ML20211R040

Requests Addl Info Re Util 860515 Response to IE Bulletin 85-003 Concerning Method to Estimate Switch Settings & Testing Planned to Verify Switch Settings

ML20211R040
Person / Time
Site:	Cooper
Issue date:	07/22/1986
From:	Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
References
IEB-85-003, IEB-85-3, NUDOCS 8607280085
Download: ML20211R040 (2)

Text

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c In Reply Refer To: JUL 22 1986 Docket: 50-298 Nebraska Public Power District ATTN: J. M. Pilant, Manager, Technical Staff-Nuclear Power Group P. O. Box 499 Columbus, Nebraska 68601 Gentlemen:

This is in regard to your response, dated May 15, 1986, to IEB 85-03. We have reviewed your response and find that we need additional information to complete our review.

Specifically, we request that you provide us with the following:

Your method to estimate switch settings, The testing planned to verify switch settings.

This request for information to clarify your original response is covered under the blanket clearance number 3150-0011 approved by the Office of Management and Budget. Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C., 20503.

If you have any questions regarding the information requested, we would be pleased to discuss them with you.

Sincerely,

Original Signed by

'J. E. GAGLI ARDO" J. E. Gagliardo, Chief Reactor Projects Branch cc:

Guy Horn, Division Manager l of Nuclear Operations l Cooper Nuclear Station P. O. Box 98 Brownville, Nebraska 68321 ,

Kansas Radiation Control Program Director -

Nebraska Radiation Control Program Director bec: ee next page)

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GENERAL OFFICE Nebraska Public Power District " '

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hs NLS8600173 May 15,1986 Mr. Robert D. Martin - Regional Administrator I ,b@

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U.S. Nuclear Regulatory Commission g \ S )936 Region IV 611 Ryan Plaza Drive, Suite 1000 g Arlington, TX 76011 _

Subject:

Response to I E Bulletin No. 85-03 Cooper Nuclear Station NRC Docket No 50-298, DPR-46

Reference:

1) I E Bulletin No. 85-03 " Motor-Operated Valve Common Mode Failures During Plant Transients Due To Improper Switch Settings"

Dear Mr. Martin:

Pursuant to the requirements of Reference 1, Nebraska Public Power District (NPPD) is submitting the enclosed response on motor-operated valves in the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems at Cooper Nuclear Station (CNS). A schedule for completion of the switch setting program

, described in Reference 1 is also included.

Since no outage is scheduled for CNS in 1987, the District may not be able to complete the program in the time frame given in Reference 1 (November 15, 1987). The District plans to actively pursue this prcpam during the upcoming October through December 1986 outage, but due to other resource commitments, A

we cannot guarantee completion of the program at that time.

Therefore , NPPD anticipates that the valve testing will be completed during the following outage, scheduled for the period of February to April 1988, and that the final report will be submitted within 60 days of the completion of valve testing.

This is explained further in the enclosure.

Finally, the BWR Owner's Group is funding studies to identify any BWR - Unique aspects of this Bulletin with regards to valve identification, analysis methodology, valve operation requirements, and valve testing concerns. NPPD is following The Owner's Group effort and any BWR guidelines arrived at between the Owner's Group and the NRC staff on this issue will be considered for implementation at CNS.

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Mr. Robert D 'artin Pega 2 May 15,1986 i

Any changes to the enclosure that will result from implementing these guidelines will be reported to your office.

If you have any questions regarding this submittal, please contact my office.

Sincerely ,

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L. G. Kunci Vice-President - Nuclear Power Group LGK:dm12 /4(Daily 5)

, cc: Document Control Desk w/ encl.

U.S. Nuclear Regulatory Commission i

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STATE OF NEBRASKA)

)ss PLATTE COUNTY )

L. G. Kuncl, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District , a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this request on behalf of Nebraska Public Power District; and that the statements contained hu ein are true to the best of his knowledge and belief.

