ML20214J953

Forwards Revised Response to IE Bulletin 85-003, Motor- Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings. Methodology & Process Used to Identify motor-operated Valves Discussed

ML20214J953
Person / Time
Site:	Cooper
Issue date:	11/26/1986
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Martin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
IEB-85-003, IEB-85-3, NUDOCS 8612020072
Download: ML20214J953 (17)

Text

I hh Nebraska Public Power District November 26, 1986

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Mr. Robert D. Martin, Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011

Subject:

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

Reference:

1) Letter from L. G. Kuncl to R. D. Martin dated May 15, 1986, " Response to IE Bulletin No. 85-03 Cooper Nuclear 8tetion"

2) Letter from J. E. Gagliardo to J. M. Pilant dated July 22, 1986, requesting additional information on IEB 35-03

Dear Mr. Martin:

Pursuant to the requirements of IE Bulletin 85-03, Nebraska Public Power District (NPPD) submitted Reference 1 on motor-operated valves in the High Pressure Coolant Injection and Reactor Core Isolation Systen.s at Cooper Nuclear Station (CNS). In this submittal, the District stated it was following the BWR Owner's Group efforts in identifying any BWR-unique aspects of the Bulletin and would consider implementing any guidelines that resulted. The Owner's Group effort is now complete and NPPD is revising its submittal to incorporate the developed methodology and to provide the additional information requested in Reference 2. Changes in the submittal are indicated by revision bars in the margins.

The District has reviewed this submittal against the information presented in IE Notice 86-93 "IEB 85-03 Evaluation of motor-operators identifies improper Torque Switch Settings." It was concluded that the notice was not applicable since CNS has no installed Rotork valve actuators.

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

Sincerely ,

W Jay M. Pilant Technical Staff Manager Nuclear Power Group 8612O20072 861126 JMP/gs:rt25/3(4) PDR ADOCK 05000298 Enclosure G PDR cc: '

ocument Control Desk w/ enclosure /

U.S. Nuclear Regulatory Commission h

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l NEBRASKA PUBLIC POWER DISTRICT f COOPER NUCLEAR STATION I

i Revised NPPD Response to Initial Requirements of IE Bulletin 85-03 1

l November 26 , 1986 l

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1.0 INTRODUCTION

The. objective of this report is to revise the Nebraska Public Power District's May 15, 1986 report in response to IE Bulletin 85-03.

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 safety related 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.

NPPD's initial response to the Phase I requirements of IE Bulletin 85-03 for Cooper Nuclear Station was submitted to the NRC by May 15, 1986. In this' submittal, the district stated it would follow the BWR Owner's Group efforts on discerning any BWR unique aspect of this bulletin. Subsequently, NPPD has participated in the BWROG meetings and, as a result of several clarifications and changes, is revising its licensing response to incorporate the methodology developed in BWR Owner's Group Report.[2]

R/85-03 Response, P263 Page 2 In addition, information requested by the NRC in Reference 3 is also addressed below:

o Method to Estimate Switch Settings Vendor design specifications . and component / system operational requirements will be reviewed to identify the various parameters that affect switch settings.

The maximum delta P values listed in Table 2-1 will be used to calculate the maximum required thrust and maximum required stem torque for each valve listed.

This required torque.value will be used in conjunction with the characteristics of the torque switch spring pack to determine the correct set points for the torque switches for each valve. The required operation of each valve will be analyzed to determine the correct limit switch set points.

o Switch Setting Verification A network of testing instrumentation will be utilized to detect and record dynamic baseline signatures of critical parameters (i.e., valve cycle time, torque and limit switch actuation points, developed thrust, running load and dynamic motor current).

This network of testing instrumentation will be similar to that evaluated by NUREG/CR-4380. This baseline signature information, which can be stored electronically as well as in hard copy form, will be used to develop motor running load threshold values.

p These threshold values can be monitored periodically to identify and trend Motor Operated Valve performance.

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l This network of testing instrumentation will allow dynamic torque / limit switch actuation point signature traces to be obtained periodically to verify correct

. switch setpoints are being maintained.

1.1 Background

As a result of several events at nuclear power plants during which motor-operatcd valves (MOVs) failed to function on demand, IE Bulletin 85-03 was issued. The purpose of this bulletin is to l

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o R/85-03 Responso, P263 Page 3 request licensees to develop and implement a program to ensure that switch settings on certain safety-related motor-operated 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

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the maximum differential pressure. If the maximum differential pressure cannot practicably be tested, provide justification, including the alternative to maximun differential pressure testing. In addition, stroke test the valves, to the extent practical, to verify proper imp.'.ementation 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

R/85-03 Response, P263 Page 4 Generic Letter 83-28. These procedures should include provisions to monitor valve performance to ensure the switch setting is correct.

(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.

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 required to be addressed in response to IEB 85-03.

Section 3 includes a description of the valve testing pro-gram to be completed 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/85-03 Response, 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 identifi'es the designed differential pressure and the maximum differential pressure expected during opening and closing of the selected valves during design basis events.

