ML20203J331

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Investigation Rept on 940916.Noncompliacce Noted.Major Areas Investigated:Casual Factors Contributing to Emergency Feedwater Valve Leakage Event Described in Event Rept 94-016
ML20203J331
Person / Time
Issue date: 09/16/1994
From: Barron L, Stowers G, Taylor J
NRC OFFICE OF INVESTIGATIONS (OI)
To:
Shared Package
ML20203J264 List:
References
FOIA-97-401 1-96-040, 1-96-40, NUDOCS 9712190189
Download: ML20203J331 (122)
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{{#Wiki_filter:- t C Emergency Feedwater Valve Leakage Event Investigation Report { Team Members: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED i l James Taylor - Senior Nuclear Safety Engineer, NSEG ( i Event Date: 08/05/94 Report Date: 09/16/94 3 N 0- 1-96-040 EXHIBIT PAGE,/ OF8/_PAGE(S) \\ h l 9pai;g}e,97212 CHRISTI97-40: PDR J

l i C Meetify causal factors contributing to the Emergency Feedwater Valve g Isakage event describext in Licensee Event Report 94 016, Emergency Feedwater Isolation Valve Leakage. Prodde recommendations for reducing the probability of such an event .I securing in the future.

3. g Assess potentail generic implications of any deficiencies noted.

eg Sutunit report on results of investigation within aproximately one week. i s C g., o 6 ' 0 4 0' :- EXHIBlT (. PAGE F OF_/_f/_PAGE(S) 2

2.0 EXECUTIVE SI'MMARY. At 1220, on August 4,1994 with the reactor in a cold shutdown condition, plant operators determined that an Emergency Feedwater isolation valve. 'Or #1 Steam Generator was leaking by. Further investigation identified similar leakage in the Emergency Feedwater supplies to #2 & #3 Steam Generators. Subsequently it was determined that under accident conditions which require isolation of Emergency Feedwater, isolation valve leakage could exceed Safety Analysis assumptions. Maintenance activities were initiated to reduce Emergency Feedwater valve leakage. In addition, administrative controls were implemented to ensure Emergency Feedwater leakage is maintained within the bounds of Safety Analysis assumptions during accident conditions. ( [ 3 CASE NO. 1-9G-040 gxgigi7 PAGE 3 0F8/._.PAGE(S) 3

3.0 RECOMMENDATIONS ( EMBIT 8 l CASEND. ]. g r,. g,{ g PAGE F OF.///_ PAGE(S) 4 I l

l T 4.0 NARRATIVE C' - On August 4,1994 Maine Yankee was in a r:old shutdown condition making preparations to restart the plant following a maintenance outage to correct Steam Generator tube leakage. At approximately 1220 while performing a leak test of the Emergency Feedwater (BA) Isolation Valve (ISV) for #1 Steam Generator using normal system instrumentation, it was determined that EFW-A 338 leaked by at a rate of approximately 400 GPM. Imkage past #1 Steam Generator Emergency Feedwater Regulating Valve (FCV) EFW-A 101 at 75 GPM was also identified. Subsequent investigation determined that the actuator for EFW-A 338 was coupled to the disk approximately 180 degrees out of alignment. " Itis misalignment is believed to have occurred during valve maintenance performed during our 1992 Refueling Outage Maintenance activity was initiated to reposition the actuator / disk to the correct orientation. At 1700 on August 5,1994 it was determined that Maine Yankee's Steam Line Break Safety Analysis assumes zero leakage past Emergency Feedwater isolation and regulating valves. Therefore, at 1957 the NRC was appraised of the situation via the Emergency Notification System in accordance with the provisions of 10 CFR 50.72 (b)(2)i. A follow up notification was made at 1151 on August 6,1994 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW A-340. On August 10,1994 comprehensive testing was performed on the Emergence Feedwater isolation and regulating vahes for all three Steam Generators using precision test equipment. C These tests revealed leakrates for various individual valves and combinations of valves from O to 36 GPM per Steam Generator (See Atachment A). Safety Analysis sensitivity studies conducted in parallel with these tests established the following revised acceptance criteria: 1) Leakage past individual isolation / regulating valves should be ler,s than 40 GPM. The purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating valve to close. 2) Leakage past both the isolation and regulating valve to each Steam Generator should be less than 10 GPM. The purpose of this limit is to provide at least 30 minutes for operators to insure EFW is isolated to a faulted Steam Generator following a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to overfill. 3) Zero leakage to a Steam Generator with a tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due to overfill following a Steam Generator Tube Rupture, in order to meet the revised acceptance criteria, EFW-A 101 was adjusted to reduce leakage past EFW A 101 & 338 to zero GPM and administrative controls were implemented to ensure termination of any potential future leakage to a ruptured Steam Generator within 30 minutes using manual isolation valves (V). C_ EXHlBIT 8 cuno. 1 - 9 6 - 0 4 0 PAGE I OF/f/_.PAGE(S) 5

j

5.0 CONCLUSION

S, FACTS & RECOMMENDATIONS C Concluslos A: The EDCR process umi by Maine Yankee circa 1983 84 did not provide sufficient guidance to ensure all safety significant design funtions of EFW A. 338,339,340 were verified post instaliation. Facts: Recounseendations: ,e ( G OF8/_kGE(S) 6 PAGE

L Conchannos 5: The process used by the IST program to determine functional test requirements did not ensure that the function of EFW A 338,339 & {.- 340 to stop Emergency Feodwater flow was period'cally verified under i the IST program.. Facts: f = Recomunesdalloas: e t 1 p CASE NO, 1-96 040 .( EXHIBIT 3 PAGE 7 0F/f/_PAGE(S) 7 y --m.,.- ry.. - ---v. v --w--.. . - + - .m--- +-

i Concluslos C: The process used to identify functional test requirements did not ensure C. that all safety significant functions of EFW A 338,339, & 340 were verified following completion of maintenance activites on these valves. t Facts: Reca===odations: i l emH0. 1-96-040 s. C EXHIBIT 7 PAGE f 0F./fgPAGE(S). 8

e 4 Ceeclassion D ( Facts: Recommmendations: T 4 P s LC 1 4 + 8 CASE N0. 'l - 9 6 - _0 4 0 EXHIBIT 3 9 PAGE f OF./f4PAGE(S) .-.. ~.

a +m.* M.-. we-_-A-a.4.---aba4mm.-Ja-.,* wee t = ') t s k t Conchasion E: I (- I

r. cia:

Recosnawodations: l I f i I i I I r h i s e p. 8 6 f t I 4 r ? I 1 t i f t t 4 castno. 1 94-044 1 EXHIBIT 3 PAGE M -_OF/fg_PAGE(S)-, 10 ,,n ,m.~. -w w w sn --+m- ,w-w- -.w.. e -w-v~u-v - - - nn+ e-br*= ww -~ w r m n e-ree me s -e w-m - oo-rw w wemam,s-==w o me - w w o e 'n* www sww www w w ww s ,-*w-se-~,*w~w<n w ww - ~,W~w+w'w-*

d L Cance F; - ( Facts: l - Recomunesdations:- ? v r ? P 1 i L (.. P k ? v castNo. l-li-9 6 - 0 4 0 l E1 EXHlBIT II PAGE // __0F/f/_ PAGE(S .s w r,- a r,, ,-w.r,w,--+.+i.,.-,,,n. m.,, -,.n. --,,w

.. ~ _ - ? i t L h - 1 6,0_ - SEQUENCE OF EVENTS F i f E 1 i L I ? k i I + P-3 ? ? a. + b b' i EXHIBIT = 12 *' k OF Q pAggggy 1 .6 ~.s -.-,-,w.,.,,,,, en,..-m,_. .,,,.,,n..,. ,-e-~w., -.+. -.., - .,-,..,.c.,,-m, r, < + >, - r

t 4 ATTACHMENT A C Test As Found Results AsI. cit S.Q Valve (s) Clod 08/04/94 08/10/94 08/11/94 i EFW A 101 75 GPM 36 GPM 0GPM t EFW A 338 400 GPM 20 GPM 24 GPM 15 GPM 0GPM EFW-A 101 & 338 0GPM 0GPM EFW-A 101,338 & EFW 102 0GPM 0GPM 2 EFW A 201 17 GPM 17 GPM EFW A 339 0GPM 0GPM EFW-A 201 & 339 0GPM 0GPM EFW A 201,339 & EFW-202 OGPM 0GPM 3 EFW A 301 36 GPM 23 GPM EFW A 340 4 GPM 0GPM EFW-A 301 & 340 0GPM 0GPM EFW A-301,340 & EFW 302 ( t j two. 1 -i.0- 04 0 {- 3 EXHIBIT PAGE_/_,{ 0F./f/ PAGE(S) 13

1 i \\LO'MWYWWSWY 9 1 96-040 ( cettto. EXHIBl? 3 PAGE4.__OF_/ff_ PAGE 14

Z. 4 C Emergency Feedwater Valve Isakage Event Investigation Report Team Members: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED James Taylor - Senior Nuclear Safety Engineer, NSEG ( Event Date: 08/05/94 Report Date: 09/20/94 3 Q _ O EXHIBIT PAGE/I0F PAGE(S)

l } 1.0 CHARTER 'C 1. Identify causal factors contributing to the Emergency Feedwater Valve Lankage event described in LW Event Report 94-016, Emergency Feedwater Isolation Valve Lankage. i 2. Provide recommendadons for reducing the probability of such an event recurr'ng in the future. l 3. Assess potential generic implications of any deficiencies noted. 4. Submit report on results of investigadon within approximately one week. i I r ( CASE ND. 1-96-040 EXHIBIT 3 PAGE10FpPAGE(S) =. ---.-.-...-..-.--.--.a....-.---

f l 4 s i ^ 2.3 EXECUTIVE

SUMMARY

f C. At 1220, on August 4,1994 with the reactor in a cold shutdown condition, plant i operators determined that an Emergency Feedwater isolation valve for #1 Steam Generator was leaking by. Further investigation idendSed similar leakage in the l Emergency Feedwater (EFW) supplies to #2 & #3 Steam Generators. Subsequently ~ it was determined that under accident conditions which require isolatior. of Emergency Feedwater, isolation valve leakage could exceed Safety Analysis assumptions.. Maintenance activities were initiated to reduce Emergency Feedwater_ valve leakage. In addition, administrative controls were implemented to ensure Emergency j Feedwater leakage is maintained within the bounds of Safety Analysis assumptions during accident conditions. A review of this event has revealed the following weaknesses which contributed to this event. EFW flow isolation assumptions in our Safety Analy*' are not clearly o I presented in design basis reference documents. Our IST program does not require periodic verification of the ability of the 3 o EFW isolation and feed regulating valves to isolate EFW flow. I The EDCR under which the EFW isolation valvet were installed did not o require verification of the ability of these valves to isolate EFW flow as part ( of post installation functional testing, Post maintenance functional testing performed on EFW A 338 in 1992 did not o require verification of the valve's ability to isolate EFW flow, Maintenance procedures lack sufficient guidance to ensur: consistent proper o assembly of the EFW isolation valves. Lack of a single, comprehensive, readily availabe reference for Safety Analysis inputs and assumptions, seems to have played a key role in this event. The improved documenitation and retresvibility of Safety Analysis information to be offered by the Safety Analysis inputs & Assumpdons (SAID) which is cumatly under development, 4 should significantly reduce the the possibility of similar events once it becomes available for use. Management should continue to emphasiae/suppost dmely =:#= of this effort. Other significant activities which are' recommended include the following: initiation of an independent, broad scope, review of EDCRs and W.O.s to o determine the overall adequacy of functional testing performed at Maine Yankee. C Review of the IST program to ensure the current scope of the program and the o " 8#

CAstNO, 1-96-O40-PAGE OF3[_ PAGE(S)

.- -. =. - -.

criterie used to establish the functional et requirements for specific components are adequate. ~ Efforts to increase awareness among those who develop, review, and approve o functional tests, that the functional testing required by our IST program should not be relied upon to ensure adequate post maintenance functional testing. Revision of maintenance proceduits to provide additional guidance for o assembling the EFW isolation valves to ensure consistent proper orientation of the valve disk, body, seat, and actuator. ( INEND. 1-96-040 EXHIBIT 3 4 PAGE /f _OF$PAGE

3.0 RECOMMENDATIONS 1. Management should continue to emphasize / support t!wly completion of the SAID project. 2. Perform an independent review of a representative sample of EDCRs to more completely assess the overall adequacy of EDCR functional testing, j 3. Reclassify EFW-A-3M,339 & 340 as IST Category A valves. r Modify PED's exisdng in :,sretation of the definition of IST Category A valves to 4. include valves which are assumed by our Safety Analysis to provide flow isolation. 5. If recommendation 2 is adopted, the IST program should be reviewed to ensure all valves which meet the revised criteria are included in the program and properly classified. 6. Modify E IST program description to include the preciw criteria used to classify valves as Category A, D & C. i 7. Perform an independent review of a representative sample of W.O.s to more fully asseas the overall adequacy of W.O. functional testing. ( 8. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing required by the IST program does not necessarily ensure adequate post maintenance functional testing. 9. Revise procedure 5 5510, Maintenance of EFW Air Operated Trip Valves, to provide sufficient guidance to ensme valves covered by this procedure are properly assembled following maintenance. 10. The valve stem and valve body of EFW A-338 and 339 should be marked in a manner similar to EFW-A 340 to provide external indication of disk position, and flow orientation. 11. - Revise procedure 5 5510 to identify the correct gaskets to be used in the upstream and downstream flanges. 12. Either 1) Add specific instructions to procedure 611 1, Ia.strurnentation and Controls Valve Calibration and Checks, to address the details necessary to ensure of proper attachment of the operators for EFW-A 330, 339, and 340 or; 2) In the future, - include specific instructions for attachirig the operator in work order technical l instructions. h 13. Maintenance department management should continue to stress the importance of good EXH! BIT. 3 . CASE NO. 1-96-04 0 5 PAGE /[- OF_/_Y/_PAGE(S 3

turnovers and coordination between different crafts working on the same job. Reladve task performance observed during work on EFW-A-338 / EFW-A-340 would C. provide a useful comparative example. 1 C s { 35mo.- 1 f 9 6 - 0 4 0 ' EXHIBIT _ [ 6 PAGEM OF.lf/_PAGE(S) ~ q .n

h 4.0 - NARRATIVE { , [- On August 4,1994 Maine Yankee was in a cold shutdown condition making preparations to restart the plant following a maintenance outage to correct Steam Generator tube leakage. At approximately 1220 while performing a leak test of the Emergency Feedwater Isolation Valve (ISV) for #1 Steam Generator using normal system instrumentation, it was determined that - l EFW A-338 leaked by at a rate of approximately 400 GPM. Imakage past #1 Steam Genemtor Emergency Feedwater Regulating Valve EFW-A 101 at 75 GPM was also identified. Subsequent investigation determined that the actuator for EFW-A-338 was-coupled to the disk approalmately 180 degrees out of alignment. His misalignment is believed to have occurred during valve maintenance performed during our 1992 Refueling Outage. Maintenance activity was initiated to reposition the actuator / disk to the correct - orientation. At 1700 on August 5,1994 it.was determined that Maine Yankee's Steam Line 3 cak Safety Analysis assumes zero leakage past Emergency Feedwater isolation and regulating valves. Therefore, at 1957 the NRC was appraised of the situation via the Emergency Notification 3 System in accordance with the provisions of 10 CFR 50.72 (b)(2)i. A follow up notification was made at 1151 on August 6,1994 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW-A-340. 10,-1994 comprehensive testing was performed on the Emergency Feedwater On August ! solation and regulating valves for all three Steam Generators using precision test equipment. , (- These tests revealed leakrates for various individual valves and combinations of valve O to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studies conducted in parallel witn these tests established the following revised acceptance criteria: Leakage put individual isolation / regulating valves should be less than 40 GPM. The - 1) purpose of this limit is to prevent excessive couldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating valve to close, Leakage past both the isolation and regulatmg valve to each Steam Generator should i 2) be less than 10 GPM. He purpose of this limit is to provide at least 30 minutes for operators to insure EFW is isolated to a faulted Steam Generator following a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to ovr611. 3) Zero leakage to a Steam Generator with a' tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due to overfill following a Steam Generator Tube Rupture. L - In order to meet the revised acceptance criteria, EFW-A-101 was adjusted to reduce leakage past EFW-A-101 & 338 to zero GPM and administrative controls were implemented to ensure termination of any potential future leakage to a ruptured Steam Generator within 30 minutes using manual isolation valves. psERD. 1 040 EXHIBIT 3 PAGE 3/ OF8/__PAGE(S) 7

5.0 CONCLUSION

J, FACTS & RECOMMENDATIONS Conclusion At 7he EDCR process used try Maine Yankee circa 1983-84 did not provide suBictent guidance to ensure all safety sigrupcant design functions of EFW-A-338, 339 & 340 wre verfied post installation. Facts: '$e valves installed under EDCR 83-29 (EFW-A-338,339, 1. A basic design fu r 340) is to ensurr ' EFW flow to faulted SG(s) under certain accident conditions. 2. Procedure 17, ~,;ina.6 Design Change Request, requires verification of design functions by. + *

sting unless such testing is not practical.

3. It is practical to perform -

sting and stroke timing with the EFW pumps in operation.'

4. EDCR 83-29 did not require either leak testing nor valve stroke timing with the EFW pumps in operation. 5. The post installation functional testing performed on EFW-A-338,339 & 340 did not _( include either leak testing nor valve stroke timing with the EFW pumps in operation. 6. A review of 4 rece.; EDCRs was perforrned by PED. No instances ofinadequate functional testing were discovered. 7. The current EDCR process contains more guidance for selecting appropriate functional test requirements than was provided 1983-1984. Reconunendations; 1. The review of recent EDCRs performed by PED provides an indication that the functional testing specified in recent EDCRs may be adequate. Hewever, due to time constraints the review performed was extremely limited in scope. In addition, utilization PED personnel to perform the review made it difficult to ensure the It is recommended that Let review of a i evaluation was wmMy objective. ,s.tative sample of EDCRs be performed to more accuratly assess the overall adequacy of EDCR functional testing. CASENO. 1-96-040 EXHIBIT 3 PAGE W OFMPAGE(S)

4 Coackasion B: The process used by the ISTprogram to detenninefuncticnal test requirements does not ensure that the ability of EFW-A-338, 339 & 340 (.. to isolate Finergency Feedmuerflow is periodically wn)ed. Facts: Maine Yankee's IST program defines Category A valves as, " valves for which seat 1. leakage is limited to a specific maximum amount in the closed position for fulfillment - of their safety function." 2. Maine Yankee's IST program defines Category B valves as, " valves f9r which seat leakage in the closed position is inconsequential for fulfillment of their safety function." 3. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EFW ! solation valves. '4. Maine Yankee's IST program classifies EFW-A-338, 339 & 340 as Category B valves. 5. The definition for Category A valves is interpreted within the Performance Engineering Group as applying only to valves for which a specific maximum allowable leak rate (e.g., less than 20 GPM) is specified in the FSAR, Technical Specifications, the IASD, or a Design Basis Summary Document. 6. The interpretation presented in item 5 is informal and not documented within the IST program description, or memorandum. Recomunendations: l. Reclassify EFW-A-338, 339 & 340 as IST Category A valves. 2. Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. 3.- If recommendation 2 is adopted, the IST program should be reviewed to ensure valves which meet the revised criteria are included in the program and properly classified. 3. To ensure consistent classification, and valid audit results, IST program description should be_modifi.ed to include the precise criteria used to classify valves as Category A, B & C. h CASENO. 9 6 - 04 0 EXHIBIT 3 9 PAGE MOFMPAGE(S) s w .,,,ry, -m_. ..sm, ,w.y- -, -. _ --m,- a

