it has been determined that operation at 25% power will induce no significant diverse operating conditions which would adversely affect system operation.

c.

J; o Reactor-Recirculation System This system is power dependent by virtue of controlling the flow rate. As a result of this

. review, it.has been determined that operation at' 25%' power will induce no' diverse'~ operational

-conditions'or system degradation. This has been confirmed by General Electric.

~o Main Steam System

c. The main steam system, although power dependent, is designed.to operate over the entire range of power operation. As a result-of this review, it has been determined that operation at 25% of rated thermal power will induce no;significant diverse operating 7-conditions which would affect the main. steam system.

.o Feedwater System The'feedwater system,.although power dependent, is designed to operate over the entire range of power operation. However, as.a result of our review, additional investigation of the inboard and outboard containment isolation check valves was warranted. The results of this additional review-are reported in the response to question 2.

Question 2. For those systems and components identified in Item 1 above, discuss and evaluate the potential impact of:a prolonged low power operation on the reliability of equipment. For example, where flow.

at low power is less than the-flow required to cause a check valve:to fully open, recent b experience has~shown that: rapid valve' degradation may occur (See INPO SOER No. 86-03). An industry stady showed that failure of certain check valve internals'could be caused by accelerated wear and fatigue damage due to flow induced vibration.

p Response 2: Feedwater Check Valves L The feedwater check valves investigated included i two swing type check valves and one globe type stop check, all installed in series (see Figure 10.4.7-2 of the USAR). All three valves are located in the recommended horizontal position. Together these three valves offer the degree of redundancy  !

necessary to ensure complete isolation in the event j of a pipe break outside the primary containment.

These types of valves were selected to minimize the ,

system pressure drop while still providing for rapid isolation.

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One aspect of the investigation involved _the

' development of minimum flow velocity requirements t

as specified by the valve manufacturer and in conformance with industry standards. A comparison of'these minimum flow velocities to the velocities expected during reduced power operation was then performed.

The results showed that on the basis of these calculations, failure of the check valves is not anticipated. In particular, f or the two swing check valves, it has been determined that the discs i of these valves will remain in a stable, partially opened position since feedwater flow will remain relatively steady due to the inexistence of flow control devices upstream of the valves to cause instability. The third valve, a globe type stop check, is specifically designed for a varying spectrum of flow conditions. This valve has been verified by the manufacturer to have sufficient flow to maintain the disc off the seat and is not expected to experience any operational difficulties during operation at 25% power.

The combination of these three valves in series coupled with the analysis which indicates little potential for failure, will result in no degradation in the performance of the safety function of these valves during operation at 25%

power.

In addition, LILCO has reviewed the Sar Onofre events documented in NUREG 1190. The report indicates that there were two principal causes for the valve failures:

1. Unstable flow conditions due to ficw disturbances upstream of the failed check valves.

2. Lack of positive locking devices in the check valve design of the disc-retaining nuts.

The flow disturbances present in the San Onofre f ailure do not exist for Shoreham. This design differs significantly from the San Onofre situation which incorporated pressure reducing valves just upstream of the failed check valves. Further, whereas the San onofre design did not incorporate positive locking devices on the disc-retaining nuts, the Shoreham design does.

4

. i I

I LILCO is also evaluating INPO SOER 86-03 which  !

addresses check. valve misapplication as a frequent cause of check valve failure. At Shoreham, a review of the check valve sizing criteria as recommended by the valve manufacturer was performed j for all three check valves. The results of the  !

review indicate that the expected flow velocities  ;

are in the vicinity of the minimum flow requirements for full disc lift as identified by the manufacturer. This situation, in addition to the valve type and valve location discussed above, should not result in rapid valve degradation at 25%

power operation. SOER 86-03 identifies the'need to  ;

~

establish a check valve preventative maintenance program to identify and rectify impending check valve problems. LILCO is still reviewing the recommendations concerning this preventative maintenance program.

To further support LILCO's analysis concerning  ;

operation of the feedwater check valves at 25% >

power, it should also be pointed out that Shoreham  ;

has operated at 5% power levels for some period of time. In particular, during the period of July 1985 to October 1985 (66 days), Shoreham operated at various power levels between zero and five percent. In January 1986, the inboard feedwater swing check valves (18V-1103) were disassembled for inspection and no problems were found. As we had ,

discussed in our May 8 memorandum,-the inboard  !

(18V-1103) and outboard ( AOV-036 ) feedwater check i valves are subject to Appendix J testing in accordance with the Technical Specifications. Of these four valves, three valves are due to be '

tested in January 1988 and the fourth in January 1989. Any resulting anomalies will be thoroughly investigated prior to returning the valves to service.

Question 3. Discuss and propose compensatory measures for monitoring accelerated equipment failures. Our experience indicates that the regular inservice testing including containment leak testing may not be adequate to detect certain valve internal failures. Valve disassembly should be specifically addressed as a compensatory measure. Discuss where the compensatory measures will be reported and updated in existing plant programs or procedures.

Discuss actions that will be taken as a result of ,

the outcome of compensatory measures.

Response 3: Whereas analysis and operating experience has shown that these valves could operate at reduced flow for extended periods, the feedwater swing check valves (18V-1103 and AOV-036) will be physically inspected f or indications of valve internal degradation.

This inspection will include a disassembly of the valves and inspection of the internals for signs of degradation. It will be performed at the first refueling outage following 25% power operation. ,

Should any anomalies be detected during the i scheduled inspections or leak rate testing, an  !

investigation into the root cause will be initiated I and compensatory actions taken prior to returning  :

the valve to service. The results of this i investigation shall be incorporated into the INE7 l NPRDS program and used as input to overall reliability studies of the feedwater system.

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