L. G.' Kuncl Subscribed 1n my presence and sworn to before me this /5d day of '//)mi , 1986

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NOTARY /PUpLIC m ent m u n un w nee n COLLEEN 88. KUTA er man es as 4.nu E

COOPER NUCLEAR STATION N PPD RESPONSE TO I N ITIAL REQL'IREME NTS OF IE BULLETIN 85-03 MAY 15,1986 NEBRASKA PUBLIC POWER DISTRICT Ammatt ,

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, NEBRASKA PUBLIC POWER DISTRICT i COOPER NUCLEAR STATION l

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NPPD Response to Initial Requirements of IE Bulletin 85-03 l

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MAY 15, 1986 i-1 i

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1.0 INTRODUCTION

The objective of this report is to respond to the require-ments of IE Bulletin 85-03 for Nebraska Public Power District's Cooper Nuclear Station (CNS).

IE Bulletin 85-03 pertains to motor-operated valve common mode failure resulting from improper switch settings during plant transients and accidents. The requirements of this bulletin can be broken down into two phases. Phase I requires the licensee to identify and document the design basis for all motor-operated valves in high pressure systems, establish a tentative schedule for implementation of Phase II, and submit this information to the NRC by May 15, 1986. The second phase includes valve testing under actual design pressures or providing justification for alternate method, establishment and implementation of a switch-

. setting program, and preparation of a final report to be submitted to the NRC by November 15, 1987.

This report is NPPD's formal response to the Phase I re-p quirements of IE Bulletin 85-03 for Cooper Nuclear Station.

, 1.1 Background As a result of several events at nuclear power plants during

which motor-operated valves (MOVs) failed to function on demand, i

IE Bulletin 85-03 was issued. The purpose of this bulletin is to l

request licensees to develop and implement a program to ensure

that switch settings on certain safety-related motor-operated i

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F R/IE Bulletin, P263 Page 2 valves are selected, set, and maintained correctly. These switch settings should accommodate the maximum differential pressures expected on these valves during normal as well as abnormal events within the design basis of the station.

In' general, licensees are required to implement a program to ensure that torque switch, torque bypass features, position limit switches, and overload relays for active motor-operated valves of high pressure safety-related systems are selected, set, tested, and maintained properly. To achieve these objectives, the following tasks were identified as being required by IE Bulletin 85-03:

(1) Review and document the design basis for the operation of each valve. This documentation should include the maximum differential pressure expected during opening and closing in both normal and abnormal events.

, (2) Establish the correct switch settings, including a program to review and revise, as necessary, the methods for selecting and setting of all switches.

(3) Change the individual valve settings as appropriate and demonstrate operability by testing the valves at the maximum differential pressure. If the maximum differential pressure cannot practicably be tested, provide justification, including the alternative to A maximum differential pressure testing. In addition, stroke test the valves, to the extent practical, to verify proper implementation of the switch settings.

(4) Prepare or revise procedures to ensure that correct switch settings are determined and maintained through the life of the plant. These procedures are to be consistent with the requirements of Item 3.2 of

Generic Letter 83-28. These procedures should include l provisions to monitor valve performance to ensure the

switch setting is correct.

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R/IE Bulletin, P263 Page 3 (5) A written report to the NRC containing the results of Item (1) above, including a program and schedule to accomplish Items (2) through (4) is to be submitted by May 15, 1986. Items (2) through (4) are to be completed by November 15, 1987, and a report submitted to the NRC 60 days subsequent to the completion of these activities.

The NRC Technical contact identified in IEB 85-03 was asked to clarify several points. The following information, pertinent to CNS, was obtained verbally over the telephone:

(1) The only "high pressure" systems of concern at CNS are the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems.

(2) All valves required for the system to perform its intended safety function (s) including inadvertent operation of valves should be considered in the 4

analysis. For example, if a valve is normally open and it has to stay open to perform its intended safety function, an inadvertant operation could result in it's i closure. Therefore, valve testing has to be performed to demonstrate the capability to open this valve.