ConclusioC C: -The Work Orderprocess did not ensure that F11sfor EFW.A 338 & C 340 tested the ability of these wives to perform their safetyfunction, nor wnfed the efectiwness of repair activities. Facts: 1. Procedurn 0-16-3,'_ Work Order Process, requires that post maintenance Functional Test Instructions (FTI), "_ Verify the ability of a compceent or system to perform its intended function." and " Demonstrate that the original deficiency has been corrected.", 2. In 1992 W.O. 92-1746 was initiated to correct EFW-A-333 valve leakage. The FTI associated with this Work Order did not verify the ability of the valve to isolate flow. 3. ' The FTIs associated with W.O. 94-3007 (EFW-A-338 valve leakage), did not verify the ability of the valve to isolate flow. 4. 'Ihe FTIs associated with W.O. 94-3075 (EFW-A-340 valve leakage), did not vering the ability of the valve to isolate flow. 5. 'Ihe FTIs associated with W.O. 92-1746,93-3007, and 93-3075 mimicked IST requirements for the associated valves. 6. A review of 30 recent Work Orders was performed by PED. No instances of inadequate functional testing were identified. Recammendations: 1. The review of recent Work Orders performed by PED pmvides an indication that the functional testing specified in recent W.O. may be adequate. However, the review performed was very limited in scope when compared to the number of W.O. generated annually. In addition, at least three recent W.O.s, 92-1746, 94-3007, and 94-3075 provided inadequate functional testing. As with the EDCR review, utilization of PED personnel to perform the review made it difficult to ensure tlw - evaluation was completely objective. 'Iherefore, it is recommended that an independent review be performed of a representative sample of W.O.s to more accuratly assess overall adequacy of W.O. functional testin.g. 2. Individuals responsible for specifying, reviewing, and approving W.O. functional test - requirements should be sensitized to the fact that specifying the component testini;; required by the IST pmgram does not ensure adequate post maintenance functional testing. _tAsttio. 1-96-040 PAGEdY _OF[PAGE(S 10

Conclusion D: Procedure 5-5510 and Work Orders 921746, 94-3008, 94-3025 did not provide adequate intructions regarding orientation of valw body to CL flow, nor orientation of vain actuator to disc shaft. Facts: l. Work Order 92-1746 for EFW-A-338 was not revised to reflect removal and reinstallation of valve to correct body orientation and wrong gaskets. This includes the actuator. 2. The "as found" valve disc / seat body position on W.O. 92-1746 was correct for EFW-A 338. 3. No procedure 6-11-1 nor other work order for I&C removal / reinstallation of the actuator can be located for W.O. 92-1746, 4. The "as found" condition for EFW-A-338 on Aug.1994 found the disc was 180' out from the correct orientation. 5. The valve seat / disc is placed in the closed position prior to, assembly in the piping. 6. EFW-A-338, 339, and 340 are fail open actuators. 7. EFW-A-338, 339, and 340 had u pe,nanent external markings to indicate valve disc position (during repair activity for EFW-A-340, W.O. 94-3025 disc shaft was marked). Recoenmendationn Procedure 5 55-10 is currently being revised to address this conclusion. Valves EFW-A-338 and 339 should be marked similar to EFW-A-340 to provide extemal indication of disk position and flow orentation. { CM. 1-96-040 PAGE II OF/$/._.PAGE 11 _ ~

1 I i i The correct gasket orientation was not idennfied in Procedure 5 55-10, Conclusion E:. nor in the wrk onter instructions for 92-1746, N-3008, and 94-3025. .C, Facts: 1. Work Order 92-1746 was not revised to reflect removal and reinstallation of valve tc correct body orientation and the use of incorrect gaskets during the intitial attempt to reassemble the valve. 2. Work on EFW-A-338 under Work Order 92-1746 installed 3" gaskets for both flanges. A 2%" gasket is required for the seat, or downstream flange (this fact is based on interview results). 3. W.O. 94-3025 for EFW-A-340 installed gaskets in reverse orientation, causing the valve to leak externally. 4 EFW-A 338,339 & 340 failed functional testing under EDCR 83-29 due to leaking gaskets. 5. As a result of Work Order 92-1746 for EFW-A 338, DCR 92-170 was generated to correct the vendor drawing 7.69-46 and change Mfr. Inst. Manual C-9-2 to reflect gasket orientation. (.. 6. A copy of the vendor drawing 7.69-46 could not be located in the 92-3025 work package for EFW-A-340. Reconunendations: Procedure 5-55-10 is currently being revised to address this conclusion. CASE NO. 1 - 9o- 040.. C. 3 EXHlBIT PAGE.4 OF4PAGE(S) 12-

i . Coactusboa F:- Procedure 6111 provides inadequate guidance to ensure consistent proper reassembly of EFW-A 338, 339, and 340. Facts' This conclusion is based on interviews with several 1&C technicians and supervision coupled-with the following facts: 1. For identical work:- Step 5.2.1 of Procedure 6-11-1 was N/A'd for valve EFW-A-338 under W.O. o' 94-3008, whereas signed completed for EFW-A-340 under W.O. 94-3025. o Step 5.3.1 wa signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-l 3008. I Step 5.7.1 was N/A'd completely on W.O. 94-3008, but one sub step signed o as, completed on 94-3025. Step 6.4 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-o 3008. 2. No completed copy of 611-1 could be found for W.O. 92-1746 (EFW-A-338 actuator temoval/ reinstallation). 3. He valve actuator for EFW-A-340 was reinstalled twice under W.O. 94-3025 'Ihere I was only one complete 6-11-1 procedure. 4, The required documentation for Step 5.2.2.b of Procedure 611-1 is missing from W.O. 94-3025. 5. Step 5.2.2.a of Procedure 611-1 for W.O. 94-3025 was signed off as complete. However, there are no manufacturer's instruction manual position marks to verify. i Recoaumeadatloas: l. Procedure. 6-11-1 is a combination of generic and specific instructions (i.e. main feed reg, valves). Two options could be considered: 1) add specific instructions to 6-11-1 for EFW-A-338, 339, and 340, or 2) use work order technical instructions for EFW-L A-330,339, and 340. 2. Some improvement' in mechanical maintenance and 1&C interface / communication would appear necessary. For the work on EFW-A-338, there appeared to be good - communications. In spite of insufficient instructions, the I&C technicisa and maintenance mechanic determined proper orientation and avoided the mh'*b made under W.O. 92-1746. Work on EFW-A-340 did not appear to have as good communications with both maintenance mechanics and I&C technician being confused } EXHIBIT WBEll0. 1 - 9 6 '- 0 4 0. 13. PAGE_ N OFMPAGE(S) w ..-....-,e, w<,--,--- .n~ v-

r + as to proper reassembly. _ When good communications were established, correct (; orientation was determined. Again, this was accomplished without sufficient guidance L in the work instructions. ( CASENO..1-96-040 EXHlBIT 3 PAGE M OF8/_.PAGE I4

~. = l Conclusloa G:. The diskfor EFW-A-33? was mis positioned during performance of . W.O. 92-1746 (Mar. 92; = ~ Facts: 1. 05-2 6-87. Stroke time testing performed cc EFW-A-338,339 & 340 under; conditions representative of a main steam line rupture. Stroke times within acaptance criteria. Test repo:t noted that the valves demonstrated very good shutoff (0-2 GPMf.. 2. 11-23-91 Leakage rate identified on W.O. 94-7585 was 612 GPM with EFW-A-338 shut. W.O. 92-1746 generated for repairs. . 3. 03 30-92 Completed repairs to EFW-A-338 under W.O. 92-1746 (no seat leakage test). 4 0844-94 Leakage rate identified on W.O. 94-3007 was 300+ GPM with EFW-A-338 shut. 5. 08 05-94 Completed repairs to EFW-A-338. lukage test satisfactory. 6. 08-10-94 Leak test of EFW-A-339& 340 were 17 GPM and 36 GPM respectively. 7. Review of equipment history (both mechanical and I&C) revealed'that EFW-A-338 ' ,( was the only valve to be disassembled (or have actuator removed). EFW-A-339 has not been disassembled since original installation. EFW-A-340 was disassembled under W.O. 94-3025, after the leak test performed on 08/10/94. Recomunendations: None N 1-96-040 EXHIBIT 3 PAGEdOF$PAGE(S) 15 i .-m.,-., --.c -,e ~,,.

MISCELLANEOUS OBSERVATIONS: OBSERVATION A. Design change information from EDCR 83 29 (i.e. flange gasket. orientation and welded disk key to shaft) was not incorporated into Maintenance instmetions for the valve, i.e. Pmcedure 5-55-10, nor into the instruction manual C-9-2. A second opportunity occurred in 1992 under W.O. 92-1746, DCR 92-170 changed technical manual instructions (C-9-2), and the vendor dwg. 7.69-46 to include gasket . orientation instructions. 'Ihis information was not incorporated into Maintenance instructions. RECOMMENDATION: During EDCR 83-29 implementation, the design. notification form was the process to identify changes to respective departments. To further enhance this process and include other areas, the Configuration Management Program (Procedure 046-

8) was developed. Some increased awareness of this process is recommended.

OBSERVATION B. There was considerable rework involved in completing reinstallation of EFW-A-340 under W.O. 94-3025. In summary: (- 1. The valve body was installed 180' out of position. 2. The flange gaskets were installed incorrectly. 3. The valve was torqued in the "open" position, causing binding, RECOMMENDATION: Maintenance department management should continue to stress the importance of good tumovers and coordination between different crafts working on the same job. Five mechanics and three I&C techs. covering three shifs were involved with this work activity. Relative task performance observed during work on EFW-A-338 / EFW A-340 would provide a useful 4 comparative example. - psno. 1 - 9 6 - 0 4 0,, EXHlBIT 3 PAGE 3d OF/f/ _PAG 16 ,. ~ .m ,. ~ - -..

... ~ ATTACHMENT A C Test As Found Results As left 10-Valve (sTimed 08/04/94 08/10/94 08/11/94 1 EFW-A-101 75 GPM 36 GPM 0GPM EFW-A-338 400 GPM 20 GPM 24 GPM 15 GPM 0GPM EFW-A-101 & 338 0GPM 0GPM EFW-A-101,338 & EFW-102 0GPM 0GPM 2 EFW-A-201 17 GPM 17 GPM EFW-A-339 0GPM 0GPM EFW-A 201 & 339 0GPM 0GPM EFW-A-201,339 & EFW-202 O GPM 0GPM 3 EFW-A-301 36 GPM 23 GPM EFW-A-340 - 4 GPM 0GPM EFW-A-301 & 340 0GPM 0GPM EFW-A-301,340 & EFW-302 ( t EXHIBIT f PAGE 3/ OF8_ PAGE(S)

.~ T.tLeesADtYweNamnsposTM ( 1 - 96 ~ 04 h, CASE NO-EXHlBIT $ PAGE '1V0FdPAGE(S) 18

3. C Emergency Feedwater Valve Leakage Event Investigation Report Team Members: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED ~ James Taylor - Senior Nuclear Safety Engineer, NSEG -(. Event Date: 08/05/94 Report Date: 09/20/94 Submitted by: Nuclear Safety Specialist, (Team Leader) Approved by: Manager, Plant Engineering CASEND. 1 - O 6- 04 Q EXHIBIT 3 PAGE 93 0F$PAGE(S)

1.0 CHARTER 1. Identify causal factors contributing to the Emergency Feedwater Valve Leahge event described in Licensee Event Report 94-016, Emergency Feedwater Isolation Valve Leakage. 2. Provide recommendations for reducing the probability of such an event recurring in the future. 3. Assess potential generic implications of any deficiencies noted. 4. Submit report on results of investigation within approximately one week. - [' CASENO. 9 6 - 0 4 0 EXHlB1T 3 PAggq op4PAGE(S)

A b 2.0 ' EXECUTIVE

SUMMARY

C At 1220, on August 4,1994 with the reactor in a cold shutdown condition, plant - operators determined that an Emergency Feedwater isolation valve for #1 Steam Generator was leaking by. Further investigation also identified leakage in the Emergency Feedwater (EFW) supplies to #2 & #3 Steam Generators. Subsequently it _was determined that under accident conditions which require isolation of Emergency Feedwater, isolation valve leakage could exceed Safety Analysis assumptions. Maintenance activities were initiated to reduce Emergency Feedwater valve leakage. In addition, administrative controls were implemented to ensure Emergency Fedwater leakage is maintained within the bounds of Safety Analysis assumptions during neeMent conditions. A review cf this event has revealed the following weaknesses which contributed to this event. EFW flow isolation assumptions in our Safety Analysis are not clearly I o presented in design basis reference documents. Our IST program does not require periodic verification of the ability of the o EFW isolation and feed regulating valves to isolate EFW flow. The EDCR under which the EFW isolation valves were installed did not ' (*- o require verification of the ability of these valves to isolate EFW flow as part of post installation functional testing. Post maintenance functional testing pe4 formed on EFW-A-338 in U92 and o EFW-A-101 in 1991 & 1994, did not require verification of these valves ability to isolate EFW flow, Maintenance procedures lack sufficient guidance to ensure consistent proper o assembly of the EFW isolation valves. Lack of a single, comprehensive, readily available reference for Safety Analysis inputs and assumptions, seems'to have played a key role in this event. The improved documentation and retrievability of Safety Analysis information to be offered by the Safety Analysis Inputs & Assumptions (SAID) which is currently under development, should significantly reduce the possibility of similar events once it becomes available for use. Management should continue to emphasize / support timely completion of this effort. Other significant activities which are recommended include the following: Initiation of an independent, broad scope, review of EDCRs and W.O.s to o determine the overall a&quacy of functional testing performed at Maine .( - Yankee. I titJ h1 q. g o - 0 4 0 EXHIBIT 3 PAGE M OFf/__.PAGE(S)

Review of the IST program to ensure the current scope of the program and the o criteria used to establish the functional test requirements for specific fV components are adequate. Efforts to increase awareness among those who develop, review, and approve o-functional tests, that the functional testing required by our IST program should not be relied upon to ensure adequate post maintenance functional testing. Revision of maintenance procedures to provide additional guidance for o assembling the EFW isolation valves to casure consistent proper orientation of the valve disk, body, seat, and actuator. ( CASE NO. 1-96-040 C EXHIBIT I PAGE 34 0F/f[_PAGE(S) 4

j l'm. 1 g 3.0 RECOMMENDATIONS ( 1.- Management should continue to emphasize / support timely completion of the action " items identified in the QPD, NSEG report titled

  • Maine Yankee Safety Analysis j

Inputs & Assumptions Review Project Final Report", dated July 8,1994 l - 2. Perform an' independent review of a mys.e;.atative sample of EDCRs to more - completely assess the overall adequacy of EDCR functional testing.- l 3. Reclassify EFW-A-101,201,301,338,339 & 340 as IST Category A valves. I ~ 4. - Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. 5. If recommendation 2 is adopted, the IST program should be reviewed to ensure all valves which meet the revised criteria are included in the program and properly classified. 6. ' Modify the IST program description to include the precise criteria used to classify - valves as Category A, B & C. Perform an independent review of a representative sample of W.O.s to more fully > -{- assess the overall adequacy of W.O. functional testing. 7. 8. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing required by the IST program does not -urily ensure adequate post maintenance functional testing. 9. Revise procedure 5-5510, Maintenance of EFW Air Operated Trip Valves, to: 4 Provide sufficient guidance to ensure valves covered by this procedure are o properly assembled followinF mainte.unce, i Identify the correct gaskets to be used in the upstream and downstream o flanges. 10' Ihe valve stem and valve body of EFW-A-338 and 339 should be marked in a manner similar to EFW-A-340 to provide external irwticarian of disk position, and . flow onentation.

11. -

Either 1) Add specific instructions to procedure 6-11-1, Instrumentation and Controls {- Valve Calibration and Checks,. to address the details nemary to ensure of proper attachment of the operators for EFW-A-330,339, and 340 or; 2) In the future, CASENO.' 1_.96-040- " 8IT # 5 PAGE.37 OF PAGE(S)-

...a .m, a. m s a.. .a A:, ~. 1 includ; specific instructions for attaching the operator in work order technical i instruc tions. {. 12. Maint: nance department management should continue to stress the importance of good turno"ers and coordination between different crafts working on the same job.- Rela' ave task performance observed during work on EFW-A-338 / EFW-A-340 would provide a useful comparative example. tMENO-1-E(040 ( EXHIBIT 8 PAGE.J$_OFK PAGE(S) 6

4.0 ' NARRATIVE. C-On August 4. '1994 Maine Yankee wu in a cold shutdown condition making prepamtions to restart the plant following a maintenance outage to correct Steam Generator tube leakage. At . approximately 1220 while performing a leak test of the Emergency Feedwater Isolation Valve (ISV) for #1 Steam Generator using normal system instrumentation, it was determined that EFW A-338 leaked by at a rate in exass of 300 GPM. Leakage past #1 Steam Generator. Emergency Feedwater Regulating. Valve EFW A-101 at 75 GPM was also identified. Subsequent investigation determined that the actuator for EFW-A-338 was coupled to the disk approximately 180 As out of alignment. His misalignment is believed to have occurred w during valve maintenana performed during our 1992 Refueling Outage. Maintenance activity was initiated to reposition the actuator / disk to the correct orientation. At 1700 on August 5,1994 it was determined that Maine Yankee's Steam Line Break Safety Analysis assumes zero leakage past Emergency Feedwater isolation and regulating valves. Derefore, at 1957 the NRC was appraised of the situation via the Emergency Notification System in accorsiance with the provisions of 10 CFR 50.72 (b)(2)i. A follow up notification . was made at 1151 on August 6,1991 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW-A-340. i On August 10,1994 comprehensive testing was performed on the Emergency Feedwater isolation and regulating valves for all three Steam Generators using PED test equipment. C These tests revealed leak rates for various individual valves and combinations of valves from -J O to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studies conducted in parallel with these tests established the following revised acceptance criteria: 1) Ilakage past individual isolation / regulating valves should be less than 40 GPM. The purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating valve to close. 2) Leakage past both the isolation and regulating valve to each Steam Generator. should be less than 10 GPM. De purpose of this limit is to provide at least 30 minutes for operators to insure EFW is isolated to a faulted Steam Generator following a Steam - Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to overfill. 3) Zero leakage to a Steam Generator with a tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due to overfill following a Steam Generator Tube Rupture. In order to meet the revised Snce criteria, EFW-A-101 was adjusted to reduce leakage past EFW-A-101 & 338 to zero GPM and administrative controls were implemented to ensure termination of any potential future leakage to a ruptured Steam Generator within 30

minutes using manual isolation valves.

CASEND. 1-9G-040 EXHIBIT 3 7 PAGE 57 0F/f/__PAGE(S) .-.y&,- .wy, --m,,, .-4..