NRC inputs and methodology adopted by the Boiling Water Reactor Owner's Group (BWROG) were utilized to select safety-related motor-operated valves in high pressure systems. Vendor design documents and formulation recommended in the BWROG report [2] were then employed to determine the design and maximum expected differential pressure for 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, which includes the valve, the' valve operator, and the motor; no further distinction will be made.

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

Single Failure - Single equipment failure or inadvertent equipment operation such as inadvertent valve closure or opening.

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

R/85-03 Response,-P263 Page.6 Abnormal Event - Plant transients and accidents caused by

. component failure, personnel error, or design basis events' (DBE).

2.2 Component / Operational Requirements Identification From the BWR09 deport, the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems were

. identified as the focus' of this response for Cooper Nuclear Station. Please note that the BWROG Report provides a generic response to IEB 85-03 that is applicable to the spectrum of BWR/3, BWR/4, BWR/5, and BWR/6. Thus, a number of system and piping arrangements addressed in the BWROG Report (i.e., HPCS -

. high pressure core ' spray). have been reviewed and found not i . applicable to Cooper Nuclear Station. 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 break. It was indicated in the BWROG Report that HPCI and RCIC are not required

, to perform active safety action / function during normal events.

In the original IEB 85-03 submittal, MOVs and their required operation were selected based on the assumption that single equipment failures and inadvertent equipment operation (such as 4

inadvertent valve closures or opening) that are within the plant

' design basis should be considered. Thus, a normally open MOV, which must remain open to achieve system safety function, was considered to be an active component. However, in this revised IEB 85-03 submittal, system analysis methodology prescribed in the BWROG Report has been adopted and HPCI and RCIC systems are J

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R/85-03 Response, P263 Page.7 assumed in their normal standby condition at the start of a design basis event. This assumption was made based on the low probability of the system being in a test mode or out of service during the occurrence of an abnormal event. Thus the required active safety actions of the MOVs are established assuming the systems are in their normal standby condition unless the event can result in the valves changing position.

2.3 Summary of Revisions Based on the methodology modification identified in Section 2.2, a number of valves and their required active operation were revised f;om the May 15, 1986 submittal. A summary of revisions made along with their justification is provided below, o RCIC-MOV-MO 15, 16 and HPCI-MOV-MO 15, 16 - HPCI and RCIC Steam Inboard and Outboard Isolation Valves Valves normally open and required to remain open to provide core cooling supply. Thus, their only required active operation is to close following a system steam line break.

o RCIC-MOV-MO 18 and HPCI-MOV-MO 17 - System Suction from Emergency Condensate Storage Tank Isolation Valve Valve normally open and required to remain open to provide pump suction from ECST. Thus, their only

, required active operation is to close when system pumps are taking suction from the torus.

o RCIC-MOV-MO 20 and HPCI-MOV-MO 20 -

System Pump Discharge Block.

Valves normally open and required to remain open to provide core cooling supply. Not primary containment isolation valves, thus they do not have any required active operation.

7, R/85-03 Response, P263 Page 8 o RCIC-MOV-MO 30 and HPCI-MOV-MO 21 - System Pump Test

[ Bypass to Emergency Condensate Storage Tank

, Valves normally' closed and required to remain closed to prevent diverting core cooling supply. It has no required active operation.

o HPCI-MOV-MO 19 - HPCI Injection f, Valve normally closed and required to open to provide core cooling supply. Not a primary containment isolation valve, thus its only required active operation is to open.

The list of active MOVs along with their component description is provided in Table 2-1.

2.4 System Differential Pressure The design differential pressure and the maximum

) differential pressure expected during a design basis event were identified for each MOV. The design differential pressure was obtained from the original vendors' component design specification [1] [4] and can be found in Table 2-1.

Formulation developed in the BWROG Report were then utilized to derive the maximum expected differential pressure each MOV will experience during a design basis event. The calculated maximum expected differential pressure are presented in Table 2-1.

R/85-03 Response, P263 Page 9 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 and procedures 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 valves considered " inoperable" as a result of Task 1.

(3) Change valve switch settings if necessary, to

• hose established in Task 1. Prior to any switch setting adjustment, a summary of findings as to valve operability vill be prepared.

(4) Demonstrate valve operability by testing the valves that have an active safety function at the maximum differential pressure (MDP) calculated in Phase I (see Table 2-1) or provide justification for alternate

,a TO L R/85-03 Response, P263 Page 10' method _of determining operational readiness. As indicated in IEB 85-03 (page 5), those valves whose only active safety action is to close to isolate a break in the line_ containing the valve.(i.e., HPCI-MO-15, 16 and RCIC-MO 15, 16) are not required to be tested.. Table 2-1 provides the list of valves, involved in Phase II.

(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 20 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

" valve testing will be completed during the following outage,

R/85-03 Response, P263 Page 11 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. l i

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4.0 REFERENCES

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

2. General Electric BWR Owner's Group Report on the Operational Design Basis of

, Selected Safety-Related Motor-Operated Valves, NEDC-31322, September, 1986.