(3) For the majority of BWRs (including CNS), the torque switches in the opening circuits of HPCI and RCIC MOVs are jumpered. In addition, the overload relays for these valves are wired for alarm only. However, the valve operability has to be tested to ensure valve operation and appropriate switch setting as necessary.

(4) All motor-operated valves, including those in the low

pressure pipes of HPCI and RCIC systems (suction lines)

! are covered under IEB 85-03.

A In responding to the requirements of IEB 85-03 for CNS, the

above clarifications have been considered and are factored into r

the analysis. I I

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R/IE Bulletin, P263 Page 4 1.2 Report Overview Section 2 of this report describes the methodology and process used to identify motor-operated valves in high pressure systems at CNS. Table 2-1 includes a list of MOVs considered in the response to IEB 85-03.

Section 3 includes a description of the valve testing p~ro-gram to be completed by November 1987, and a tentative schedule for this program. A bar chart depicting the schedule is also provided. Section 4 provides a list of references used in performing the analysis.

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R/IE Bulletin, P263 Page 5 2.0 ANALYSIS METHODOLOGY The objective of this section is to describe the process used to identify motor-operated valves in response to IE Bulletin 85-03. The analysis methodology also identifies the design basis differential pressure and the maximum differential pressure ei-pected during opening and closing of each valve for both normal and abnormal events.

Nuclear Safety Operational Analysis (NSOA), NRC inputs, and CNS component classification results were utilized to identify high pressure systems and the active MOVs required to support the systems' safety functions. Vendor design documents for selected MOVs and General Electric system design specifications were then entered into a systems operational analysis to determine the design and maximum expected differential pressure.of each MOV.

2.1 Definitions The following terms have been used in the report and are defined herein for clarification:

Motor-Operated Valve (MOV) - The entire valve assembly, g which includes the valve, the valve operator, and the motor; no further distinction will be made.

l High Pressure System - A system that experiences peak nuclear pressure while performing its required safety function.

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Single Failure - Single equipment failure or inadvertent  !

equipment operation such as inadvertent valve closure or opening.

Nornal Event - The normal BWR operating states and planned operations such as power operation or refueling, from which transients, accidents, and special events are initiated.

Abnormal Event - Plant transients and accidents caused by component failure, personnel error, or design basis events (DBE).

2.2 System / Component Identification The Nuclear Safety Operational Analysis (NSOA) was used to identify the high pressure systems and their respective safety functions that would be used to mitigate abnormal design basis events at CNS. The MOVs required to support the system safety function (s) were then identified using the Q-List, Piping &

Instrumentation Diagrams, and System Operating Procedures. The CNS NSOA provides a methodology for establishing the plant safety requirements at the systems level. From the NSOA, a list of essential systems identified in the normal and abnormal design F

basis events protective sequence was compiled. Information obtained from the NRC IE Bulletin 85-03 technical contact was then utilized to select the systems of concern and their respec-tive safety functions. The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems were identified as the focus of this response. ,

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l R/IE Bulletin, P263 Page 7 Based on the system level safety requirements established in the NSOA,. further evaluation of the plant requirements were per-zormed to identify the system's safety classification. The safety functions of the HPCI and RCIC systems are to provide reactor coolant makeup during plant accidents and transients, and to automatically isolate in the event of a steam supply line i break.

For each MOV in the HPCI and RCIC system, component opera-

• tional requirements necessary to facilitate system safety 1

functions were analyzed to identify the active MOVs. IE Bulletin 85-03 states that single equipment failures and inadvertent i

equipment (such as operations inadvertent valve closures or openings) that are within the plant design basis should be i assumed. Thus, a normally open MOV, which must remain open to achieve system safety function, is considered to be an active l

component in this response. The list of active MOVs along with i .

their component description is provided in Table 2-1.