5.0 _ _ CONCLUSIONS, FACTS & RECOMMENL,$TIONS { Conclusion At The process used by Maine Yankee to ensure consistency betwen our Safety Analysis and plant perfonnance did not ensure that; o The safetyfunctions of EFW A 338, 339 & 340 wre adequately wnfed during post installation testing. The safetyfunction of EFW-A-101, 201, 301, 338, 339 & 340 to o isolateflow wre periodically wnfed under the ISTprogram. The safetyfunctions of EFW-A-101, 338 & 340 were adequately o test during post maintenance tes:ing. Facts: A basied' sign function of the valves installed under EDCR 83-29 (EFW-A-338,339, 1. e - 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident conditions. 2. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EFW isolation / feed regulating valves. 3. Procedure 17 21-2, Engineering Design Change Request, requires verification of design functions by functional testing unless such testing is not practical. 4. It is practical to perform leak testing and stroke timing with the EFW pumps in operation. 5. EDCR 83-29 did not require either leak '.esting nor valve stroke timing with the EFW pumps in operation. 6. The post installation functional testing performed on EFW-A-338,339 & 340 did not include either leak testing nor valve stroke timing with the EFW pumps in operation. 7. Maine Yankee's IST program classifies EFW-A-101, 201, 301, 338, 339 & 340 as Category B valves. 8. Maine Yankee's IST program defines Category B valves as, " valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety function." 9. In' 1992 W.O. 92-1746 was initiated to correct EFW-A-338 valve leakage. The FTl associated with this Work Order did not verify the ability of the valve to isolate flow. 10. The FT!s associated with W.O. 94-3007 (EFW-A-338 valve leakage), did not verify y

p. 9 G - O' 7 2

EEC 3 PAGE h OF /f[PAGE(S)

the ability of the valve to isolate flow. 11. 'Ihe FTis associated with W.O. 94-3075 (EFW-A-340 valve leakage), did not verify the ability of the valve to isolate flow. I 12. 'the FTIs associated with W.O. 91-7584 (EFW-A 101 ve've leakage), did not verify - the ability of the valve to isolate flow. 13. The FTIs==Ma'M with W.O. 94-1997 (EFW-A-101 not completely shut), did not verify the ability of the valve to isolate flow. Recommendations: 1. Management should continue to emphasizelsupport the timely completion of the action items identified in the QPD, NSEG report titled " Maine Yankee Safety Analysis Inputs & Assumptions Review Project Final Report", dated July 8,1994 CASEHO. 1 96-040 EXHlBiT 3 PAGE 9/ OF$PAGE(S) 9

Cwu Et The EDCR process used by Maine Yankee circa 1983-84 did not {. prodde suffelent guidance to ensure all sqfety signflicant design ? pnctions of EFW-A 338, 339 & 340 wie wrified post installation. Facts: 1. A basic design function of the valves installed under EDCR 83-29 (EFW-A-338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident conditions. 2. Procedure 17-21-2, Engineering Design Change Request, requires verification of design functions by functional testing unless such testing is not practical. 3. It is practical to perform leak testing and stroke timing with the EFW pumps in operation. EDCR 83-29 did not require either leak testing nor valve stroke timing with the EFW s pumps in operation. 5. The post installation functional testing performed on EFW-A-338, 339 & 340 did not include either leak testing nor valve stroke timing with the EFW pumps in operation. 6. A review of 4 recent EDCRs was performed by PED. No instances of inadequate { functional testing were discovered. 7. The current EDCR process contains more guidance for selecting appropriate functional test requirements than was provided 1983-1984 Recommends'.lons: 1. The review of recent EDCRs performed by PED provides an indication that the functional testing rpecified in recent EDCRs may be adequate. However, due to time constraints the review performed was extremely limited in scope. In addition, utilization of PED personnel to perform the review made it difficult to ensure the evaluation was completely objective. It is recommended that independent review of a representative sample of EDCRs be performed to more accurately assess the overall adequacy of EDCR functional testing. CASE ND. 1 - 9 G - 0 4 0,- EXHIBIT 3 PAGE 9).- OFjff_.PAGE(S)

~. - 4, Coachesion Ct. The process used by the ISTprogram to determinefunctional test (,- regulismens does not ensure that the ability of EFW-A-101, 201, 301, 338, 339 & 340 to isolate Emergency Feedwaterflow is periodically wriffed. Facts: - 1. Maine Yankee's IST program defines Category A valves as, " valves for which seat - leakage is limited to a specific maximum amount in the closed position for fulfillment of their safety function." 2. Maine Yankee's IST program defines Category B valves as, " valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety. function." 3. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EFW isolation / feed regulating valves. 4 Maine Yankee's IST program classifies EFW-A 101, 201, 301, 338, 339 & 340 as Category B valves. 5. The definition for Category A valves is interpreted within the Performance Engineering Group as applying only to valves for which a specific maximum allowable leak rate (e.g., less than 20 GPM) is specified in the FSAR, Technical ( Specifications, the IASD, or a Design Basis Summary Document. 6. The interpretation presented in item 5 is informal and not documented within the IST program description, or memorandum. Recommendations: 1. Reclassify EFW-A-101,201,301,338,339 & 340 as IST Category A valves. 2. Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. 3. If recommendation 2 is adopted, the IST program should be reviewed to ensure valves which meet the revised criteria are included in the program and properly classified. 4 To ensure consistent classification, and valid audit results, IST program description should be modified to include the precise criteria used to classify valves as Category A, B & C. CASE NO. 1-96-040 XHIBIT 8 e 11 PAGE 93 0F/[/_PAGE(S

J 1 C==*t==ta= Ds-7he Wort Onter process did not ensure that F77 sJbr EFW-A 101, 338 a (* & 340 tested the aNiity qf these veins to perfbrm their sqfetyJimetion, nor wriffed the dectinness of repair activities. Facts- ' l. Procedurt 0516 3, Work Order Process, requires that post maintenance Functional . Test Instructions (FTI), " Verify the ability of a component or system to_ perform its- ) intsaded function." and " Demonstrate that the original deficiency has been corra:ted." j 2. Ir 1992 W.O. 921746 was initiated to correct EFW-A-338 valve leakage. De FTl associated with this Work Order did not verify the ability of the valve to isolate flow. 3. The. Fris associated with W.O. 94-3007 (EFW-A-338 valve leakage), did not verify' the ability of the valve to isolate flow. '4 The FT1s associated with W.O. 94-3025 (EFW-A 340 valve leakage), did not verify the ability of the valve to isolate flow. 5. The FTIs associated with W.O. 91-7584 (EFW-A 101 valve leakage), did not verify : the ability of the valve to isolate flow. 6. De FTIs associated with W.O. 94-1997 (EFW-A-101 not completely shut), did not verify the ability of the valve to isolate flow. -(< 7. De FT!s associated with W.O. 91-7584,94-1746,94-1997,94-3007, and 94-3025 basically mimicked IST requirements for the associated valves, 8. A review of 30 recent Work Orders was performed by PED. No instances of a inadequate functional testing were identified. Reconumendations: 1. De review of recent Work Orders performed by PED provides an indication that the functional testing specified in recent W.O. may be adequate. However, the review performed was very limited in scope when compared to the number of W.O. generated annually. In addition, at least four recent W.O.s, 92-1746, 94-1997, 94-3007, and 94-3025 provided inadequate functional testing. As with the EDCR P review, utilization of PED personnel to perform the review made it difficult to ensure - the evaluation was completely objective. Therefore, it is recommended that an in-f+;-=-f=,t review be performed of a representative sample of W.O.s to more ^ accurately assess overall adequacy of W.O. functional testing. l -2.. I'ndividuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing- . required by the IST program does not ensure adequate post maintenance functional testing. EXHIBIT PAGE ff 0F8/ PAGE(! p i 7 p" 12 1 f $-. ..w,

Conclusion E: Procedure 5 3510, Maintenance of EFW Air Operated Trip Valus, ( m.d Work Orders 92-1746, 94-3008 & 94-3025 did not provide adlyuate instructions regarding orientation of valw body topow, nor . orientat'on of vain actuator to disc shap. Facts: 1. Work Order 92-1746 for EFW-A-338 was not revised to reflect removal and reinstallation of valve to correct body orientation and wrong gaskets. This includes the actuator. 2. The "as found" valve disc / seat body position on W.O. 92-1746 was correct for EFW-A-338, 3. No procedure 6-11-1 nor other work order for 1&C removal / reinstallation of the actuator can be located for W.O. 97-1746. 4. The "as fo'und" condition for EFW-A-308 on Aug.1994 found the disc was 180' out from the correct osientation. 5. The valve seat / disc is placed in the closed position prior to assembly in the piping. (.. 6. EFW-A-338,339, and 340 are fail open actuators. 7. EFW-A-338,339, and 340 had no permanent external markings to indicate valve disc position (during repair activity for EFW-A-340, W.O. 94 3025 disc shaft was marked). Recommendations: Procedure 5 55-10 is currently being revised to address this conclusion. Valves EFW-A-338 and 339 should be marked similar to EFW-A-340 to provide extemal indication of disk position and flow orientation. ( Cf.SE NO. } 9 q - () 4 0 EXHIBIT E PAGE_f5I OF4 PAG 13

Conckaaion F: The correct gasket orientation ws not idenn) fed in Procedure 5 55-10, Maintenance of EFW Air Operated Trip Valus, nor in the work onier instructionsfor 92-1746, N 3008, and N-3025. Facts: 1. Work Order 921746 was not revised to reflect removal and reinstallation of valve to correct body orientation and the use of incorrect gaskets during the initial attempt to reassemble the valve. 2. Work on EFW-A-338 under Work Order 92-1746 installed 3" gaskets for both flanges. A 2%* gasket is required for the seat, or downstream flange (this fact is based on interview results). 3. W.O. 94-3025 for EFW-A-340 imtalled gaskets in reverse orientation, causing se i valve to leak externally. 4 EFW-A-338,339 & 340 failed functional testing under EDCR 83-29 due to leaking gaskets. 5. As a result of Work Order 92-1746 for EFW-A-338, DCR 92-170 was generated to correct the vendor drawing 7.69-46 and change Mfr. Inst. Manual C-9-2 to reflect gasket orientation. ( 6. A copy of the vendor drawing 7.69-46 could not be located in the 94-3025 work package for EEW-A-340. Recommendations: Procedure 5-55-10 is currently being revised to address this conclusion. C-CASE NO. 1 - 9 C. - 0 4 r EXHIBIT I PAGE Id' OF/f/__ PAG 14

Camekata= Gt Procedure 6-11-1, instnenentation and Controls Valve Calibration and Owcks, provides inadequate guidance to ensure conshtent proper reassernbly of EFW A-338, 339, and 340. Facts: 1. Based on information obtained during interviews, I&C technicians were uncertain as to the proper application of procedure 6-11-1 to the EFW valves. Guidance concerning checking the seat / disk for proper orientation was considered particularly confusing. 2. For identica$ work: Step 5.2.1 of Procedure 6-11-1 was N/A'd for valve EFW-A-338 under W.O. o 94 3008, whereas signed completed for EFW-A-340 under W.O. 94-3025. o Step 5.3.1 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-3008. Step 5.7.1 was N/A'd completely on W.O. 94-300p, but one sub step signed o as completed on 94-3025. Step 6.4 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-o 3008. 3. The valve actuator for EFW-A-340 was reinstal'ed twice under W.O. 94-3025. There was only one complete 611-1 procedure. 4 The required documentation for Step 5.2.2.b of Procedure 6-11-1 is missing from W.O. 94-3025. 5. Step 5.2.2.a of Procedure 6-11-1 for W.O. 94-3025 was signed off as complete. However, there are no manufacturer's position marks identified in the vendor manual / drawings to verify. 6. No completed copy of 6-11-I could be found for W.O. 92-1746 (EFW-A-338 actuator removal / reinstallation). Recommendations: 1. Procedure 6-11-1 is a combination of generic and specific instructions (i.e. main feed reg. valves). Two options could be considered: 1) add specific instructions to 6-11-1 for EFW-A-338,339, and 340, or 2) use work order technical instructions for EFW-A-330,339, and 340 2. Some improvement in mechanical maintenance and I&C interface / communication EXHlBli 8 15 l*SE n 1 - 9 G - 0 4 0 PAGE 97 0Fj[/_PAGE(f-)

would appear necessary. For the work on EFW A 338, there appeared to be good comm=W. In spite of insufficient instructions, the IAC technician and C' maintenance mechanic determined proper orientation and avoided the mistake made under W.O. 921740 Work on EFW A 340 did not appear to have as good communications with both maintenance mechanics and I&C technician being confused as to prtter reassembly. When good communications were established, correct orientadon was determined. Again, this was accomplished without sufficient guidance in the work instructions. J i i CASE NO.

1. O n7 04 h EXHlBlT f FAGE N CF/ff._PAG 16

.= Ceedesloe Ht The diskfor EFW A 338 uns mis postrioned during performance of (' W.O. 921746 (Mar. 92) i Facts: 1. 05 26 87 Stroke time tesdng performed on EFW-A 338,339 & 340 under conditions representative of a main steam line rupture. Stroke times within acceptance criter'.a. Test report noted that the valves demonstrated, "very good shutoff' (0-2 GPM). 2. 11-23 91 Imkage rate identified on W.O. 917585 was 612 GPM with EFW-A-338 shut. W.O. 921746 generated for repairs. 3. 03-30-92 Completed repairs to EFW A 338 under W.O. 921746 (no seat leakage test). 4. 08 04 94, Leakage rate identified on W.O. 94 3007 was 300+ GPM with EFW-A 338 sut. 5. 08-05 94 Completed repairs to EFW A 338. Leakage test satisfactory. 6. 38-10-94 Leak test of EFW A 339& 340 weit 17 GPM and 36 GPM respectively. ( 7. Review of equipment history (both mechanical and I&C) revealed that EFW. A 338 was the only valve to be disassembled (or have actuator removed). EFW A-339 has not been disassembled since original installation. EFW-A-340 was disassembled under W.O. 94 3025, after the leak test performed on 08/10/94. Reccenmendations: None CASE NO. 1 0C 0 4 b' EXHIBIT Y PAGE @ 0F/ff PAGE(S 17

MISCELLANEOUS OBSERVATIONS: OBSERVATION /.. Design change information from EDCR 83 29 (i.e. flange guket orientation and welded disk key to shaft) was not incorporated bio Maintenance instructions for the valve, i.e. Proceoure 5-5510, nor into the instruction manual C 9 2. A second opportanity occurred in 1992 under W.O. 921746, DCR 92170 changed technical manual instructions (C-9 2), and the vendor dwg. 7.69-46 to include gasket orientation instructions. His information was not incorporated into Maintenance instructions. RECOMMENDATION: During EDCR 83 29 implementation, the design notification form was the process to ;dentify changes to respective departments. To further enhance this process and include other areas, the Configuration Management Program (Procedure 0 8) was developed. Some increased awareness of this prxess is recommended. OBSERVATION B. There was considerable rework involved in completing reinstallation of EFW A 340 under W.O. 94 3025. In summary: l. The valve body was installed 180' out of position. ( 2. The flange gaskets were installed incorrectly. 3. The valve was torqued in the "open" position, causing binding. RECOMMENDATION: Maintenance department management should continue to stress the importance of good turnovers and coordination between different crafts working on the same job. Five mechanics and three I&C techs, covering three shifts were involved with this work activity. Relative task performance observed during work on EFW-A 338 / EFW A-340 would provide a useful comparative example. WE ND. ]. 9 0. p.; 9 { EXHGIT-18 PAGE_ff_._OFM/ GEM i

l ATTACHMENT A C WW44BB M44M "E p.34

>0

-*- M 11 arw4 ass ps4ari e me -C4-- M E1 P HC -C4-- N SA ( Test As Found Results As Uft SD Valve (s) Closed 08/04/94 Q3/10/94 08/11/94 1 EFW-A-101 75 GPM 36 GPM 0GPM EFW A-338 400 GPM 20 GPM 24 GPM 15 GPM 0GPM EFW-A 101 & 338 0GPM 0GPM EFW-A 101,338 & EFW 102 0GPM 0GPM 2 EFW-A-201 17 GPM 17 GPM EFW A-339 0 GPM 0GPM EFW A 201 & 339 0GPM 0GPM EFW-A-201,339 & EFW 202 OGPM 0GPM 3 EFW A 301 36 GPM 23 GPM EFW-A 340 ' 4 GPM 0GPM EFW-A 301 & 340 0GPM 0GPM EFW A-301, 340 & EFW 302 ( CASE NU. 1-96*O'10 E EXHIBIT 19 PAGE f/ OF./f/_ PAGE(!

bw...-- 4 C .C-cut No. 1 - 9 G - 04 0 5 EXHIBIT PAGEMOF84._ PAGE(

9, ~ Emerger::y Feedwater Valve tankage Evet Investigation Report Team Members: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED James Taylor - Senior Nuclear Safety Engineer, NSEG ( - Event Date: 08/05/94 Report Date: 09/20/94 Submitted by: Nuclear Safety Specialist, (Team Leader) Approved by: Manager, Plant Engineering CASE NO. ] 9 (; - 0 4 () { [ EXHIBIT PAGE IJ OF.///_PAGE

1.0 CHARTER (' l. Identify causal facsors contribudng 2 the Faergency Feedwater Valve Lankage event described in Ucensee Event Report 94-016, Emergency Peedwater Ixnation Valve 14akage. 2. Provide recommendations for reducing the probability of such an event recurring in the future. 3. Assess patential generic implications of any de% noted. 4 Submit report on results of investigation within approximately one week. e CASENO. 1-96-040 EXHIBIT 8 ( PAGE 6f OF./f4PAGE(5

i l 2.0 EXBCUUVE

SUMMARY

At 1220, on August 4.1994 with the renuar in a cow thutdown condidon, plant opersion desermined that an Emergency readwaier isoladon valve for #1 steam Generator was leaking by. Further investigaden also idendnad leakage in the t Emergency Pendwater (EFW) supplies to #2 & #3 Steam Generaton. Subsequently k was deterudnad that under accident condidons which seguire isolation of Emergency l Feedwater, isolados valve leakage could escoed safety Analyds assumpdons. Maintenance activities were initiated to reduce Fmergency Pendwater valve leakage. In addition, administrative controls were implemented to ensure Emergency Pendwater lealmge is maintained within the bounds of Safety Analysis assumpdons l during accident oceditions. A review of this event has revealed the following weaknesses which contributed to this event.- EFW How isolation assumptions in our Safety Analysis are not clearly o pfenented in design basit referaw documents. Our IST program does not require periodic veri 5 cation of the ability of the o EFW isolation and feed reguladng valves to isolate EFW Gow. 'the EDCR under which the EFW isolation valves were installed did not o require verincation of the ability of these valves to isolate EFW Dow as part l (- of post installation functional tesdag. l Post maintenance functional testing performed on EFW A 338 in 1992 and o EFW A 101 in 1991 & 1994, did not require verification of these valves ability to isolate EFW flow. Maintenance procedures lack sufficient guidance to ensure consistent proper o assembly of the EFW isolation valves. l Lack of a single, comprehensive, readily available reference for Safety Analysis inputs and assumpdons, seems to have played a key role in this event. The improved documentation and retrievability of Safety Analysis information to be offered by the Safety Analysis inputs & Assumptions (SAID) which is currently under development, should significantly reduce the possibility of similar events once it becomes available for use. Manat,ement should continue to emphasize / support timely completion of this effort. i Other significant activities which are recommended include the following: l Initiation of an independent, broad scope, resiew of EDCRs and W.O.s to o determine the overall adequacy of fimetional testing performed at Maine EXHlBlT 8 PAGE_If 0FB/_.PAGE(S) - D 1-O 6' O4 0 3

4 Review of the IST program to ensure the current scope of the program and the o criteria used to establish the funedonal test requirements for specific C-components are adequate, Efforts to increase awareness among those who develop, review, and approve i o funedonal tests, that the funedonal tesdag required by our IST program should not bt'i relied upon to ensure adequate post maintenance functional testing. Revision of maintenance procedures to provide addidonal guidance for o assembling the EFW isoladon valves to ensure consistent proper orientation of the valve disk, body, seat, and actuator. I ( EXHIBIT 7 CASEND. ] ,96-01P og,c $4 0F/f/_.PAGE 4 (

3.0 RECOh04ENDA*!10NS l l. Managemer't should continue to =pha=Imalsupport dmely compledon of the action items identifled in the QPD, NSEO report titled

  • Maine Yankee Safety Analysis inputs & Assumpdons Review Project Final Report", dated July 8,1994.

i 2. Perform an ir2+,i^ review of a representative sample of EDCRs to more completely assess the overall adequacy of EDCR functional tesdng. 3. Reclassify EFW A 101,201,301,338,339 & 340 as IST Category A valves. 4. Consideradon should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. 1 5. If recommendation 2 is adopted, the IST program should be reviewed to ensure all l valves which meet the revised criteria are included in the program and properly classifiet!. 6. Modify the IST program description to include the precise criteria used to classify valves as Category A, B & C. 1 l . 7. Perform an independent review of a representative sample of W.O.s to more fully (. assess the overall adequacy of W.O. functional testing. 8. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing required by the IST prog am does not necessarily ensure adequate post maintenance functional testing. 9. Revise proce.iure 5 5510, Maintenance of EFW Air Operated Trip Valves, to: o Provide sufficient gvWw to ensure valves covered by this procedure are properly assembled following maintenance, Identify the correct gaskets to be used in the upstream and downstream o flanges.