3. Letter from J. E. Gagliardo (NRC) to J. M. Pilant (NPPD) dated July 1986.

4. Crane Company drawings K-7190 and K-8104, Contract E69-4, 1969.

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9 Enclosure 1-P263 NPPD-CHS IEC 85-03 RESPONSE TABLE 2-1 MOV Data Summary Table (RCIC)

Component Design Max. Expected Required Identification Component Operational Requirements Differential Operational Active l Code (CIC) Description Pressure Differential Operation Comment Pressure RCIC-MOV-MOIS RCIC Steam Open for Core 1146 pst 1091 psi Close Valve not required Inboard Isolation Cooling Supply to be tested at Close for System Steam manimum OP Line Isolation RCIC-MOV-M016 HCIC Steam Open tur Core 1146 psi 1U91 psi Close Valve not. required l Outboard 1%nlation Cooling _ Supply to be tested at Close for System Steam maximum O P Line Isolation RCIC-MOV-M010 RCIC Supply from Open/Close for Core 50 psi 14 psi Close Emeroency Conden- Cooling Supply sate Storage Tank RCIC-MOV-MO2O System Pump Open for Core Cooling 1925 pst 1212 pst None Valve not required l Discharge Block Supply to be tested at i maximum dP RCIC-MOV-M021 RCIC Injection to Open for Core Cooling 1925 psi 1222 psi Open Reactor l Supply RCIC-MOV-M027 RCIC Pump Minimum Close/Open for Core 1500 psi 1335 pst Open Flow Recirc to Cooling Supply 1338 psi Close Torus RCIC-MOV-MU3d RCIC Test Close for Core 1925 psi 1212 pst Hone valve not required Return to ECST Cooling Supply to be tested at manimum 6P RCIC-MOV-MO41 RCIC Supply from Close/Open for Core 50 psi H8 pst Open Tarus Cooling Supply 65 psi Close RCIC-MOV-M0131 RCIC Steam Supply Open fnr Core Cooling 1146 pst 1091 psi Open to RCIC Turbine Supply RCIC-MOV-M0132 Auxiliary Cooling Open for Core Cooling 1500 psi 1311 pst Open Supply Supply Page t of 2 an -

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Enclosure 1 -

p263 NPPD-CNS ,

IEB 85-03 RESPONSE TABLE 2-1 MOV Data Summary Table (HPCI)

Component Design . Man. Espected Required Identification Component Operational Requirements Differential Operational Active Code (CIC) Description Pressure Differential Cperation Ccmment Pressure HPCI-MOV-M014 Steam Supply to Open for Core Cooling 1146 psi 1091 psi -Open Turbine Supp1y HPCI-MOV-M015 Steam Supply Open for Core 1146 pst 1091 psi Close Valve not required Inboard Isol.ition Cooling Supply to be tested at Close for System Steam Line Isolation maximum d P HPCI-MOV-M016 Steam Supply Open for Core 1146 psi 1091 psi Close Valve not required Outboarel !sonation Cooling Supply to be tested at Close for System Steam Line Isolation maximum OP HPCI-MOV-M017 Pump Suction from Open/Close for Core 150 psi 14 psi Close Emergency Conden- Cooling Supply sate Storage Tank HP C I -MOV- M019 HPCI Injection Open/Close for Core 1325 pst 1118,2 psi Open Cooling Supoly HPCI-MOV-M020 Steam Pump l Open for Core _1925 psi 1212 psi Discharge Block None Valve not required l Cooling Supply to be tested at maximum O P HPCI-MOV-M021 HPCI Pump Test Close for Core 1925 pst 1212 psi None Bypass to ECST Valve'not required Cooling Supply to be tested at manimum O P 1

HPCI-MOV-M025 HPCI-P-MP Minimum Close/Open for Core Flow Bypass Line 1500 psi 1014 ps i_ Open CoolinD Supply 1030 psi Close Isolatton HPCI-MOV-M058 HPCI Pump Suction Close/Open for Core 150 pst 98 pst Open from Suppression Cooling Supply 64 psi Pool Close l

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TABLE 3-1 COOPER NUCLEAR STATION hEBRASKA PUBLIC POWER DISTRICT IE BULLETIN 85-03 RESPONSE PROPOSED SCHEDULE FOR COMPLETION OF PHASE II ACTIVITIES SCHEDULED OUTAGES l l

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1 I I

I l l l l l l 1 l l TASK l 1 i i i i i l l l l l l

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

2. WRITE JCOs l l . I l J l l l I (if necessary) I I i i 1 1 I I I I I I I I I I I I I. I l .I l 1 1

3. WRITE SUWARY OF l l l l l l FINDINGS AND CHANGE l l l l l l l l I l l l I SWITCH SETTING l I I I I I I I I I

4. VALVE TESTING, I l. ,I l l l - l .l

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. I l l i l I l l l l PRESSURE

!  ! I I I I I I I VALVE TESTING, l l l l l l I

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STROKE l l l l l l I I l l 1 I I I I I I I

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6. WRITE / REVISE SWITCH I l l i I I i l l SETTING PROCEDURE I I i I I I I I l~ l I

7. FINAL REPORT l l l  !. I I I l .I I I I l~ l i i i ~l l 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 i June Jan. June Jan. June l 1986 1987 1988 1 l

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