2.3 System Differential Pressure i e The design differential pressure and the maximum differential pressure expected during normal and abnormal operational modes vere identified for each MOV. The design differential pressure i

j was obtained from the original vendor's component design specifi-cation [1] and can be found in Table 2-1.

. I R/IE Bulletin, P263 -Page 8 j f General Electric System Specifications [2,3] and system flow i I

diagrams [4, 5) were then utilized to derive the maximum expected i di f f e rential pressure each MOV will experience during design basis events. The maximum expected differential pressure is conservatively' considered to be the maximum upstream or downstream pressure, whichever is larger. No credit is taken for the ambient pressure. Thus, the maximum expected differential pressure presented in Table 2-1 will be the most conservative enveloping differential pressure that could be experienced by the MOVs during various plant. operational modes.

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I 3.0 SCHEDULE FOR PHASE II ACTIVITIES The objective of this section is to establish a tentative schedule for. implementation of Phase II activities consistent with the requirements of IE Bulletin 85-03. Tasks required to complete the Phase II effort are identified. In addition, plant outages scheduled for 1986 - 1988 are included since valve test-ing will have to be performed during plant shutdown conditions.

NPPD reserves the right to deviate from requirements it imposes upon itself in this section, if required due to unforeseen operation constraints.

In order to complete Phase II activities, the following tasks must be performed:

(1) Using the results from Phase I (Table 2-1), establish the correct switch settings. This includes a program to review and revise as necessary the methods for selecting and setting all switches for the required valve operation (open, close). It should be noted that all torque switches in opening circuits of valves identified in Phase I are jumpered and, therefore, are not of any concern. In addition, overload relays for the valves of concern are wired for alarm only.

(2) Provide justification for continued operation (JCO) in accordance with the CNS Technical Specifications for

! A valves considered " inoperable" as a result of Task 1.

(3) Change valve switch settings if necessary, to those j established in Task 1. . Prior to any switch setting l adjustment, a summary of findings as to valve operability will be prepared.

' (4) Demonstrate valve operability by testing the valve at .'

the maximum differential pressure (MDP) identified in Phase I (see Table 2-1) or provide justification for alternate method of determining operational readiness. i

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1 R/IE Bulletin, P263 Page 10 (5) Stroke test each valve to the extent practical.

(6) Prepare or revise procedures, as necessary, to ensure that correct switch settings are determined and maintained.

(7) Submit a report to the NRC summarizing the results of Tasks 1 through 6 above. This report will include a verification of completion of Phase II activities consistent with the requirements of IEB 85-03.

Several Phase II tasks specified in the previous section must be performed during plant shutdown. The following is a list of the scheduled outages during 1986-1988:

Year Duration 1986 October 5 - December 14 1987 None 1988 February 28 - April 10 NPPD will attempt to demonstrate the operational readiness of these valves during the 1986 outage. However, because of the existing time constraints and other major plant modifications, which have been previously planned, it may not be possible to safely complete the required testing for all valves listed in Table 2-1 during this outage. Therefore, NPPD anticipates that A valve testing will be completed during the following outage, scheduled for the period of February to April 1988. The final report required by Phase II will be submitted within 60 days of the completion of the required actions.

In order to comply with the requirements of IE 85-03, NPPD has prepared a tentative schedule for completion of Phase II activities. Table 3-1 includes a bar chart depicting the pro-posed schedule.

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4.0 REFERENCES

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l 1. Anchor Equipment Company, List of MOVs and Their Respective i Design Specifications. Contract No. E69-7, 1969. ,

i 2. General Electric High Pressure Coolant Injection Data Sheet. I GE Document 257HA354AN, Rev. O, Oct. 1968.

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3. General Electric Reactor Core Isolation Cooling Data Sheet.

l GE Document 22AI354, Rev. 4, Oct. 1968.