10. -

The valve stem and valve body of EFW A-338 and 339 should be marked in a - manner similar to EFW A-340 to provide external indication of disk position, and flow orientation. I1. Either 1) Add specific instructions to procedure 6-11-1, Instrumentation and Controls Valve Calibration and Checks, to address the details necessary to ensure of proper ( attachment of the operators for EFW A-330, 339, and 340 or; 2) In the future. EXHIBlT 3 5 CASE NO, ].90 040 PAGEM_OF.8/_ PAGE(S)

include specific instructions for attaching the operator in work order techrdeal instructions. { 12. Malam dopamnant management should continue to senas the importance of good tumovers and coordinarian between different crafts working on the same job. Relative task performance observed during work on EFW A-338 / EFW-A 340 would prwide a useful comparative exampk. I i i l 4 ) EXHIBIT - NE 7 -9I'-040 PAGE 67 0FhPAGE(S) 6 r- --+i--yweg -gs-- &y e WemwrrT--ver-w~-wr- --m'-wreWW -r ei-4

  • 'e

'--w w w-9--w m-re' N f a'-+- T-

{ 4.0 NARRATIVE On August 4,1994 Maine Yankee was in a cold shutdown condition making preparations to restart the plant following a snaintenance outage to correct Steam Generator tube leakare. At approximmealy 1220 whi's performing a leak test of the Emergency Feedwater Isolation Valve (ISV) for #1 Steam Generator using nonnat system instrumentadon, it was determined that i EFW A 338 leaked by at a rate in excess of 300 GPM. I t past #1 Steam Generator Emergency Feedwater Regulating Valve EFW-A 101 at 75 GPM was also identified. i Subsequent l'ivestigation determined that the actuator for EFW A 338 was coupled to the disk approximately 180 degrees out of alignment. 'this misalignment is believed to have occurred during valve maintenance performed during our 1992 Refueling Outage. Maintenance i activity was initiated to reposition the actuator / disk to the correct orientadon. { At 1700 on August 5, D94 it was determined that Maine Yankee's Steam Line Break Safety Analysis assumes aero leakage past Eir g.g Feedwater isoladon and regulating valves. [ Therefore, at 1957 the NRC was appraised of the situation via the Emergency Notif. cation System in accordance with the provisions of 10 CFR 50.72 (b)(2)l. A follow up notification H was made at 1151 on August 6,1994 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW A-340.. On August 10,1994 comprehensive testing was performed on the Emergency Feedwater isolation and regulating valves for all three Steam Generators using PED test equipment. These tests revealed leak rates for various individual valves and combinations of valves from O to 36 GPM per Steam Generator (See Attachment A). Safety.halysis sensitivity studies conducted in parallel with these tests established the following revised acceptance criteria: 1) Lankage past individual isolation / regulating valves should be less than 40 GPM. *!?e l purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating valve to close. 2) Lankage past both the isolation and regulating valve to each Steam Generator should be less than 10 GPM. '!he purpose of this limit is to provide at least 30 minutes for operators to insure EFW is isolated to a faulted Steam Ger erator following a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to overfill. 3) - Zero leakage to a Steam Generator with a tube rupture within 30 minutes. The 1 purpose of this limit is 'o prevent lifdag SG safety valves due to overft11 following a Steam Generator Tube Rupture. In order to _ meet the revised acceptance criteria, EFW A 101 was adjusted to reduce leakage i past EFW A-101 & 338 to aero GPM and administrative controls were implemented to 3 ensure termination of any potential future leakags to a ruptured Steam Generator within 30 minutes using manual isolation valves. If 3 EXHIBIT 53f MD. ) 9G-010 PAGEdf_OF/# PAGE(S) 7 l l ! a..,_, -. _ _. _ _ -. _. _ _ -. _ _ ,. _.. ~.:

5.0 CONCLUSION

S, FACIS & RECOMMENDATIONS C'= M As The process used by Maine Yanker to enswe consistency betwen our 54rty Analysis and plant perfonnance dkt not enswe shat: c 1he sq ety)knctient $EFW-A 338, 339 a 340 wre tadequately i o wrVied durin pont installation sessing. j The s4rty)knaion <EFW A 101, 201, 301, 338, 339 & 340 to o isolateflow wre periodically wrVied weder the ISTprogram. 1he stery)knctions < EFW A 101, 338 a 340 wre adequately a rest dwing pan raintenance oestins, j Factst 1. A basic design function of the valves lanulled under EDCR 83 29 (EFW A 338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident - conditions. 2. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is i terminated by closure of the EFW isolation / feed regulating valves. ,( 3. Procedurs 17-212, Engineering Design Change Request, requires verification of design functions by functional testing unless such testing is not practical. 4. It is practical to perform leak testing and stroke timing with the EFW pumps in operation. 5. EDCR 83 29 did not require either leak testing nor valve stroke timing with the EFW pumps in operation. 6. De post inem1wion functional testing performed on EFW A-338,339 & 340 did not include either leak testing nor valve stroke timing with the EFW pumps in operation. i 7. Maine Yankee's IST program classifies EFW-A 101, 201, 301, 338, 339 & 340 as Category B valves.- 8, Maine Yankee's IST program defines Categwy B valves as, " valves for which seat l leakage in the closed position is inconsequential for fulfillment of their safety. function.' 9. In 1992 W.O. 92-1746 was initiated to correct EFW A-338 valve leakage. The FTI- . associated with this Work Order did not verify the ability of the valve to isolate flow. 1 10. De FTis associated with W.O. 94-3007 (EFW-A-338 valve leakage), did not verify 8 EXHIBIT.3 CGE NO. 1-96"()4 4 PAGEff)_0Fh PAGE(S)

~ l the aldlity of the valve to isolate flow. 11. The IT1s adw with W.O. 94-3075 (EFW A 340 valve leakage), did not verify {- the ability of the valve to isolate flow. 12. 'the Fns associated with W.O. 91-7584 (EFW A 101 valve leakage), did not verify the ability of the valve to isolate flow. 13. The Fris associated with W.O. 94-1997 (EFW A 101 not completely shut), did not verify the ability of the valve to isolate flow. Receaumendatione 1. Management should continue to emphamirs> support the timely completion of the action items idendfied in the QPD, NSEO report titled

  • Maine Yankee Safety Analysis inputs & Assumptions Review Project Final Report", dated July 8,1994

( C CASENO. 1-96-040 EXHlBIT 3 PAGE Q OF/'[/ PAGE(S) .-r. .-~- -. _..

Ceadadem Et ne RDCR procau used by Maine Yankee circa 1983-84 did not prmide adicient guidance so ensure all sdtry signyl cant design f l2:2 <DW-A 338, 339 a 340 wrr wrW post installarion. l Facts: 1. A baalc design funcdon of the valves installed under EDCR 83 29 (EFW A 338,339, 340) is to ensure isoladon of EFW flow to fauhed SO(s) under certain accident tenditions. 2. Procedure 17-212, Engineering Design Change Request, requires verification of da.lgo functions by fkmr+ianal testing unless such tesdag is not practical. l 3. It is praedcal to perform leak tesdag and stroke timing with the EFW pumps in operadon. 4. EDCR 83 29 did not require either leak testing rur valve stroke timing with the EFW pumps in operation. 5. De post installation functional testing performed on EFW-A 338,339 Ac 340 did not include either leak testing nor valve stroke timing with the EFW pumps in operation. t 6. A review of 4 recent EDCRs was performed by PED No instances of inadequate funedonal testing were discovered. 7. De current EDCR process contains more guidance for selecting appropriate functional test requirements than was provided 1983 1984. Reca==wedations: 1. The review of recent EDCRs performed by PED provides an indication that the functional testing specified in recent EDCRs may be adequate. However, due to time constraints the review performed was extremely limited in scope, in addition, udliradon of PED personnel to perform the review made it difficult to ensure the evaluation was completely objective. It is recommended that independent review of a representative sample of EDCRs be performed to more accurately asses.s the overall adequacy of EDCR functional testing. C CASE NO. ] gg. 04 0 EXHIBIT 3 10 PAGE C k 0FN/ PACE (S)

i Coachashen Cs The proonss used by he ISTprogram.'s desenmneJimestonal test reqadrements does not answe Ast he mary qf EFW-A 101, 201, 301, 338, 339 a 340 no isolate Dnergency Feedmeter)fow is periodknlly wipiet. f.xts: j 1. Maine Yankes's IST program dennes Category A valves as, ' valves for which sent leakage is limited to a specific manhoum amount in the closed position for fulfillment of their safety function." i l 2.- Maine Yankes's IST program dennes Category B valves as, ' valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety function." l 3. Maine Yankee's Safety Analysis for Main Steam Ilne Brw k assumes EFW flow is i terminated by closure of the EFW isolation / feed regulating valves. 5 4. Maine Yankee's IST prograra classifies EFW-A 101,201,301,338,339 & 340 as j Category B valves. i 5. 'the definition for Category A valves is interpreted within the Performance Fegineering Group as applying only to valves for which a specific maximum allowable leak rate (e.g., less than 20 GPM'> ls specified in the FSAR, Technical i Specifications, the IASD, or a Design Basis Summary Document. f 6. The interpretation presented in item 5 is informal and not documented within the IST prognm description, or memorandum. Recomme=dah* 1. Reclassify EFW A 101,201,301,338,339 & 340 as IST Category A valves. 2. Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our j Safety Analysis to provide flow isolation. 3. If recommendation 2 la odopted, the IST program should be reviewed to ensure valves which mre'. the revised criteria are incloded in the program and properly classified. 4 To ensure consistent classification, and valid audit results, IST program description should be modified to include the precise criteria used to classify valves as Category ~ A, B & C. PASE N0. 1 o4;O4 0 11 EXHIBIT .7 PAGElf__OF/l/_. PAGE(S)

P Ceneheden Ds The Wort Onder rmcess ed not enswe that FT12)>r FJW A 10), 338 l a M0 sessed the aMilty gthese ulws no per)brm their 24rtyfluaction, l (- nor wrped the dherinness of repair actMoses. Factst j 1. Procedure 016-3, Work Order Pmoens, requires that post maintenance Funcdonal I Test lastructions (FTI), ' Verify the ability of a component or system to perform its inunded funedon." and 'Demonstram that the original deficiency has been corrected.' 2. In 1992 W.O. 921746 was initiated to correct FFW A 333 valve leakage. 'the FTI maaarla w with this Work Order did not verify the ability of the valve to isolate flow. 3. 'the FTis maaaelaw with W.O. 94-3007 (EFW A 338 valve leakage), did not verify the ability of the valve to isolate flow. 4 The Fris associated with W.O. 94 3025 (EFW A 340 valve leakage), did not verify the ability of the valve to isolate flow. 5. The Fils duociated with W.O. 917534 (EFW A 101 valve leakage), did not verify the ability of the valve to isolate flow. 6. The Fris associated with W.O. 94-1997 (EFW A 101 not completely shut), did not verify the ability of the valve to isolate flow. ( 7.- The Fils associated with W.O. 91-7584,94-1746,941997,94-3007., and 94 3025 basically mimicked IST requirements for the associated valves. 8. A re iew of 30 recent Work Orders was performed by PED. No instances of twuate functional testing were identified. Receaunendatlomst 1. The review of recent Work Orders performed by PED provides an indication that the functional testing specified in recent W.O. may be adequate. However, the review performed was very limited in scope when comparea to the number of W.O. generated annually. In addition, at least four recent W.O.s,92-1746,94-1997,94-3007, and 94-3025 provided inadequate functional testing. As with the EDCR review, utilization of PED personnel to perform the review made it difficult to ensure the evaluation was completely objective. 'therefore, it is recomme< ed that an indeoendent review be performed of a mya,tative sample of V'.O s to more accurately asses.s overall adequacy of W.O. functional testing. 2. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing required by the IST program does not ensure adequate post maintenance functional testing - EXHlBlT 3 l ', M-1 - 9 ti - 0 4 0 PAGE @ 0FM PAGE(S)

C=u Et Procedure 3 5510, Mahuenance ofEFW Air Operosed THp Valns, and Wort Orders 921746, %3006 & %3025 ud not provide adequate instrsalons regarung orientation of ulw body toflow, nor orientation of vain octuator so disc shmt. Facts: 1. Work Order 92-1746 for EFW A 338 was not revised to reflect removal and

=installatlan of valve to correct body orientation and wrong gaskats. This includes the actuator.

2. 'Ihe 'as found" valve disc / seat body posidan on W.O. 92-1746 was correct for EFW-A 338. 3. No procedure 6-11-1 nor other work order for 1&C removal / reinstallation of the actuator can be located for W.O. 921746. 4. 'Ihe 'as.found" condition for EFW-A 338 on Aug.1994 fcand the disc was 180' out from the correct orientation. 5. The valve seat / disc is placed in the closed position prior to' assembly in the piping. 6. EFW-A-338, 339, and 340 are tn't open actuators. ( 7. EFW-A-338, 339, and 340 had no permanent external markings to indicate valve disc position (during repair activity for EFW-A 340, W.O. 94-3025 disc shaft was marked). Recan===adations: Procedu:e 5 5510 is currently being revised to address this conclusion. Valves EFW A-338 and 339 should be marked similar to EFW A-340 to provide external indication of disk position and flow orientation. ~ s !( l CASE ND, 1 - 9(i- 04 0 EXHIBIT 8 l PAuE SI OF]f/_ PAGE(S)

i Coachadee Fs The correct gasket oriensanlon nr not identilfat in Procedurt 3 3510, (- Mainnenance ofEFW Air Operased THp Valm, nor in the work order instructientpr 921746, N-3000, and N-3025. Facts: 1. Work Order 92-1746 was not revised to reflect renmal and relantallatian of valve to correct body orientation and the use of incorrect gaskets during the initial attempt to reassemble the valve. 2. Work on EFW-A-338 under Work Order 921746 lastallad 3' gaskets for both flanges. A 2%" gasket is required for the seat, or downstream flange (this fact is based on interview results). 3. W.O. 94-3025 for EFW.A 340 installed gaskets in reverse orientation, causing the valve to leak externally. 4 EFW A-338, 33) & 340 failed functional testing under EDCR 83-29 due to leaking gaskets. 5. As a result of Work Order 921746 for EFW A-338, DCR 92-170 was generated to correct the vendor drawing 7.69-46 and change Mfr. Inst. Manual C-9-2 to reflect gasket orientation. '( A copy of the vendor drawing 7.69-46 could not be. located in the 94 3025 work 6. package for EFW A-340. i Reca==endatloas: Procedure 5 5510 is currently being revised to address this conclusion. l l i t' + t f cAsnio. 1-96-040 EXHIBIT 8 PAGE_Id_OF_/f_ PAGE(3) ,,p. ._.5_.,_...~.mm. ,,._-....,l__. _.. ~.,,.. - _ _.. ,,,m...

4 Coachaden Gt Procedure 6111, lastnenentation and Controls Valw Calibration and Checks, prm4 des inadequase gsadance 80 enrure consissent proper reassent:j of EJW-A 333, 339, and 340. Facts: 1. Based on informados obtained during interviews,1&C technicians wert uricertain as to the proper application of procedurs 6-11 1 to the EFW valves. Ouidance concerning checking the sent/ disk for proper orientation was e=W particulady confusing. 1 2. For identical work: Step 5.2.1 of Procedust 6-11 1 was N/A'd for valve EFW-A 338 under W.O. o 94 3008, whereas signed completed for EFW-A-340 under W.O. 94 3025. Step 5.3.1 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-o 3008. Step 5.7.1 was N/A'd completely on W.O. 94-3008, but one sub step signed o as completed on 94 3025. Step 6.4 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-o 3008. c 3. 'Ihe valve actuator for EFW-A-340 was reinstalled twice under W.O. 94-3025. 'Ihere was only one complete 611-1 procedure. 4. The required documentation for Step 5.2.2.b of Procedure 6-11-1 is missing from W.O. 94-3025, 5. Step 5.2.2.e tf Procedure 6-11-1 for W.O. 94-3025 was signed off as complete. However, there are no manufacturer's position marks identified in the vendor manual / drawings to verify. 6. No completed copy of 6-11-1 could be found for W.O. 92-1746 (EFW-A-338 actuator removal / reinstallation). Reca==-4tions: 1. Procedure 611-1 is a combination of generic and specific instructions (i.e. main feed reg, valves). Two options could be considered: 1) add specific instructions to 611-1 for EFW A-338,339, and 340, or 2) use work order technical instructions for EFW-A 330,339, and 340. b 2.. Some improvement in mechanical maintenance and I&C interface / communication tASEND. 1 - 9G- 04 0 15 EXHIBIT 3 PAGE O OF./f.PAGE(S)

would appear necessary. For the work on EFW A-338, there appeared to be good E-'==. In spite of insuf5cient instructions, the IAC technician and C maintamance mechanic determined proper orientation and avoided the mistake made under W.O. 92-1746. Work on EFW-A-340 did not appear to have as good communications with both maintenance mechanics and IAC WhaWn being confused as to proper reassembly. When good communications were established, correct orientation was determined. Again, this was accomplished without sufficient guidance in the work instructions. ( C CASE NO, ) qq- 04 0 EXHIBIT 3 16 PAGE_tE_OF./g/_ PAGE(S) l'

. =... - _ i I r==ru Bt The disk)6r EfW.A 338 war mis positioned during performance of w.o. 92-n46 (Mar. 92) Factst 4 1. 05 26 87 Stroke time testing performed on EFW-A-338,339 & 340 under conditions representadve of a main steam line rupture. Stroks Smes within acceptance criteria. Test report noted that the valves demonstrated, 'very good shutoff" (0 2 GPM). 2. !!-23-91 Lankage rate idend5ed on W.O. 91-7585 was 612 GPM with EFW A-4 338 shut. W.O. 92-1746 generated for repairs. 3. 03-30-92 Completed repairs to EFW A-338 under W.O. 92-l' 46 (no seat 7 leakage test). 4. 08-04 94 Lankage rate idend5ed on W.O. 94-3007 was 300+ GPM with EFW-A-338 shut. i. 08 05 94 Completed repairs to EFW-A 33? Imakage test satisfactory, i 6. 08 10-94 Laak test of EFW-A-339& 340 were 17 GPM and 36 GPM respectively. 7. Review of equipment history (both mechanical ud I&C) revealed that EFW-A-338 was the caly valve to be disassembled (or have actuator removed). EFW-A 339 has not been disassembled since original inemilation. EFW A-340 was disassembled under W.O. 94-3025, aner the leak test performed on 08/10/94. Recomsmendations: None 4 EXHIBIT 3 CASE NO. ].- 4 G - 0 4 0 17 PAGE__ff 0F/_f/_ PAGE(S)

i i hDSCELLANEOUS OBSERVATIONS: (' OBSERVATION A. - Design change information from EDCR 83-29 (i.e. flange gasket orientation and welded disk ley to shaft) was not ' :-:-yerH into + Maintenance instructions for the valve, i.e. Procedure 5 5510, nor into the instruction manual C-9-2. A second opportunity occurred in 1992 i under W.O. 921746, DCR 92170 changed technical manual i instructions (C-9 2), and the vendor dws. 7.69-46 to include gasket orianamelan instructions. 'this information was not incorporated into Maintenanew instructions. RECOMMENDATION: During EDCR 33 29 implementation, the design notification form was the process to identify changes to respective departments. To further enhance this process and include other areas, the Configuration Management Program (Procedure 006-

8) was devdsi.d. Some increased awareness of this process is recommended.

t J OBSERVATION B. There was considerable rework involved in completing reinstallation of EFW A-340 under W.O. 94-3025. In summary: 1. '!he valve body was installed 180* out of position. (' 2. 'Ihe flange gaskets were installed incorrectly. 3. 'the valve was torqued in the "open" position, causing binding. . RECOMMENDATION: Maintenance department management should continue to stress the importance of good tumovers and coordmation between l different crafts working on the same job. Five mechanics and three I&C techs. covering three shifts were lavolved with this i work activity. Relative task performance observed during work on EFW A 338 / EFW A-340 would provide a useful comparative example. 1 C CASEll0. ]. - 9 6 - () 4 f) EXHlBIT 3 E 18 PAGE ~/O OF/f/_ PAGE(S)

ATTACHMENT A C + u.i WW 1M -C4-- M LA. womas es+mt 1 P see -t4-- M LS. ui \\ w.ast g, -C4-- N S.S. C Test As Found Results As 1.4ft SQ Valve (s) Closed 08/Q4/$ 08/10/94 08/l1/94 1 EFW-A 101 75 GPM 36 GPM 0GPM EFW-A-338 400 GPM 20 GPM 24 GPM 15 GPM 0GPM EFW-A-101 & 33h 0GPM 0GPM EFW-A-101, 338 & EFW-102 0GPM 0GPM 2 EFW A 201 EFW A 339 17 GPM 17 GPM 0GPM 0GPM EFW A 201 & 339 0GPM 0GPM EFW A-201,339 & EFW-202 OGPM 0GPM 3 EFW A-301 36 GPM 23 GPM ' EFW-A-340 4 GPM 0GPM EFW A-301 & 340 0GPM 0GPM EFW-A-301,340 & EFW 302 EXHIBIT 3 CASENO. ] - 9G- 04 0 OF/f/_kGE(S) 19 PAGE ~h

(r v m unenemtm e ( EXHIBIT 3 CASE 110. 1.-96-010 PAGE 71-CFM'__ FAGE( 20

S. C To: G. M. Leitch, Vice President Operatiora l Via: R. W. Blackmore From: G. N. Stowers e al Plant Root Cause Evalaution Report #190 Emergency Feedwater Valve Leakage ( PRCE Team Members: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED James Taylor - Senior Nuclear Safety Engineer, NSEG Event Date: Augius 5,1994 Report Date: August 1,1996 U. SEND, } 9(i- 04 0 EXHlBIT 3 PAGE.77.0F./f[ PAGE(s

7 1.0 CHARTER ( l. Identify musal factors contsbudng to the Eiregs.cy Feedwater Valve IAakage event described in 11amaan Event Report 94-016, Emergency Feedwater Isoladon Valve tankage. l 2. Provide recommendadons for reducing the probability of such an event recurring in the future. 3. Assess potential generic implicadons of any deficiencies i;oted. 4. Submit report on results of invesdradon within approximately one week. CASW. 1-96-040 ( EXHIBIT 3 PAGE 79 0F.$/_.PAGE(S) 2

.~ i 2.0 EXECimVE

SUMMARY

( At 1220, on August 4,1994 with the reactor in a cold shutdown condition, plant operators deserudnad that an Emergency Feedwater isolation valve for #1 Steam Generator was leaking by. Further investigation also identined leakage in the Emergency Feedwater (EFW) eupplies to #2 & #3 Steam Generators. Subsequently it was determined thet under accident conditions which require isoladon of Emergency Feedwater, isoladon valve leakage could enceed Safety Analysis assumpdons, j Maintenance activities were initiated to reduce Emergency Feedwater valve leakage. In addition, adplalatrative controls were implemented to ensure Emergency Feedwater leakage is maintained within the bounds of Safety Analysis assumptions during accident conditions. A review of this event has revealed the fouowing weaknesses which contributed to this event. EFW flow isolation assumptions in our Safety Analysis are not clearly -o presented in design basis reference documents. t Our IST program does not require periodic verification of the ability of the o EFW isolation and feed regulating valves to isolate EFW flow,

  • Ihe EDCR under which the EFW isolation valves were installed did not o

require verification of the ability of these valves to isolate EFW flow as part { of post installation functional testing, o Post maintenance functional testing performed on EFW A-338 in 1992 and EFW A 101 in 1991 & 1994, did not require verification of these valves ability to isolate EFW flow, Maintenance procedures lack sufficient guidance to ensure consistent proper o assembly of the EFW isolation valves. Lack of a single, w wpresasive, readily available reference for Safety Analysis inputs and astumptions, seems to have played a key role in this event. 'Ihe improved documentation and retrievability of Safety Analysis information to be offered by the Safety Analysis inputs & Atsumptions (SAID) which is currently under development, should significantly reduce the possibility of similar events once it becomes available for use. Management should continue to emphasize / support timely completion of this - effort. Other significant activities which are recommended include the following: Initiation of an independent, broad scope, review of EDCRs and W.O.s to o determine the overall adequacy of functional testing performed at Maine Yankee', 3 EN CASEl10. ] 9G- 04 0 PAGE 7f 0F./f/_PAGE(S) 3

o Review of the IST program to ensure the current scope of the program and the criteria used to establish the functional test requirements for specific ) -( componer.ts are adequae. o Efforts to increase awareness among those who develop, review, and approve functional tests, that the functional testing required by our IST program should i not be relied upon to ensure adequae post maintenance functional testing. o Revision of maintenance procedures to provide additiod guidance for assembling the EFW isolat m valves to ensure consistent proper orientation of i the valve disk, body, seat, and actuator. e ( ( CASE ND. ] 040 EXHIBIT 3 PAGE % OF$PAGE S) 4 t

...... ~ - ~. - - - - - - - -.... - F 3.V: - RECOMMENDATIONS .-( -1. Management should continue to emphasize / support timely completion of the action - items identified in the QPD, NSEO report titled _" Maine Yankee Safety Analysis Inputs & Assumpdons Review Project Final Report", dated July 8,1994 2, Perform an independent review of a representative sample of EDCRs to more completely assess the overall adequacy of EDCR functional testing. - 3. Reclassify EFW A-101,201,301,338,339 & 340 as IST Category A valves. 4. Consideradon should be given to modifying PED's =leting interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. 5. If Iacommendation 2 is adopted, the IST program should be reviewed to ensure all valves which meet the revised criteria are included in the program and properly Classified. 6. Modify the IST program description to include the precise criteria used to classify valves as Category A, B & C. /" 7. Perform an lah;+ ht review of a representative sample of W.O.s to more fully l \\ assess the overall adequacy of.W.O. functional testing. 8. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing required by the IST program does not r-~~ily ensure adequate post maintenance functional testing. ' 9. Revise procedure 5-55-10, Maintenance of EFW Air Operated Trip Valves, to: Provide sufficient guidance to ensure valves covered by this r.rocedure are o properly assembled following maintenance, Identify the corrt.t gaskot to be used in the upstream and downstream o flanges. 10. The valve stem and valve body of EFW-A-338 and 339 should be marked in a manner similar to EFW-A-340 to provide external indication of disk position, and - flow orientation. I1. Either 1) Add specific instructions to procedure 6-11-1, Instrumentation and Controls {- Valve Calibration and Checks, to address the details necessary to ensue of proper attachment of the operators for EFW-A-330,339, and 340 or; 2) In the future, CASENO. ] - 9 fi - 0 4 0 5 EXHIBlT 3 PAGE ~)) 0F_8/_ PAGE(S)

include specific instructions for attaching the operator in work order technical instructions. e-k Maintenance department management shNid matinue to stress the importance of good 12. turnovers and coordination between diffenzt crafts working on the same job. Relative task performance observed during work on EFW-A-338 / EFW-A-340 would provide a useful comparative example. CASE NO. } _ g, tj - 0 4 0 EXHIBIT 3 ( PAGE 71I 0F //// PAGE(S) 6

4.0 NARRATIVE On August 4,1994 Maine Yankee was in a cold shutdown condition making prepantions to restart the plant following a maintenance entage to correct Steam Generator tube leakage. At approximately 1220 while performing a leak test of the Emergency Feedwater Isolation Valve (ISV) for #1 Steam Generator using normal system instrumentation, it was determinea that EFW A-338 leaked by at a rate of 400 GPM. Leakage past #1 Steam Generator Emergency Feedwater Regulating Valve EFW-A-101 at 75 GPM was also identified. Subsequent investigation determined that the actuator for EFW-A-338 was coupled to the disk approximately 180 degrees, out of alignment. This misalignment is believed to have occurred during valve maintenance performed during our 1992 Refueling Outage. Maintenance activity was initiated to reposition the actuator / disk to the correct orientation. At 1700 on August 5,1994 it was determined that Maine Yankee s Steam Line Break Safety Analysis assumes zero leakage past Emergency Feedwater isolation and regulating valves. Thuefore, at 1957 the NRC was appraised of the situation via the Emergency Notification System in accordance with the provisions of 10 CFR 50.72 (b)(2)i. A follow up notification was made at 1151' on August 6,1994 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW-A-340. On August 10,1994 comprehensive testing was performed on the Emergency Feedwater isolation and regulating valves for all three Steam Generators using PF.D test equipnent. These tests revealed leak rates for various individual valves and combinations of valves from (. O to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studies conducted in parallel with these tests established the following revised acceptance criteria: 1) Leakage past individual isolation / regulating val es should be less than 40 CPM. The purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating valve to close. 2) Leakage past both the isolation and regulating valve to each Steam Generator should be less than 10 GPM. The purpose of this limit is to provide at least 30 minutes for operators to insum EFW is isolated to a faulted Steam Generator following a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to overfill. 3) Zero leakage, to a Steam Generator with a tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due to overfill following a Steam Generator Tube Rupture. In order to rncet the revised acceptance criteria, EFW-A-101 was adjusted to redum leakage past EFW-A-101 & 338 to zero GPM and administrative controls were implemented to ensure termination of any potential future leakage to a ruptured Steam Generator within 30 minutes using manual isolation valves. CASE ND. 3 040 EXHlBiT 3 7 PAGE M OF /4'/ PAGE(S)

4

5.0 CONCLUSION

S, FACTS A RECOMMENDATIONS C Cnarka=Lan At The per> cess used by Maine Yankee to ensure consistency betwen our Sqfety Analysis and plant performance did not ensure that; lhe sqfetyJimctions ofEFW-A-338, 339 & 340 wre adequately o wit / fed during post lastallation testing. The sqfety,fimction of EFW-A-101, 201, 301, 338, 339 & si to o isolateflow wie periodically wrfled under the ISTprogram. The sgetyf.metionr of EFW A.101, 338 & 340 wre adequately o test during post maintenance testing. Facts: 1. A basic design function of the valves installed under EDCR 83-29 (EFW-A-338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident conditions. 2. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EFW isolation / feed regulating valves. ( 3. Procedure 17 21-2, Engineering Design Change Request, requires verification of design functions by functional testing unless such testing is not practical. 4. It is practical to perform leak testing and stroke timing with the EFW pumps in operation. 5. EDCR 83-29 did not require either leak testing nor valve stroke timing with the EFW pumps in operation. 6. The post installation functional testing performed on EFW-A-338, 339 & 340 did not include either leak testing nor valve stroke timing with the EFW pumps in operation. 7. Maine Yankee's IST program classifies EFW-A-101,201,301,338,339 & 340 as Category B valves. 8. Maine Yankee's IST program defines Category B valves as, " valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety function." 9. In 1992 W.O. 92-1746 was initiated to correct EFW-A-338 valve leakage. The FTI associated with this Work Order did not verify the ability of the valve to isolate flow. 10. - The FTIs associated with W.O. 94 3007 (FJW-A-338 valve leakage), did not verify ew 2 8 ~ PAGE TO OF /Y/ PAGE(S)

the ability of the valve to isolate flow. I1. De PTis associated with W.O. 94-3075 (EFW-A-340 valve leakage), did not verify the ability of the valve to isolate flow. i 12. De FTIs associded with W.O. 91-7584 (EFW-A-101 valve leakage), did not verify the ability of the valve to isolate flow. 1 13. The FTIs associated with W.O. 94-1997 (EFW-A-101 not completely shut), did not verify the ability of the valve to isolate flow. Mmenmanendatlans. 1. Management should continue to emphasize / support the timely completion of the action items identified in the QPD, NSEG report titled

  • Maine Yankee Safety Analysis Inputs & Assumptions Review Project Final Report", dated July 8,1994.

C ( CASE ND-1 040 g EXHIBIT f/ OF /f/ PAGE(S) 9 PAGE t

Coackasion Et The EDCR pruess used by Maine Yankee circa 1983-84 did not prvHde ssdicient gsddante to ensure all sqfety signipcant design (- fimettons of EFW-A-338, 339 a 340 were wrtfled post trutallation. Facts:- 1. A basic design function of the valves installed under EDCR 83-29 (EFW A-338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident conditions. 2. Procedure 17-21-2, Engineering Design Change Request, requires verification of design functions by functional testing unless such testing is not practical. 3. It la practical to ptforia leak testing and stroke timing with the EFW pumps in operation. ~4. EDCR 83 29 did net require either leak testing nor valve stroke timing with the EFW pumps in o'peration. 5. The post installation functional testing performed on EFW-A-338, 339 & 340 did not include either leak testing nor valve stro'.:e timing with the EFW pumps in operation. 6. A review of 4 recent EDCRs was performed by PED. No instances ofinadequate funcdonal testing were discovered. { 7. The current EDCR process contains more guidance for selecting appropriate functional test requirements than was provided 1983-1984. Recommendations: 1. The review of recent EDCRs performed by PED provides an indication that the functional testing specified in recent EDCRs may be adequate. However, due to time constraints the review performed was extremely limited in scope, in addition, utilization of PED personnel to perform the review made it difficult to ensure the evaluation was completely objective. It is recomW that independent review of a representative sample of EDCRs be giferswd to more accurately assess the overall adequacy of EDCR functional testing. C-CASENO. 1-96-040 EXHlBIT PAGE @ OF8f.PAGE(S) 10

Conclusion Ct The process used by the ISTprogram to determinefuncnonal test requirements does not ensure that the ability of EFW-A 101, 201, 301, .C 338, 339 & 340 to isolate Emergency Feedmuerflow is periodically wryled. Facts: 1. Maine Yankee's IST program defines Category A valves as, " valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their safety function." 2. Maine Yankee's IST program defines Category B valves as, " valves for which seat leakage in the closed position is inconsequential fr fulfillment of their safety '~ funcdon." 3. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EF1V isolation / feed regult. ting valves. Maine Yankee's IST program classifies EFW-A-101,201,301,338,' 339 & 340 as 4. Category B valves. 9 5. The definition for Cate6ory A valves is interpreted within the Performance Engineering Group as applying only to valves for which a specific maximum allowable leak rate (e.g., less than 20 GPM) is specified in the FSAR, Technical Specifications, the IASD, or a Design Basis Summary Document. 6. The interpretation presented in item 5 is informal and not documented within the IST program description, or memorandum. Recomunendations: 1. Reclassify EFW-A-101,20;,301,338,339 & 340 as IST Category A valves. 2. Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. 3. If recommendation 2 is adopted, the IST progrsm should be reviewed to ensure valves which meet the revised criteria are included in the program and properly classified. 4. To ensure consistent classification, and valid audit results, IST program description should be modified to include the precise criteria used to classify valves as Category A, B & C. 3 CASE NO. ] - 9G- 04 0 EXHIBIT j 11 PAGE O OF/#/ PAGE(S) l

Concknsion Dt 7he Wort Order process did not ensure that FTl sfor EFW-A-101, 338 & 340 tested the ability of these unins to perform their sqfetyfunction, i (. nor wri6ed the efectinness of repair activities. Facts: 1. Procedure 016-3, Work Order Process, requires that post maintenance Functional Test Instructions (FrD, " Verify the ability of a component or system to perform its intended function.' and " Demonstrate that the original deficiency has been' corrected." i 2. In 1992 W.O. 92-1746 was initiated to correct EFW A 338 valve leakage. The FTI associated with this Work Order did not verify the ability of the valve to isolate flow. 3. The FT!s associated with W.O. 94-3007 (EFW A-338 valvo leakage), did igg verify the ability of the valve to isolate flow. 4. 'the FTIs associated with W.O. 94-3025 (EFW-A 340 valve leakage), did not verify the ability of the valve to isolate flow. 5. The FTIs' associated with W.O. 91-7584 (EFW-A 101 valve leakage), did not verify the ability of the valve to isolate flow. 6.. 'Ihe FTIs associated with W.O. 941997 (EFW-A-101 not completely shut), did not verify the ability of the valve to isolate flow. ( 7. 'the FTIs associated with W.O. 91-7584,94-1746,94-1997,94-3007, and 94-3025 basically mimicked IST requirements for the auociated valves. 8. A review of 30 recent Work Orders was performed by PED. No instances of inadequate functional testing were identified. Recomunendations: 1. 'the review of recent Work Orders performed by PED provules an indication that the functional testing specified in recent W.O. may be adequate. However, the review performed was very limited in scope when compamd to the number of W.O. generated armually. In addition, at least four recent W.O.s,92-1746,94-1997,94-3007, and 94-3025 provided inadequate functional testing. As with the EDCR review, utiliration of PED personnel to perform the review made it difficult to ensure the evaluation was completely objective. 'Iherefore, it is recommerded that an let review be performed of a representative sample of W.O.s to more accurately assess overall adequacy of W.O. functional testing.

2. -

Individuals responsible for specifying, reviewing, and approving W.O. fuW test requirements should be sensitiud to the fact that specifying the component testing required by the IST program does not ensure adequate post maintensace functional testing. C N CASE ND. 1 - 9 G - 04 0 EXHIBIT PAGE M OF./_dPAGE( 12

Coactusion E: Procedure 5-5510, Mcintenance of EFW Air Operated Trip Valws, and Work Orders 92-1746, 94-3008 & 94-3025 did not provide C. adequate instructions regarding orientation of m!ve body toflow, nor orientarion of m!ve actuator to disc.thap. Facts: 1' Work Order 92-1746 for EFW A-338 was not revised to reflect removal and rein #allation of valve to correct body orientation and wrong gaskets. This includes the actuator. 2. The 'as found" valve disc / seat body position on W.O. 92-1746 was correct for EFW-A-338, 3. No procedure 6-11-1 nor other work order for I&C removal / reinstallation of the actuator can be located for W.O. 92-1746. 4. The *as found" condition for EFW-A-338 on Aug.1994 found the disc was 180' out from the correct orientation. 5. The valve seat / disc is placed in the closed position prior to assembly in the piping. 6. EFW-A-338,339, and 340 are fail open actuators. ( 7. EFW-A-338, 339, and 340 had no permanent external markings to indicate valve disc position (during repair activity for EFW-A-340, W,0. 94 3025 disc shaft was marked). Recommendations: Procedure 5 55-10 is currently being revised to address this conclusion. Valves EFW-A-338 and 339 should be marked similar to EFW-A-340 to provide extemal indication of disk position and flow orientation. EASE NS. }.96-040 N EXHIBIT PAGE 60F/f/_.PAGE 13

Conclusion F:- The correct gasket orientation us not identined in Procedure 5-55-10, - Maintenance of EFW Air Operated Trip Valws, nor in the wik order C instructionsfor 92-1746, 94-3008, and 94-3025. Facts: 1. Work Order 921746 was not revised to reflect removal and reinstallation of valve to correct body orientation and the use of incorrect gaskets during the initial attempt to seassemble the valve. 2.. Work on EFW-A-338 under Work Order 92-1746 install J 3" gaskets for both flanges. A 2%" gasket is required for the seat, or downstream flange (this fact is bued on interview results). - 3. W.O. 94-3025 for EFW-A-340 installed gaskets in reverse orientation, causing the valve to leak externally. 4.. EFW-A-338,339 & 340 failed functional testing under EDCR 83-29 due to leding ~ - gaskets. 5. As a result of Wor! Order 921746 for EFW-A-338, DCR 92-170 was generated to correct the vendor drawing 7.69-46 and change Mfr. Inst. Manual C-9-2 to reflect gasket orientation. 6. A copy of the vendor drawing 7.69-46 could not be located in the 94-3025 work package for EFW-A-340. Recommendations: Procedure 5-55-10 is currently being revised to address this conclusion. UE 1GG-040 EXHIBIT 3 PAGE Ed OF/l/_.PAGE 14

Conchasion Gt Pmcedure 6111, instrumentation and Controls' Valw Calibration and Checks, providos inadequate guidance to ensure consistent proper (- reassembly of EFW-A-338, 339, and 340. Facts: ) 1. Based on information obtained during interviews, I&C technicians were uncertain as to the proper application of procedure 6-11-1 to the EFW valves. Guidance concerning checking the nest / disk for proper orientation was considered particularly confusing. 2. For identical work: o Step 5.2.1 of Procedure 611-1 was N/A'd for valve EFW-A-338 under W.O. 94-3008, whereas signed completed for EFW-A-340 under W.O. 94-3025. o Step 5.3.1 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-3008. ' Step 5.7.1 was N/A'd completely on W.O. 94-3008, but one sub step signed o as completed on 94-3025. o Step 6.4 was signed as completed on W.O. 94-3025 and N/A'd on W.O. 94-('. 3008. 3. 'The valve actuator for EFW-A 340 was reinstalled twice under W.O. 94-3025. There was only one complete 6-11-1 procedure. 4. The required documentation for Step 5.2.2.b of Procedure 6-11-1 is missing from W.O. 94-3025, 5. Step 5.2.2.a of Procedure 6-11-1 for W.O. 94-3025 was signed off as complete. However, there are no manufacturer's position marks identified in the vendor manual / drawings to verify. 6. No completed copy of 6-11-1 could be found for W.O. 92-1746 (EFW-A-338 actuator removal / reinstallation). Recamawadatloas: 1. Procedure 6-11-1 is a combination of generic and spectfic instructions (i.e. main feed reg. valves). Two options could be considered: 1) add specific instructions to 6-11-1 for EFW-A-338, 339, and 340, or 2) use work order technical instructions for EFW-A-330, 339, and 340. ( _2.- Some improvement in mechanical maintenance and I&C interface / communication CASE NO. ]-96-040 15 EXHIBIT PAGE @ OFjf PAGE(S

i would appear r-==y. For the work on EFW-A-33P there appeared to be good communications. In spite of insufficient instructions, the I&C technician and mainte= nee mechanic determined proper orientation and avoided the mieka made under W.O. 92-1746. Work on EFW-A-340 did not appear to have as good communications with both rraintenance mechanics and I&C technician being confused as to proper reassembly. When good communications were established, correct orientation was determined. Again, this was accomplished without sufficient guidance in the work instructions. ( t b EN 9 6 - 0 4 0 EXHIBIT 8 PAGE %OFMLPAGE( 16

N MISCELLANEOUS OBSERVATIONS: OBSERVATION A. Design change information from EDCR 83-29 (i.e. flange gasket i orientation and welded disk key to shaft) was not incorporated into Maintenance instructions for the valve, i.e. Procedure 5 55-10, nor into the instruction manual C-9-2. A second opportunity occurred in 1992 under W.O. 92-1746, DCR 92170 changed technical manual instructions (C-9-2), and the vendor dwg. 7.69-46 to include gasket orientation instructions. This information was not incorporated into Maintenance instrxtions. RECOMMENDATION: D. iring EDCR 83-29 implementation, the design notification - form was the process to identify changes to respective - departments. To further enhance this process and include other areas, the Configuration Management Program (Procedure 04

8) was developed. Some increased awareness of this process is recommended.