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4. " Reactor Core Isolation Cooling System", General Electric (APED) Drawing No. 729E719BC, Sheet 1 of 1, Rev. O, Aug. 13, i i

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5. "High Pressure Coolant Injection System", General Electric I (APED) Drawing No. 729E720BB, Sheet 1 of 1, Rev. O, Aug. 27, 1969.  ;

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• NPPD-CNS IED B5-03 RESPONSE TABLE 2-1 -

Mov Data Summary Table (RCIC~ -

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f,.i.. ( t. !( } l Description I l Pressure i Diffs.rential l Operation ! Comment l [

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t 49 II - MOv- MO IS l RCIC 5 team I ntina r d l Open for Core C7al. Supply i 1146 psi i 1135 psia l l _Open l See Note 5 l 1 I Isolation l Close f or Nys. Steam Line l l ( lose j j f l l l ! solation l l l l l L 1

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11:15 ps i a Open l Nee Note 5 '

P it -MOV-Mulb l H( IC Steam Out- l Open for Cure Cool. Suppb l 1146 psi l l l l tinar d Isolation l C l o .ie f of Sys. Steam L iese l l l LIuse l l 1 Isolatinn f l l l l l

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, I  ; i I I I I fa : i wpv-Mo1H l RCIC Supply from i Open/Close for Core 1 50 psi l 2 14 psia l Active l See Note 2 l  :

l E m.a r pen c y t om ten- 1 Coolinel Supply l l l l l f i l sate Storage Tat,k l l l l l l

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i m It Mov Mf t.'o l RCIC pump l Open for Core Cooling l 1925 psi l 1712 psia l Open i See Notes 315 i j t l Discharge l Supply l l l l l  :

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l I I 1 l l i l I i i l i f m ir uov.uo21 l p;IC Injection to 1 Open for Core Cooling  ! 1925 psi l 1712 psia Open See Notes 3& 5 l [

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l l l l l l 1 I I I I I Ni li Vov MO2/ l RCIC Pump Minimum l Close/Open for Core i 1500 pst i 1212 psia Active l See Notes 3 4 & St 1

l Flo. Recirc to l Cooling supply l l l l To: us  ! l l l l s

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we f*-90v-Main l RCIC T*st Return l Close for Core Cooling l 1923 pst 1212 psia l Close See Notes 3, 4 & Sl i I poot l Supply l ,

! l l l I I I l l l f.  !( -MOv-M041 l RCIC Supply from Close/Open for Core 1 50 psi l 17.3 psia 1 Active l l l i l l Torus Cooling Supply l l l l l i I I I I l l l l 1 l l l l in i t MOV-M0111 l RCIC Steam supply l Open for Core Cooling l 1146 psi l 1135 psia l Open l See Note 5 l t I to RCIC Tur tilne l Supply I l l l i - l I l l l l

! I I I  ! I l I pr It "OV-Molt? l Ausillary Coo l itig 1 Open for Core Cooling l 1500 psi  ! 1212 psia l Open l See Notes 4 &5 l l Supply Supply l l l l l

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i . ..,.g i. .. ,e e i t l l l Design l Man. Espected l Required l l i g . i. . . , i f 6. ,. e i on l Component i Operatinnal Requirements IDIfferential lOperatianal l Operation l t ,

4 4 ..t.. ( Lli ) l Desce'ption l l Ps essue e lDiffeeenttal l l Comment l L l l l iPressure l l l l i

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I rth ! *k)v - Va l :. l Steam Supply in i Open for C.n e Cooling I 1146 psi l 1135 psia l Open l See Note 5 1  :

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. l Open for Cor e Cool. Supply l 1146 psi l 1135 psia l Open l See Note 5 l

\ tm.te rt Isolatinn l Ctnse int Nys. Steam L i ne l l l ( lose j l l l !sulatnun l l l l 1 I I I i -1 1 I I I l l  ;