OBSERVATION B. "Ihere was considerable rework involved in completing reinstallation of EFW-A-340 under W.O. 94-3025. In summary: l. The valve body was installed 180' out of position. 2. 'Ihe flange gaskets were installed incorrectly. 3. The valve was toequed iri the "open" position, causing binding. RECOMMENDATION: Maintenance department management should continue to stress the importance of good turnovers and coordination between different crafts working on the samejob. Five mecteics and three I&C techs. covenng three shifts were involved with this work activity. Relative task performance observed during work on EFW-A-338 / EFW-A-340 would provide a useful comparative example. t I E ( tem. 1 9o-04o-18 PAGE _OFlf/_PAGE(S) l

i ATTACHMENT A T WW4438 W541M WW182 -C4-- n s.a. new+am Fe 6 aiii W Mt w. W54448 WW4 art P 3EC -N-- as u. ) Test As Found Results As left SQ Valve (OClosed C3/04/94 08/10/94 08/11/94 1 EFW-A-101 75 GPM 36 GPM 0GPM EFW-A-338 400 GPM 20 GPM 24 GPM 15 GPM 0GPM EFW-A-101 & 338 0GPM 0GPM EFW-A-101,338 & EFW-102 0GPM 0GPM 2 EFW-A-201 EFW-A-339 17 GPM 17 GPM 0GPM 0GPM EFW-A 201 & 339 0GPM 0GPM EFW-A-201,339 & EFW-202 OGPM 0GPM 3 EFW-A-301 36 GPM 23 GPM EFW-A-340 4 GPM 0GPM EFW-A-301 & 340 0GPM 0GPM EFW-A-301, 340 & EFW-302

(

SNEN0. 1-96-040 d EXHIBIT __ 19 PAGE 94 0F.lf/_PAGE(S)

g

w 1

l:. $.tASE0.- 1. 9 g. 0 4 0 f.. exsisi7 h PAGE f/ /f/_PAGE(S) ~ OF. I 1 m

4 To: G. M. Leitch, Vice President Opentions Via: R. W. Blackmore From: G. N. Stowers et al P ant Root Cause Evaluation Report #190 Emergency Feedwater Valve Leakage ~~ ( PRCE Team Memt.crs: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED James Taylor - Senior Nuclear Safety Engineer, NSEG Event Date: August 5,1994 Report Date: August 2,1996 CASENO. ]. 9(;. 04 0 EXHlBIT 8 PAGE 9} OF8L.PAGE(S)

1.0 CHARTER - Charter provided by Manager, PED and endorsed by Vice President, Operations 1. Identify causal factors contributing to the Emergency Feedwater Valve Lankage event described in Ucensee Event Report 94416, Emergency Feedwater Isoladon Vr.lve Lankage. -2. Provide recommendations for reducing the probability of such an event recurring in the future. 3. - Assess potential generic implications of any deficiencies noted. e

s..

( CASENO. 1-96# 04 0 2 EXHIBIT d PAGEkOF/f/.PAGE(S)

.... - ~. - 2.0 EXFCUTIVE

SUMMARY

( At 1220, on August 4,'1994 with the reactor in a cold shutdown condition, plant operators determined that an Emergency Feedwater isolation valve for #1 Steam Generator was leaking by, Further investigation also identified leakage in the Emergency Feedwater (EFW) supplies to #2 & #3 Steam Generators. Subsequently it was determined that under accident conditions which require isolation of Emergency Feedwater, iso'.ation valve leakage could exceed Safety Analysis assumptions. Maintenance activities were initiated to reduce Emergency Feedwater valve leakage. In addition, administrative controls were implemmted to ensure Emergency. Feedwater leakage is maintained within the bounds of Safety Analysis assumptions during accident conditions. A review of this event has revealed the following causal factors: Maintenance instructions lack sufficient guidance to ensure consistent proper o assembly of the EFW isolation valves. An' erroneous belief that specifying the functional testing required by the IST o program for a given comporient, ensures adequate post maintenance functional testing of all the component's safety fuactions.

  • Ihe EDCR ur der which the EFW isolation valves were installed did not o

require verification cf the ability of these valves to isolate EFW flow as part -( of post installation functional testing, Weaknesses in our self assessment process resulted in several missed o opportunities to identify / correct deficiencies which contributed to this event. EFW flow isolation assumptions in our Safety Analysis are not clearly o presented in design basis reference documents. To summarut recommended actions to reduce the probability of recurrence: Maintenance procedures should be revised to provide additional guidance for o assembling the EFW isolation valves to ensure consistent, proper orientation of the valve disk, body, seat, and actuator, Action should be taken to increase awareness among tho:e who develop, o review, and approve fun.ctional tests, that the functional testing required by our IST program should not be rehed upon to ensure adequate post maintenance functional testing. The EDCR, Work Order, and IST prre*<= should be reviewed and o enhanced as miary to ensure that functional testing specified by the,e programs consistently verifies the ability of components to adequately perform C. their intended safety functions. 3 EXHlBlT 3 CASE NO-1-96-040, PAGE f_OFMPAGE

A broad scope review of EDCRs and W.O s should be conducted to determine o the overall adequacy of functional testing performed at Maine Yankee. -( Maine Yankee's self assessment process should be reviewed, and enhanced as o r**"ary, to address inadequacies which contributed to the failure to previously identify and correct the deficiencies identified in this report. Management should continue to emphasize / support timely completion of the o action items identified in the QPD, NSEG report titled

  • Maine Yankee Safety Analysis Inputs & Assumptions Review Project Final' Report", dated July 8, 1994.

e e p e ema 3 90- on 8 c EXHIBIT 4 PAGE_7I__OF_/$ PAGE(S)

3.0 RECOMMENDED' ACTIONS TO PREVENT RECURRENCE ,( l. Provide impromd guidance for the assembly of EFW-A-338, 339 & 340 to ensure V - consistent, proler reassembly of these and similar 5 al.es. ( See Causal Factors A & B)- Valves EFW-A-335,339 & 340 should oc marked to provide external indication of

2..

disk position, seat position, and correct flow orientation. ( See Causal Factor A) A representative sample of maintenance procedures which hand off to other 3. procedures to complete a maintenance activity should be reviewed to determine averall adequacy of maintenance procedure interfaces. ( See Causal Factor A) Clear guidance for adjusting air operated valves needs to be provided to I&C L 4. Technicians. (See Causal Factor C) 5. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements'should be sensitized to the fact that specifying the component testing required by the IST program does not necessarily ensure adequate post maintenance - functional testing.- ( See Causal Factor D) 6. An independent review should be performed of a representative sample of W.O.s to. more accurately assess overall adequacy of W.O. f=+iaa=1 testing. (See Causal Factor D) 7. Consideration should be given to establishing programmatic requirements which ensure that safety significant functions of plant components are periodically verified. Such verification could be implemented as part of our surveillance program, IST program, or some other program. (See Causal Factor E) 8 PED should evaluate the a mropriateness of classifying EFW-A-101, 201, 301, 338, 339 & 340 as IST Category B valves. (See Causal Factor E) 9 Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. (See Causal Factot E) 10. If recommendation 9 is adopted,- the IST program should be reviewed to ensure valves which meet the revised criteria are included in the program and properly classified. (See Causal Factor E). I1. To ensure enneiennt classification, and valid audit results, IST program description should be revised to include the.w.ed interpmWs of those criteria used to'- classify valves as CQsri A, B & C. (See Causal Factor E) L12. - An independent review of a representative sample of EDCRs be performed to more 8 EXHEi CASENO. 1 9 6 - 0 4 0 5 PAGE$__OF8LPAGE(S, m-+ev r-re-av.- se e. .w- ---rw ,r-me-w-w- r-w-a----, --er-e- v rw

0 c accurately assess 'the overall adequacy of EDCR functional testing. (See Causal Factor F)- C~ 13. Maine Yankee's self assessment process should be reviewed to identify any inadequacies which may have contributed to the failure to previously identify and correct the deficiencies identified in this report. (See Causal Factor G) 14. Management should continue to emp:maiw/ support the timely completion of the action ite.as identified in the QPD, NSEG report titled " Maine Yankee Safety Analysis Inputs & Assumptions Review Project Final Report", dated July 8,1994. (See Causal Factor H) l 15. Management should continue to stress the importance of procedure compliance / attention to detail. (See: Causal Factor C, Fact 5; Causal Factor D, Fact 1; Causal Factor G, Facts 8,9,11, & 15 0 (~ CASE ND. 1-96-040 6 EXHIBIT I PAGE 9[ OFf/_ PAGE(S

--.--...-~-- i I 4.0. NARRATIVE Maine Yankee's Emergency Feedwater (EFW) Isolation Valves EFW-A-338,339 & 340 were installed under EDCR 83-29 in May of 1984. The purpose of this modification was to provide redundant isolation of FFW to affected Steam Generators during a Main Steam Rupture event. Post inuallation ating of these valves verified: system integrity, valve stroke times, and the capacity and lntegrity of the valves air operating system. The ability of l the isolation valves to stroke under ocsign conditions of pressure and flow and their ability to J-actually interrupt flow was not tested. As design basis documents did not specify a specific maximum allowable leak rate for these valves, they were classified as IST Category B i valves. In accordance with the provisions of Maine Yankee's IST program for Category B valves, penodic testing rymrements included ondy verification of stroke time and pc:ition indication. In 1987 during development of a Design Basis Summary Document for the EFW system, it was noted that the ability of the valves to stroke under design conditions had not been verified during the post installation testing specified by EDCR 83-29. Follow up tasting conducted on May 5,1987 confirmed satisfactory stoke times for EFW-A-301 & 340 under design conditions of pressure and flow. Although a determination of the leak tightness of these valves was not a test objective, it was noted in the test report that the valves exhibited i "very good shutoff" (0-2 GPM). In November,1991 operations noted leakage past EFW-A-101 & 338 and initiated work

(

orders for their repair. In March,1992 EFW-A-338 was disassembled and the cause of the j'\\ leakage was determined to be erosion of the seat. (It is believed that the disk for EFW-A-?38 was installed incorrectly at this time). An ine-ion of EFW-A-101 conducted during the same time frame confirmed proper seating of the valve. Post maintenance functional testing for both valves consisted of a stroke test and a check for external leakage. In May,1994 a work order was initiated by Operations against EFW-A-101 due to the valve j. not fu'!y closing. Repair actions consisted of increasing the output of the air regulator supplying the valve operator from 32 to 35 PSIG. It was noted on the work order that J regulator output could not be increased much further. Post maintenance testing consisted of stroking the valve. On July 16,1994 a plant shutdown was performed due to Steam Generator tube leakage. While Steam Generator levels were being maintained using the Emergency Feedwater i system, plant operators noted a gradual, ur==W increase in #1 Steam Generator level. SuWng leakage past the EFW feed regulating valve for #1 Steam Generator (EFW-A-101) [ as the source of the naare! increase, operators redirected EFW flow through the main feed system and isolated the EFW feed regulating valves. An anempt to recreate the 4 apparent icekage later in the shift was unsuccessful and the source of the unanticipated level . mcrease remamed unidentified.- L ~ On August 4,1994 Maine Yankee was in a cold shutdown condition making preparations to t restart the plant following completion of Steam Generator inspections and repair. At { approximately 1220 operators conducted a EFW feed system leak test to follow up on the j CASE NO. j gg. 04 0 7 PAGE ff 0FgPAGE(S) i -~m -m.c-m.--eg--me-3 m wg-g- .y-aq,-.my-g-wr

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i anomalies noted on July 16. -During this test it was determined that the E. Feedwater Isolation Valve for #1 Steam Generator (EFW-A-338) leaked by at a rate of ' z(_ about 400 GPM.. It was also noted that the Emergency Feedwater Regulating Valve for #1 Steam Generator (EFW-A-101) leaked _by at a rate of approximately 75 GPM Subsequent investigation determined that the actuator for EFW-A-338 was coupled to the disk appmximately 180 degrees out'of alignment. This tr.isalignment is believed to have occurred during valve maintenance performed during our 1992 Refueling Outage. Maintenance l retivity was initiated to reposition the actuator / disk to the correct orientation. Leakage past l EFW-A-101.was corrected by adjusting valve stroke. At 1700 on August 5,1994 it was determined that Maine Yankee's Steam Line Break Safety Amlysis nonmes vem 14 age past Emergency Feedwater isolation and regulating valves. Derefore, at.1957 the NRC was appnsed of the situation via the Emergency Notification System in accordance with the provisions of 10 CFR 50.72 (b)(2)i. A follow up notification was made at 1151 on August 6,1994 after leakage was identified past the Emergency 'Feedwater Isolation Valve for #3 Steam Generator, EFW-A-340. This leakage was subsequently determined to be caused by seat / disk erosion. On August 10,1994 comprehensive testing was performed on the Emergency Feedwater isolation and regulating valves for all three Steam Generators using PED test equipment. Dese tests revealed leak rates for various individual valves and combinations of valves from O to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studies conducted in parallel with these tests established the following revised acceptance criteria: '( 1) leakage past individual isolation / regulating valves should be less than 40 GPM. The purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating va ve to close. 8 2) Leakage past both the isolation and regulating valve to each Steam Generator should be less than 10 GPM. The purpose of this limit is to provide at least 30 minutes for operators to insure EFW is isolated to a faulted Steam Generator following a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be j lifted due to overfill. 3) Zero leakage to a Steam Generator with a tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due'to overfill following a Steam Generator Tube Rupture. In order to meet the revised accertance criteria, EFW-A-101 was adjusted to reduce leakage L past EFW-A-101 & 338 to zero GPM and administrative controls were implemented to - ensure termination of any potential future leakage to a ruptund Steam Generator within 30 minutes using manual isolation valves. Comprehensive "as left" leak testing conducted by l-- PED lon August 11,1994 confirmed that EFW isolation / regulating valve leakage was within the revised acceptance criteria. i Q CME NO. 1-961040 ;. 8 EXHIBIT 8 PAGEKOF 8/ - PAGE(S)

5.0 CAUSAL FACTORS, SUPPORTING FACTS, AND RECOMMEND ACTIONS TO ' PREVENT RECURRENCE Causal Esctor. At Procedure 3-55-10, Maintenance ofEFW Air Operated Trip Valus, \\ does not provide adequate guidance to ensure consistent proper assembly of EFW-A-338, 339 & 340. Facts: - 1. In March 1992, following maintenance performed under W.O. 92-1746, EFW A 338 was initially reinstalled with incorrect valve body orientation and incorrect gaskets. 2. Proceoure 5-55-10 leaves the valve in tie cloud positica prior to as:embly in the ii P P ng. 3. 'Ihe actuators for EFW-A-338, 339 & 340 are in the "open" position when they are attached to the valves. 4. The mechanical configuration of EFW-A-338,339 & 340 and their' actuators allows reassembly with the valve disk either in the correct orientation or misaligned by 180 degrees. 5. The position of the seat and disk for EFW-A-338,339, and 340 can not be determined by visual inspection after the valves have been assembled. 'Ihere are no permanent external markings to indicate valve disc position (during repair activity for (- EFW-A 340, W.O. 94-3025 disc shaft was marked). 6. Procedure 5-55-10 does not provide guidance for rotating the disk to the correct orientation prior to installing the actuator. 7. 'Ihe "as found" condition for EFW-A-338 on Aug.1994 found the disc was 180' out from the correct orientation. (W.O. 94-3008) 8. 'Ihere was considerable rework involved in completing reinstallation of EFW-A-340 under W.O. 94-3025. In summary; The valve body was installed 180* out of position. o o_ ' 'Ihe flange gaskets were installed incorrectly, The valve was torqued in the "open" position, causing binding. o Reca===adad Actions to Prevent Recurrence: 1. Provide improved guidance for the assembly of EFW-A-338,339 & 340 to ensure consistent, proper reassembly'of these valves. (Maintenance) 2. Valves EFW-A-338,339 & 340 should be marked to provide external indication of disk position, seat position, and correct flow orientation. (Maintenance) 9 EXHlBIT 3 M 1 - 9 6- 04 O, PAGE [A8 0F /f/ PAGE =,

3. A representative sample of maintenance procedures whicn hand off to other C procedures to complete a maintenance activity should be reviewed to determine overall adequacy of maintenance procedure interfaces. (Maintenance) T ( -{ CASE ND. 1-96-040 EXHlBIT 10 PAGE[d/ OFl[/_.PAGE(S)

l Causal Factor E: Procet.ure 611-1, instrumentation and Controls Valw Calibration and Checks, provides inadequate guidance to ensure consistent proper reassembly of EFW-A-338, 339, and 340. - Facts: 1. Procedure 5-55-10 leaves the valve closed position prior to assembly in the piping. 2. The actuators for EFW A-338,339 & 340 are in the "open" position wt.en they are attached to the valves. 3. The mechanical configuration of EFW-A-338,339 & 340 and their actuators allows reassembly with the valve disk either in the corteet orientatioit or sidsaligini by ISO 4 The position of the seat and disk for EFW-A-338,339, and 340 can not be readily determined by visual inspection aner the valve have been assembled. There are no permanent external markings to indicate valve disc position (during repair actisity for EFW-A-340, W.O. 94-3025 disc shan was marked). 5. The "as found" condition for EFW-A-338 on Aug.1994 found the disc was 180* out from the correct orientation. (W.O. 94-3008) 6. Rawd on information obtained during interviews, I&C technicians were uncertain as ( to the proper application of procedure 611-1 to the EFW valves. Guidance con =rning checking the sent/ disk for proper orientation was considered particularly confusing. Recamnwnded Action to Prevent Recurrence: a Provide improved guidance for the assembly of EFW-A-338,339 & 340 and similar valves to ensure consistent, proper installation of the valve operators for these valves. (Maintenance) CAstno. ) - 9 6 - 0 4 0 { EXHIBIT 3g PAGE/0> OF$PAGE(S

Causal Factor Cs inconsistent use l Procedure 611-1, instrumentation and Controls ( - Valw Calibration and Oecks, and Technical instructions by 1&C Technicians to perform seat checks and wrification of min stroke / calibration may how contributed to EFW-A-101 seat leakage and significant rewrk. Facts: 1. On Newember 23,1991 EFW-A-101 was identified as leaking by it's seat. A seat che.k was performed (actuator travel was not adjusted). A verification of valve stroke / calibration was not performed. 2. On May 9,1994 EFW-A-101 was identified as not closing when it's position controller was set to close the valve. Air set was adjusted and a valve strokeJealibration was performed. However, a seat check was not performed. 3. On August 4,1994 EFW-A-101 was identified as leaking by at approximately 75 GPM. A seat check was performed and adjustment made to actuato.r travel. A -verification of valve stroke / calibration was not performed. 4. Or. August 11,1994 EFW-A 101 was identified as leaking by at approximately 36 GPM. Repair activities included a seat check, regulator adjustment, bench set (, check / adjustment, and a verification of valve stroke / calibration were performed. Valve leakage was reduced to 0 GPM. 5. Numerous steps in of procedure 6-11-1 performed under WOs 91-7584 and 94-3070 were NA'd without written justification being provided. Recaran=nded Action to Prevent Recurnace: Clear guidance needs to be provided to I&C Technicians for adjusting air operated valves. (Maintenance) l L h CAS N - 1-96-040.. EXHlBIT 3 PAGE/03 OFMPAGE(S)

t Caisaal Faeter D: 'the Work Order pmcess did not ensure that post maintenance Funedonal Test lastruedons for EFW A 101,338 & 340 adequately J ~ tested the abuity of these valves to perform their safety functions. j Facts: 1. Procedure 016-3, Work Order Process, requires tiaat post maintenance Functional Test Instruedons (Fr!), " Verify the abuity of a component or system to perform its intended funedon.' and ' Demonstrate that the original denciency has been correcud.' i In spite of this guidaacs, several W.O.s for EFW A 101,338 4 340 failed to specify adequais Ametional testing. 2. In 1992 W.O. 921746 was initiated to correct EFW A 338 valve leakage. The FT! mamar4a w with this Work Order did not verify the ability of the valve to isolate now. 3. The Fris manar4aw with W.O. 94-3007 (EFW.A 333 valve leakage), did not verify the ability of the valve to isolate now. 4. 'the FTIs associated with W.O. 94 3025 (EFW A 340 valve leakage), did not verify the ability of the valve to isolate flow. 5. 'The FTis associated with W.O. 91-7584 (EFW A 101 valve leakage), did not verify the ability of the valve to isolate flow. (. 6. 'The FT1s associated with W.O. 94-1997 (EFW A 101 not completely shut), did not verify the ability of the valve to isolate now. 7.

  • !he Fris associated with W.O. 917534,94-1746,94-1997,94-3007, and 94 3025 renect IST requirements for the associated valves. (e.g., valve stroke & external j

lenkage) 8. During numerous interviews and informal discussions individuals stated that until recent events indicated otherwise, they believed that specifying the functinaal testing required by the IST program for a given component, ensured adequate post maintenance funedonal testing of the component's safety functions. 9. A PF.D review of 32 recent randomly selected Work Onlers idendfied no instances of indequate functional testing. FW Actlene to Prevent Recurreneet 1.- Individuals responsible for specifying, reviewing, and approving W.O. functiot sl test requirements should be sensitized to the fact that specifying the component testing required by the IST program does not necessarily ensure adequate post maintenance =1 [- functional tesdng. (PED. Operations, Maintenance) EMBT 3 CASE NO. 1 96-040 13 - PAGE$0F PAGE(S)-

0 2. PED's review of recent Work Orders indicates that the functional testing specified in rooset W.O.s rnay be adequate. However, the review performed w s limited in C. scope when compared to the number of W.O. generated annuahy. In addition, at least four recent W.O.s, 721746, %1997,94-3007, and 94-3025 provided inadequate functional testing 'Iberefore, it is recommended that an hulependent review be performed of a representative sample of W.O.s to more accurately assess overall adequacy of W.O. functional testing. (QPD, PED) t b t P CE NO. 1 - 9 6 - 04 0 c EXHIBIT 3 34 PAGE /M OF8/_.PAGE(S

Causal Factor E: Existing periodic testing programs (surveillance / IST) do not ensure C that all design funedons of EFW A 101,201,301,338,339 & 340 are ) periodically verified. Facts: - i d 1. '!he process used by the IST program to determine femtional test requ rements oes not ensure that the ability of EFW A 101,201,301,333,339 & 340 to isolate Emergency Feedwater flow is periodically verified. I 2. Maine Yankee's IST program defines Category A valves as, ' valves for which seat leakage is limited to a specinc maalmum amount in the closed position for fulfillment of thr:ir safety function.' 3. The definitior for Category A valves is interpreted within the Performance Engineering Group as applying only to valves for which a specific maximum allowable leak rate (e.g., less than 20 GPM) is specined in the FSAR, Technical Specifications, the IASD, or a Design Basis Summary Document. *Ihl interpretation is informal and not documented within the IST program descripdon, or memorandum. 4 Maine Yankee's IST program defines Category B valves as, ' valves for which seat leakage in the closed position is inconsequential for fuLTiment of their safety function." 5. Maine Yankee's IST program classifies EFW A 101,201,301,338,339 & 340 as Ca'egory B valves. 7. Category B valves are thot checked for seat leakage under the IST program. 8. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EFW isolatiord feed regulating valves. 9. 'Ihere is no other pmgram which requires periodic testing of EFW A 101,201,301, 338, 339 & 340 for seat leakage. Reca====d=t Actions to Pment Recurrence: 1. PED should evaluate the appropriateness of classifying EFW A 101,201,301,338, 339 4 340 as IST Category B valves. (PED) 2. Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. (PED) 3. If recommendation 2 is adopted, the IST program should be reviewed to ensure - C-valves which meet the revised criteria are included in the program and properly ' classified. (PED) EXHIBIT 3 . CdSE NO. 15

1. 9 6 N 0,

PAGE /M 0F PAGE(S) 4 ,-,,,w-m ,..,-.,m.,,,, m v~,r,- m ,,,-...ne, n, ,,..nw

4. To ensure consistant classification, and valid audit results, IST program description should be revised to inchwie the approved interpretations of those criteria used to ( classify valves as Category A, B & C. (PED) 5. Consideration should be given to establishing programmatic requirements which ensure that safety significant functions of plant components are periodically verified. Steh verification could be implemented as pan of our surveillance program, IST program, or some other program. (PED) i ( { CASE NO. 1-96-040 EXillBIT 3 i PAGE/d2_OF4 PAG -,,, - ~. -, - - - - -, - - -..... - - - - - - - -, - - ~

. _ ~ Causal Facter F: The EDCR process used by Maine Yankee circa 1963 84 did not ensure ( all safety sign @ cant design Jimetions of EFW A-338, 339 & 340 were wr$ed post installation. Facts: 1. RR 83 29, to install EFW A 338,339 & 340 was prepared by YNSD. I 2. 'Ihe revision of YNSD procedure WE 100 in use during 1983 required verification _ that acceptance criteria L+=-prH into design documents were sufficient to verify the design requirements (design fur +1aan) had been satisfactorily accomplished. 3. A basic design funedon of the valves lanallat under EDCR 83 29 (EFW-A-338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident i conditions. 4.- EDCR 83 29 did not require either leak testing nor valve stroke timing with the EFW pumps in operation. l 5. It is practical to perform leak testing and stroke timing with the EFW pumps in j OPemdon. l 6. Post installation functional testing on EFW A 338,339 & 340 was performed in ( accordance with Implementing Instruction 83 29 5, Aux. Feodwater System Functional Testing, which was dcv@ by Maine Yankee using the acceptance criteria provided in EDCR 83 29. The funedonal testing specified did not include either leak testing nor valve stroke timing with the EFW pumps in operation. 7. -Ihe current revision of Procedure 17 212, Engineering Design Change Roquest, requires verincation of design functions by functional testing unless such testing is not practical. 8. A PED review of 4 recent, randomly selected EDCRs identified no instances of inadequate functional testing. 9. The current EDCR process appears to contain sufficient guidance to ensure that appropriate post inmattation functional test is specified. Rec==== dad Actles to Prnest kaanrrence: PED's review of recent EDCRs provides an udcahon that the funcuonal testing specified in recent EDCRs may be adequate. However, the review performed was limited in scope. It is recommended that independent review of a representative sample of EDCRs be perfortned to [. more accurately assess the overall adequacy of EDCR funct onal testing. (QPD, PED, CED) i " 80 3 CASE ND..}. 96-040 17 PAGE /4t!I_OF/f/.PAG

l Caused Vaetoe Gt Weaknesses in our self assessment proce t resulted in sewral missed C opportunities to ident(fy/ correct deficiencies which contributed to this eWnl. Facts: i 1. EFW-A-338, 339 & 340 faded functional testing under EDCR 83 29 due to leaking gaskets. 2. Work on EFW-A-33'8 under Work Or:ler 92-1746 inusited 3* ga% for both flanges. A 2%" gaskt is required f.r the seat, or downstream flange (this fact is based on interview results). 3. Work Order 921746 was not revised to reflect removal and reinstallation of valve to correct body orientation and the use of inconcet gaskets during the initial attempt to reassemble the valve. 4. As a resuli of Work Order 921746 for EFW A 338, DCR 92-170 was generated to mrrect the vendor drawing 7.69-46 and change Mfr. Inst. Manual C-9-2 to redect gasket orientation. 5. W.O. %3025 for EFW A 340 installed gaskets in reverse orientation, causing the valve to leak externally. 6. A copy of the vendor drawing 7.69 46 could not be located in the %3025 work package for EFW-A-340. 7. For identical work: o Step 5.2.1 of Procedure 611 1 was N/A'd for talve El W A-338 under W.O. %3008, whereas signed completed for EFW A 340 under W.O. 94-3025. o Step 5.7.1 was N/A'd completely on W.O. 94-3008, but one sub step signed as completed on 94-3025, 8. The required documentation for Step 5.2.2.b of Procedure 611-1 is missing from W.O. 94-3025. 9. Step 5.2.2.a of Procedure 611-1 for W.O. 94-3025 was signed off as complete. However, there are no manufacturer's position marks identified in the vendor manual / drawings to verify. 10. No completed copy of 6-11-1 could be found for W.O. 921746 (EFW-A 338 actuator remova)/reinemitation). f 11. Design change information fror.1 EDCR 83 29 (i.e. flange gasket orientation and A welded disk key to shaft) was not incorporated into Mainteaance instructions for the 18 EXHlBIT 3 U.S E W. 't 4 s. ~ - PAGE /y 0F/P/ _ PAGE(S

valve, i.e. Procedure 5-5510, nor into the instruction manual C 9-2. A second opportunity occurred in 1992 under W.O. 921746, DCR 92170 chaged technical i manual instructions (C-9-2), and the vendor dwg. 7.69-46 to include gasket i orientation instructions. 'Ihis information was not incorporated into Msintenance instructions. 12. Inadequate post modifie. tion funedonal testing of EFW isolation and regulating valves went undetected for at least ten years. 13. EFW isolation and regulating valve leakage was not identified as a possible safety concern when leakage was noted in November 1991, and May 1994. l t 14. Post maintenance functional testing did not test essential valve isolation functions, 15. In numerous instances steps of procedure 611-1 were NA'd v.3hout documented justification. Recosamended Action to Prevent Recurreneet Maine Yankee's self assessment process should be reviewed to identify inadequacies which contributed to the falltue to previously identify and correct th: deficiencies identified in this report. (QPD, CED, PED, Maintenance, Operations) ( Some specific areas where self assessment appears to have been less than adequate include the following: Worker feedback of field experience to maintenance procedures and W.O.s (Maint.) o Supervisory review of maintenance documentation / procedure adequxy (Maint.) o CED selection / review of EDCR functional test requirements (CED) o o PED selection / review of W.O. functional test requirements (PED) Operations review of EDCR/W.O. functional test requirements (OPS) o QPD selection of QC inspection hold points / acceptance criteria (QPD) o o. QPD oversight of Maine Yankee's Self Assessment Program (QPD) "N-1 - DU- 04 0 19 EXHlBIT 3 PAGE/,/Q__OFJgj___ pgaggg

/ Causal Factee Hs The process used by Maine Yankee to ensure consistency betwen our C Sqfety Analysis and plant perfbrmance did not ensure that: 7he sqfety)knctions of EFW A 338, 339 & 340 wie adequately o wrtfled during post ins;allation testing. The sqfery)knction ofEFW A 101, 201, 301, 338, 339 & 340 to o isolate)fow wre periodically witfied. The sqktyJknctions of EFW A 101, 338 & 340 wre adequately o tested during past maintenance testing. Facts: 1. A basic design function of the valves installed under EDCR 63 29 (EFW A-338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) u,, der certain accident conditions.. 2. Maine Yankee's Safety Analysis for Main Steam Line Break assuones EFW flow is terminated by closure of the EFW isolation /foed regulating valves. 3. EDCR 83 29 did not require either leak testing nor valve stroke timing with the EFW ptmps in operation. 4. De pon inatallation functional testing performed on EFW A-338,339 & 340 under Imple nenting Listruction 83-29-5, dki not include either leak testing nor valve stroke timing with the EFW pumps in operation. 5. Maine Yankee's IST program classifies EFW A-101,201, 301, 338, 339 & 340 as Category B valves. 1 l 6. Maine Yankee's IST program defines Category B valves as, " valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety function.' 7. In 1992 W.O. 92-1746 was initiated to correct EFW A 338 valve leakage. 'Ite FT1

==arw inead with this Work Order did not verify the ability of the valve to isolate flow. 8. The Fris==arw imaad with W.O. 94-3007 (EFW A-338 valve leakage), did not verify , the ability of the valve to isolate flow. L 9. De FTIs associated with W.O. 94-3075 (EFW A 340 valve leakage), did not verify the ability of the valve to isolate flow. i 10. The FTIs associated with W.O. 91-7584 (EFW A-101 valve leakage), did not verify g( the ability of the valve to isolate flow. i 0._ ]. 9 g. 04 0 20 exsisir_ y ' PAGE//.,L 0FB/-.PAGEF

f i !1. 'the Fris associated with W.O. 94-1997 (EFW A 101 not completely shut), did not { verify the ability of the valve to isolate flow. i Recoaumended Action to Prevent Recurrences Management should continue to emphasias/ support the timely completion of the action items identified in the QPD, NSEG report titled ' Maine Yankee Safety Analysis Inputs & ~ Assumptions Review Project Final Report', dated July 8,1994. (Management, RE) e ( f-CASENO. 1-96-040 21 EXHlBIT-3 PAGE //1CFh/._ PAG 5(g

t Coachusboat 7he diskJbr F2W A 338 mar mis positioned during perfbrmance of W.O. 92-( 1746 (March 1992) Facts: 1. 05/14/84 Proper vsIve assembly and orientation verified by the cognizant field engineer for EDCR 83-29 2. 05 26-87 Stroke time testing parformed on EFW-A-301 A 340 under conditions repressatative of a main steam line rupture. Stroke dmes within acceptance criteria. Test report noted that the valves demonstrated.

  • very good shutofP (0 2 GPM).

3. 11 23-91 Lankage rate idendfied on W.O. 91-7585 was 612 GPM with EFW A-338 shut. W.O. 92-1746 generated for repairs. 4. 03-30-92 Completed repairs to EFW-A 338 under W.O. 921746. Valve body was initially reinstalled upside down but retumed to the correct orientation prior to close out of the Work Order. No seat leakage test was performed. 5. 08-04-94 Lankage rate identified on W.O. 94-3007 was 300+ GPM with EFW-( A 338 shut. 6. 0845 94 Completed repairs to EFW A 338. Leakage test sadsfactory. 7. 08-10 94 Leak test of EFW A-339& 340 were 17 GPM and 36 GPM respectively. 8. Review of equipment history (both mechanical and I&C) revealed that EFW A 338 was the only valve to be disassembled (or have actiator removed). EFW-A 339 has not t<.en dinawinbled since original inmilation. EFW-A 340 was dinawenbied under W.O. 94-3025, after the leak test performed on 08/10/94. 9. A combination of botting pattern, and mechanical interference makes it impossible to install the actuator on EFW A-338 with ti e valve body upside down. tAstNo. 1-967040 EXHIBIT 1_ 22 PAGED 3 0FAPAG'

L A1TACHMDU A Es+n8 EFA1M WW 102 p,g psome ps+sM Wem W N SA m., P 380 -C4-- # S1 Test As Found Results As Left SQ Valvemclosed 08/04/94 08/10/94 08/11/94 1 EFW-A 101 75 GPM 36 GPM 0GPM EFW-A 338 400 GPM 20 GPM 24 GPM 15 GPM 0GPM EFW-A 101 & 338 0GPM 0GPM EFW A-101,338 & EFW-102 0GPM 0GPM 2 EFW-A 201 17 GPM 17 GPM EFW-A 339 0GPM 0GPM EFW-A 201 & 339 0GPM 0GPM EFW-A 201,339 & EFW-202 O GPM 0GPM 3 EFW-A-301 36 GPM 23 GPM EFW-A-340 4 GPM 0GPM EFW-A 301 & 340 0GPM 0GPM EFW-A 301,340 & EFW-302 ( 55ENO. 1-964040 EXHl817 3 ~ 23 PAGEMOF.8/ pag

5 k k j i. I 1 p .]'.; 8 [ t l f i u.i i i I i i 7 !I )I) 4!1 i 4 ! _y-1 .i r .~ l 3,, g;l e r I I I [ ( ] g. (( iw i 7 i I I l f s} m t I ie. o. 0 4 E I ' fE f a E m 7 ._1 g i i {]I 43,] m t k>g. l!ki l}j}I 1 aLEI i i I I t k I N EXMIBIT 3 CASEE 1-96-040 i PAGE]hf_.0F$PAGE(! ( )

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=-- to 7 To: G. M. Leitch, Vice President Operations Via: R. W. Blackmore From: G. N. Stowers a al Plant Root Cause Evaluation Report #190 Emergency Feedwater Valve Leakage PRCE Team Members: George Stowers - Nuclear Safety Specialist, NSEG Lyndon Barron - Performance Engineer, PED James Taylor - Senior Nuclear Safety Engineer, NSEG Event Date: August 5,1994 Report Date: October 6,1994 CASENO. 1-963040 EXHIBli 3 PAGE_/Z2._OFj/_ PAGE(S) l L

1.0 CHARTER Charter provided by Manager, PED and endorsed by Vice President, Operadons 1. Identify causal factors contribudng to the Emergency Feedwater Valve L*=h e event described in IJcensee Event Report 94 016, Emergency t Feedwater Isoladon Valve Leakage. 2. Provide recommendadons for reducing the probability of such an event recurring in the future. 3. Assess potendal generic implications of any deficiencies noted. ( 4 C A 7 -9G J 04 0 EXHlBIT 3 2 PAGEKOF_lf/_ PAGE(S)

2.0 EXECUTIVE

SUMMARY

C. At 1220, on August 4,1094 with the reactor in a cold shutdown conditica, plant operators determined that an Emergency Feedwater isolation valve for #1 Steam Generator was leaking by. Further investigation also identified leakage in the Emergency Feedwater (EFW) supplies to #2 & (3 Steam Generators. Subsequently it was determined that under accident conditions which require isolation of Emergency Fredwater, isolation valve leakage could exceed Safety Analysis assumptions. Maintenance activities were initiated to reduce Emergency Feedwater valve leakage. In addition, administrative controls were implemented to ensure Emergency Feedwater leakage is maintained within the bounds of Safety Analysis assumptions dudng accident conditions. A tuview of this event has revealed the following causal factors: Maintenance inst uctions lack sufficient guidance to ensure consistent proper o r.ssembly of the EFW isolation valves. An erroneous belief that specifying the functional testing required by the IST o program for a given component, ensures adequate post maintenance functional testing of all the component's safety functions. The EDCR under which the EFW isolation valves were installed did not o ( require verification of the ability of these valves to isolate EFW flow as part of post installation functional testing. Weaknesses in our self assessment process resulted in s::veral missed o opportunities to identify / correct deficieccies which contributed to this event. EFW flow isolation assumptions in our Safety Analysis are not clearly o prer.cnted in design basis reference documents. To summarize recommended actions to reduce the probability of recurrence: Maintenance procedures should be revited to provide additional guidance for o assembling the EFW isolation valves to ensure consistent, proper orientation of thr. valve disk, body, seat, and actuator, o Action should be taken to increase awareness among those who develop, review, and approve functional tests, that the functional testing required by our IST program should not be relied upon to ensure adequate post maintenance functional testing, The EDCR, Work Order,9d IST processes should be reviewed and o enhanced as necessary to e.,ure that functional testing specified by these C. programs consistently verifies the ability of components to adequately perform their intended safety functions, WL 3 ~. > ' V ).,9q p g n IL JAL PAGE

A broad scope review of EDCRs and W.O.s should be conducted to determine o (. the overall adequacy of functional testing performed at Maine Yankee. i Maint,. Yankee's self assessment process should be reviewed, and e,thanced as o necessary, to address inadequacies which contribute'd to the failure to pmiously identify and correct se deficiendes identified in this report, Management should continue to emphasize / support timely completion of the o action items identified in the QPD, NSEO report titled

  • Maine Yankee Safety Analysis Inputs & Assumptions Review Project Final Report", dated July 8, 1994.