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. j Steam Suppt y Out- l Open for (' o r e Cool. S_ opp.I,y l 1146 psi l 1135 psia l_Onen i See Note 5 l  ;

l linar d Isol,stion l Cluse f ov Sys. Steam L i v e. l l l close l l ,

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-7 I I I I I l f, HH I r.mV Mit 1/ l Pump Suction t e ne l Open/Close for Core  ! ISO psi 1 2H psia l Active l See Note 1 l F

, l E rper genc y L onatet t- l Cooling Supply l I l l f l s,ete 5tnrao Tank l l l l l j

__ _ _ _ _ _ _ _ . _ _ _l l l l l l l i e th ! Mov - M01'l l HPCI Injection i Open/Close for Core l 1325 pst i 1212 psia ( Active i See Note 5 l 1 1  ! Cooling supply l l l [

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, Hsf I uo v - UO .i l l HPCI-P-MP l Open for Core Con 1 Supply l 1925 psi l 1712 psia l Open i See Note 5 l {

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' I I I I I I l 6 mai env-uo11 l HPCI-P-MP Test l Close for Lore Cooling i 1925 pst i 1212 psia l Close l See Notes 3, 4 A 5l l Bypass to Emer. I Se,pply l l l l l t

! l Contiensa t e Storage f l l l l l 1  ! Tank l l l l l l l,

i l i I l l l l k i i I i l i I l l Hh ! nov 90.". l HPCI-P-MP Minimum i Close/Open f o r' Core i 1500 psi l 1212 psia l Active l See Notes 3, 4 & 51 l i l Flow Bypass t_ine l Co olinry Supply l l l l [

] Isolation l l l l l l I l 1 l l 1 1 I l HH I Mov-MOSH l HPC1 Pump 5.setion I Close/Open for Core l 160 psi 17.1 psia l Act we l l e j trom Suppressino l CooIing Supply j l l l l l l Poul l \ l l l l

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! IEB 85-03 RESPONSE i MOV Data Summary Table (HPCI)

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4 NOTE 1: The maximum pressure drop in the piping from the reactor vessel to the turbine is 15 psi. This pressure drop has conservatively been ignored in the analysis.

NOTE 2: The maximum expected operational pressure is ' considered to be the sum of elevation difference of about130 feet between the pump suction and the emergency condensate storage tank and ambient pressure of 14.7 psi, i

NOTE 3: Pressure drop introduced by orifices has been conservatively ignored.

NOTE 4: The maximum expected operational pressure is considered to be the maximum pump discharge pressure.

NOTE 5: Values subject to change due to plant modifications as a result of the ATWS rule implementation. ,

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l Enclosure 1, P263, T/ Table 2-1 (Cont)

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4 TABLE 3-1 COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT IE BULLETIN 85-03 RESPONSE PROPOSED SCHEDULE FOR COMPLETION OF PHASE II ACTIVITIES

! SCHEDULED OUTAGES l l l l TASK l 1 l' l

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1 I I I I I I I i

1. ESTABLISH PROGRAM l. l l J l i I l l 1 1 i l I I I I I I i

2. WRITE JCOs l l . l l J l l l l  !

. (if necessary) l I i I i i l l l l 1 l l l l l l l 1

3. WRITE

SUMMARY

OF l l 1, .I I l l i FINDINGS AND CHANGE I l l~ l l l l j SWITCH SETTING l l l l l l l l l 1

1 I I I I I I I I I 4. VALVE TESTING, l I. .I l l l l . l . I 1

PRESSURE l l~ l l l l l l l

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. 5. VALVE TESTING, l l l l l l l . l . l STROKE i l l l l l 1 I I I I I I I I I I l l

6. WRITE / REVISE SWITCH l l l. l l. l l SETTING PROCEDURE l l l l l l l l l

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7. FINAL REPORT l l l l. I l I l .I I I I l~ l 1 I i 'l I I I I I I I I i

! I I I I I I I I I l i , I I I I I I I I l l June Jan. June Jan. June i 1986 1987 1988 l

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