( C peerr , - 9' 6- 04 0 EXHIBIT 3 4 PAGEA_0__OFj/,,,, PAgg(g

3.0 RECOMMENDED ACTIONS TO PREVENT RECURRENCE 1. Provide improved guidance for the assembly of EFW A 338,339 & 340 to ensure consistent, proper reassembly of these and similar valves. ( See Causal Facters A & B) 2. Valves EFW A 338,339 & 340 should be marked to provide external indication of i disk position, seat position, and correct flow orientation. ( See Causal Factor A) 3. A representative sample of rnalntenance proceduits which hand off to other procedures to complete a maintenance activity should be reviewed to dectmine overall adequacy of maintenance procedure interfaces. ( See Causal Factor A) 4. Clear guidance for adjusting air operated valves needs to be provided to I&C Technielans. (See Causal Factor C) 5. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifyin;; the component testing required by the IST program does not necessarily ensure adequate post mainter.ance funedonal testing. ( See Causal Factor D) 6. An independent review should be performed of a representative sample of W.O.s to more accurately assess overall adequacy of W.O. functional testing. (See Causal ( Factor D) 7. Consideration should be given to establishing programmatic requiremats which ensure that safety significant functions of plant components are periodically verified. Such verification could be implemented as part of our surveillance program, IST program, or some other progiaan. (See Causal Factor E) 8 PED should evaluate the appropriateness of classifying EFW A 101,201,301,338, 339 & 340 as IST Category B valves. (See Causal Factor E) 9 Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide flow isolation. (See Causal Factor E) 10. If recommendation 9 is adopted, the IST program should be reviewed to ensure valves which meet the revised criteria are included in the program and properly classified. (See Causal Factor E) 11. To ensure consistent classification, and valid audit results, IST program description should be revised to include the approved interpretations of those criteria used to classify valves as Category A, B & C. (See Causal Factor E) ( 12. An independent review of a representative sample of EDCRs be performed to more EXHlBIT 3 PAGES/_0C ff/_ PAGE(S CASEhD. 1-96-01()

_ accura'aly assoas the overall adequacy of EDCR funcdonal tesdng. (See Causal Factor D 13. Maine Yankee's self assessment process should be reviewed to identify any inadequacies which may have contributed to the failure to previously identify and correct the deficiencies idendfied in this report. (See Causal Factor G) I 14. Management should condnue to emphasize / support the dmely compledon of the action items idendfled in the QPD, N',EG report dtled

  • Maine Yankee Safety Analysis Inputs & Assumpdons Review Project Final Repart", dated July 8,1994. (See Causal Factor H) 15, Management should continue to stress the importance of procedure compliance /

attention to detall. (See: Causal Factor C, Fact 5; Causal Factor D, Fact 1; Causal Factor G Facts 8, 9, 11, & 15 k CASE NO. ] g n - n3 n p can v.. 6 PAGE]J_E_OF./f/_ PAGi(S) l

4.0 NARRATIVE Maine Yankee's Emergency Feedwater (EFW) Isoladon Valves EFW A 338,339 & 340 were installed under EDCR 83-29 in May of 1984. 'Ihe purpose of this modincadon was to provide redundant isolation of EFW to affected Steam Generators during a Main Steam Rupture event. Post installadon tesdng of these valves verined: system integrity, valve stroka times, and the capacity and integrity of the valves air operating system. 'Ihe ability of l 1 the isoladon valves to stroks under design conditions of pressure and flow and their ability to actually interrupt flow was not tested. As design basis documents did not specify a specific maximum allowable leak rate for these valves, they were classined as IST Category B l valves. In accordance with the provisions of Maine Yankee's IST program for Category B l valves, periodic testing requirements included only verination of stroke time and position indication. I In 1987 during development of a Design Basis Summary Document for the EFW system, it was noted that the ability of the valves to stroke under design conditions had not been verified during the post installation testing specified by EDCR 83 29. Follow up testing conducted on May 5,1987 confirmed satisfactory stoke times for EFW-A-301 & 340 under design conditions of pressure and flow. Although a determination of the leak tightness of these valves was not a test objective, it was noted in the test report that the valves exhibited "very good shutoff" (0 2 GPM). i In November.1991 operations noted leakage past EFW A 101 & 338 and initiated work ( orders for their repair. In March,1992 EFW A 338 was disassembled and the cause of the leakage was determined to be crosion of the seat. (It is believed that the disk for EFW A-338 was installed incorrectly at this time). An inspection of EFW-A 101 conducted during the same time frame confirmed proper seating of the valve. Post maintenance functional testing for both valves consisted of a stroke test and a check for external leakage. In May,1994 a work order was initiated by Operations against EFW-A 101 due to the valve not fully closing. Repair actions consisted of increasing the output of the air regulator supplying the valve operator from 32 to 35 PSIG. It was noted on the work order that regulator output could not be increased much further. Post maintenance testing consisted of stroking the valve. On July 16,1994 a plant shutdown was performed due to Steam Generator tube leakage. ~ While Steam Generator levels were being maintained using the Emergency Feedwater system, plant operators noted a gradual, unexpected increase in #1 Steam Generator level. Suspecting leakage past the EFW feed regulating valve for #1 Steam Generator (EFW A 101) as the source of the unexpected increase, operators redirected EFW flow through the main feed system and isolated the EFW feed regulating valves. An attempt to recreate the apparent leakage later in the shift was unsuccessful and the source of the unanticipated level increase remained unidentified. On August 4,1994 Maine Yankee was in a cold shutdown condition making preparations to l( restart the plant following completion of Steam Generator inspections and repali At approximately 1220. operators conducted a EFW feed system leak test to follow up on th: CASE ND. 1 9(;*040 7 ( at./M_48/_ PAGE(S) \\:. I ii, _,._ -,, _?, n._,.,_--,_,._..,_-_.___....

\\ i maamallas rated on July 16. During this test it was determined that the Emergency C-Pendwater Isoladon Valve for #1 Steam Generator (FEV A 338) tankad by at a rate of l about 400 GPM. It was also noted that the Emergency Feedwater Regulating Valve for #1 Steam Generator (EFW A 101) leaked by at a ra:e of approximately 75 GPM. Subsequent j inva:tigrian determined that the actuator for EFW A-338 was coupled to the disk aoproalmstely 180 degrees out of alignment. His misalignment is believed to have occuned during valve maintenance performed during our 1992 Refueling Outage. Maintenance -l activity was initiated to reposidon the actuator / disk to the correct orientadon. Imakage past EFW A 101 was conected by adjusting valve stroke. i At 17M on August 5,1994 it was determined that Maine Yankee's Steam Line Break Safety Analysis assumes zero leak.gc past Ere.srgency Fealwater isolation and regulating valves. Therefore, c 1957 the NRC was appnsed of the situation via the Emergency Notification System in accordance with the provisions of 10 CFR 50.72 (b)(2)l. A follow up notification wr.: made at 1251 on August 6,1994 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW A 340. This leakage was anbs xtuently determined to be caused by seat / disk erosion. On August 10, l994 comprehensive testing was performed oa the Emergency Feedwater isolation and regulating valves for au thne Steam Generators using PED test equipment. Dese tests rt-t'ad leak rates for various individual valves and combinations of valves from O to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studie conducted in parallel with these tests established *he following revised acceptance criteria: 1) Leskage past individual isolation / regulating valves should be less than 40 GPM. De purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with fauure of either the isolation valve or regulating valve to close. 2) Leakage past both the isolation and regulating valve to each Steam Generator should be less than 10 GPM. De purpose of this limit is to provide at least 30 minutes for operators to insure EFW is itclated to a faulted Steam Generator 'ollowing a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to overful. 3) Zero leakage to a Steam Generator with a tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due to overfill following a Steam Generator Tube Rupture. In order to meet the revised acceptance criteria, EFW-A 101 was adjusted to reduce leakage past EFW-A-101 & 338 to zero GPM and administrative controls were implemented to ensure termination of any potential future leakage to a ruptured Steam Generator within 30 -

minutes using manual isolation valves. Comprehensive *as left" leak testing conducted by PED on August 11,1994 confirmed that EFW isolation / regulating valve leakage was within the revised acceptance criteria.

( QSE ND. 1-96-040 EXHIBIT 3 PAGE /g 0F3/__FAGE(S)

5.0 CAUSAL FACTORS, SUPPORTING FACTS, AND RECOMMEND ACnONS TO PREVENT RECURRENCE Causal Factor A: Procedure 3 5510, Maintenance of EFW Air Operated Trip Valves, does not provide adequate guidance to ensure consistent proper assemMy ofEFW A-338, 339 & 340. Facts: r 1. In March 1992, following maintenance performed under W.O. 921746, EFW-A 338 was initially reinstalled with incorrect valve body orientation and incorrect gaskets. 2. Procedure 5-5510 leaves the valve in the closed position prior to assembly in the piping. 3. De actuators for EFW-A 338,339 & 340 are in the "open" position when they are attached to the valves. 4. The meclianical configuration of EFW-A 338,339 & 340 and their actuators allows reassembly with the valve disk either in the correct orientation or misaligned by 180 degrees. 5. The position of the seat and disk for EFW-A-338,339, and 340 can not be determined by visual inspection after the valves have been assembled. There are no C permanent extemal markings to indicate valve disc position (during repair activity for EFW-A-340, W.O. 94 3025 disc shaft was marked). 6. Procedure 5 55-10 does not provide guidance for rotating the disk to the correct orientation prior to installing the actuator. 7. The "as found" condition for EFW-A 338 on Aug.1994 found the disc was 180' out from the correct orientation. (W.O. 94-3008) 8. There was considerable rework involved in completing reinstallation of EFW A-340 under W.O. 94-3025. In summary: o The valve body was installed 180' out of position, o ne flange gaskets were installed incorrectly, o The valve was torqued in the "open" position, causing binding. Recommended Actions to Prevent Recurrence: 1. Provide improved guidance for the assembly of EFW A-338,339 & 340 to ensure consistent, proper reassembly of these valves. (Maintenance) 2. Valves EFW-A 338,339 & 340 should be marked to provide external indication of ( disk position, seat position, and corre::t flow orientation. (Maintenance) 9 EXHIBIT 3 CASE ND. ] -9'6 - 04 0 PAGE/M OFjf/_PAGE(S)

3. A spressatative sample of maintmar* procedures which hand off to other C procedures to complete a maintenance activity should be reviewed to determine overall adequxy of maintenance procedure interfaces. (Maintenance) \\ \\ 9 A M 7-96-040 EXHlBlT.$ PAGE/M. OF_f//__ PAGE(S) ,g

Cameal Easter Et haceda 6111, lasruwnentadon and Controls Valw Calibration and - Checks, prmWs inadequate gaddance so ensure consissent p oper ,w-29 qfEFW-A 338,339, and 340. Facts: 1.. Procedurs 5 5510 leaves the valve closed position prior to assembly in the piping. 2. The actuators for EFW-A 338,339 & 340 are in the "open" posidon when they are attached to the valves. 3. 'Ihe mechanical configuradon of EFW-A 338,339 & 340 and their actuators allows reassembly with the valve disk either in the correct orientadon or misaligned by 180 degrees. 4. The position of the seat and disk for EFW-A 338,339, and 340 can not be readily determined by visual inspection after the valve have been assembled. There are no permanent external markings to indicate valve disc position (during repair activity for EFW A 340, W.O. 94 3025 disc shaft was marked). 5. The *as found" condition for EFW A 338 on Aug.1994 found the disc was 180' out from the correct orientation. (W.O. 94 3008) ( 6. Based on information obtained during interviews, I&C technicians were uncertain as to the proper appilcation of procedure 6-11 1 to the EFW valves. Guidance concerning checking the seat / disk for proper orientation was considered particularly confusing. Recommended Action to Prevent Recurrence: Provide improved guidance for the assembly of EFW-A 338,339 & 340 and similar valves to ensure consistent, proper installation of the valve operators for these valves. (Maintenance) a 1 CASENO.. 1-96-040 EXHIBIT 3 PAGE /}7 OF/f/_.PAGE(S)

Causal Yator Cs inconsistent use of Procedure 611 1, instrwnentation and Controls ( Valw Calibration and Check, and Technicalinstructions by !&C Technicians to perform seat checks and wrification of min stroke / calibration may han contributed to EFW-A 101 seat leakage and significant rework. Factst 1. On November 23,1991 EFW-A-101 was identified as leaking by it's seat. A seat check was performed (actuator travel was not adjusted). A vetification of valve stroke 1 calibration was not performed. 2. Cn May 9,1994 EIV-A 101 was identified as not closing when it's position controller was set to close the valve. Air set was adjusted and a valve stroke /calibrat on was performed. However, a seat check was not performed. i 3. On August 4,1994 EFW A 101 was identificxi as leaking by at approximately 75 GPM. A seat check was performed and adjustment mhic to actuator travel. A verification of valve stroke / calibration was not performed. 4 On August 11,1994 EFW A-101 was identified as leaking by at approximately 36 GPM. Repair activities included a seat check, regulator adjustment, bench set check / adjustment, and a verification of valve stroke / calibration were performed. (. Valve leakage was reduced to 0 GPM. 5. Numerous steps in of procedure 6-11-1 performed under WOs 91-7584 and 94 3070 were NA'd without written justification being provided. Recommended Action to Prevent Recurrence: Clear guidance needs to be provided to I&C Technicians for adjusting air operated valves. (Maintenance) C 1 - 9 G - 04 0 EXHIBIT 8 12 PAGE/)T OF/f4PAGE(S)

1 ( Causal Factor D: De Work Order process did not ensure that post maintenance A Functional Test Instructions for EFW A-101,338 & 340 adequately tested the ability of these valves to perform their safety functions. Facts: 1. Procedure 0-16-3, Work Order Process, requires that post maintenance Functional Test Instructions (FTI), " Verify the ability of a component or system to perform its intended function.' and ' Demonstrate that the original deficiency has been corrected.' In spite of this guidance, several W.O.s for EFW-A 101,338 & 340 failed to specify adequate functional testing. 2. In 1992_W.O. 921746 was initiated to i;orrect EFW A 338 valve leakale, ne FTI associated with this Work Order did not verify the ability of the valve to isolate flow. 3. The FTIs associated with W.O. 94-3007 (EFW A-338 valve leakage), did not verify the ability of the valve to isolate flow. 4 The FTIs associated with W.O. 94-3025 (EFW A 340 valve leakage), did not verify the ability of the valve to isolate flow. 5. The FTIs associated with W.O. 91-7584 (EFW-A 101 valve leakage), did not verify { the ability of the valve to isolate flow. 6. The FTIs associated with W.O. 94-1997 (EFW A-101 not completely shut), did not verify the ability of the valve to isolate flow. 7. The FTIs associated with W.O. 91-7584,941746,941997,94-3007, and 94 3025 reflect IST requirements for the associated valves. (e.g., valve stroke & extemal leakage) 8. During numerous intervien and informal discussions individuals stated that until recent events indicated otherwise, they believed that specifying the functional testing required by the IST prognm for a given component, ensured adequate post maintenance functional testing of the component's safety functions.

9. -

A PED revie.v of 32 recent randomly selected Work Orders identified no instances of inadequate functional testing. Recommended Actions to Prevent Recurrence: 1. Individuals responsible for specifying, reviewing, and approving W.O. functional test requirements should be sensitized to the fact that specifying the component testing ( required by the IST program does not necessarily ensure adequate post maintenance functional testing. (PED, Operations, Maintenance) EXHIBIT 3 WE NO. 1-96-040 13 PAGE$OF8/_ PAGE(S)

2. PED's review of recent Work Orders indicates that the funedonal testing specified in recent W.O.s may be adequate. However, the review perfonned was limited in C.. scope when compared to the number of W.O. generated annually. In ad'idon, at least four recent W.O.s,92-1746,94-1997,94-3007, and 94-3025 provided inadequate funedonal testing. Therefore, it is recommended that an independent review be performed of a representative sample of W.O.s to more accurately assess overall adequacy of W.O. funedonal testing. (QPD, PED) k e p usm. 1 - 9 0 - 04 o s,,, y PAGE /FJ OF.lf/_ PAGE(S) _g

l l Causal Factor E: Existing periodic testing programs (survedlance/ IST) do not ensure ( that all design fuwtions of EFW-A 101,201,301,338,339 & 340 are periodically verified. Facts: 1. 'Ihe process used by the IST program to determine functional test requirements does not ensure that the ability of EFW-A 101,201,301,338,339 & 340 to isolate Emergency Feedwater flow is periodically verified. 2. Maine Yankee's IST program defines Category A valves as, " valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their safety function." 3. The definition for Category A valves is interpreted within the Performance Engineering Group as applying only to valves for which a specific maximum allowable leak rate (e.g., less than 20 GPM) is specified in the FSAR, Technical Specifications, the IASD, or a Design Basis Summary Document. This interpretation is informal and not documented within the IST program description, or memorandum. 4 Maine Yankee's IST program defines Category B valves as, " valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety function." 5. Maine Yankee's IST program classifies EFW-A-101,201,301,338,339 & 340 as Category B valves. 7. Category B valves are not checked for seat leakage under the IST program. 8. Maine Yankee's Safety Analysis for Main Steam Line Break assumes EFW flow is terminated by closure of the EFW isolation / feed regulating valves. 9. There is no other program which requires periodic testing of EFW A 101,201,301, 338,339 & 340 for seat leakage. Recommended Actions to Prevent Recurrence: 1. PED should evaluate the appropriateness of classifying EFW-A-101,201,301,338, 339 & 340 as IST Category B valves. (PED) 2. Consideration should be given to modifying PED's existing interpretation of the definition of IST Category A valves to include valves which are assumed by our Safety Analysis to provide fa w isolation. (PED) 3. If recommendation 2 is adopted, the IST program should be reviewed to ensure C^ valves which meet the revised criteria are included in the program and properly classified. (PED) EXHIBIT 3 CASE NO. 1 - 9G- 04 0 NEMCF/f PAGE(S)

f% 4. To ensure consistent classification, and valid audit results,.tST prognm description ( should M revised to include the approved interpretations of those criteria used to L classify valves as Category A, B & C. (PED) 5. Consideration should be given to establir.hing programmatic requirements which ensure Qat safety significsnt functions of plant components are periodically verified. Such verification could te implemented as part of our suntillance prog:an., IST program, or some other program. (PED) CASENO. 1-96-040 EXHIBIT PAGE/M OFE/_ PAGE(S) 16

Causal Factor F: The EDCR process used by Maine Yankee circa 1S83 84 did nct ensure c.ll safety significant design functions of EFW-k338, 339 & 340 were venfled post installation. Facts: 1. EDCR 83-29, to inst-11 EFW-A-338,339 & 340 was prepared by YNSD. 2. The revision of YNSD procedure WE-100 in use during 1983 required verification that acceptance criteria incorporated into design documents we.re sufficient to verify the tesign requirements (design functions) had been satisfactorily accomplished. 3. A basic design function of the valves installed under EDCR 83-29 (EFW-A-338,339, 340) is to ensure i olation of EFW flow to faulted SG(s) under certain accident conditions. 4. EDCR 83-29 did not require either leak testing nor valve stroke timing with the EFW pumps ih operation. 5. It is practical to perform leak testing and stroke timing with the EFW pumps in operation. { 6. Post installation functional testing on EFW-A-338,339 & 340 was perform.-d in accordance with Implementing Instruction 83 29 5, Aux. Feedwater Sy: - Functional Testing, which was developed by Maine Yankee using the acceptance criteria provided in EDCR 83-29. The functional testing specified die not include either leak testing nor valve stroke timing with the EFW pumps in o; ration. 7. The current revision of Procedure 17-21-2, Engineering Design Change Request, requires verification of design functions by functional testing unless such testing is not practical. 8. A PED review of 4 recent, randomly selected EDCRs identified no instances of inadequate functional testing. 9. The current EDCR process appears to contain sufficient guidance to ensure that appropriate post installation functional test is specified. Recommended Action to Prevent Recurrence: PED's review of recent EDCRs provides an indication that the functional testing specified in recent EDCRs may be adequate. However, the review performed was limited in scope. It is recommended that independent review of a representative sample of EDCRs be performed to ( more accurately assess the overall adequacy of EDCR functional testing. (QPD, PED, CED) MW-1 - 9 0- Os 0 17 EXHIBIT 3 PAGEM3 CF/p'_PAGE(5)

y. 1 Canaal Facter Gt - Weaknesses in our seVassessment procus rerulted in senral missed ^ L opportedries to identify / correct dq)fciendes alch contributed to this ennt. Facts: 1. EFW A-338,339 & 340 failed functional testing under EDCR 83-29 due to leaking

gukets, 2.-

- Work on EFW-A-338 under Work Order 92-1746 L Mlad 3" gasktts for both flanges. A 2%" gasket is required for the seat, or downstream flange (this fact is based on interview results). 3. Work Order 92-1746 was not revised to reflect removal and reinstallation of valve to correct body orientation and the use of incorrect gaActs during the initial attempt to reauemble the valve. 4 As a result of Work Order 92-1746 for EFW-A 338, DCR 92-170 was generated to correct the vendor drawing 7.69-46 and change Mfr. Inst. Manual C-9-2 to reflect y gasket orientation. 5. W.O. 94-3025 for EFW-A-340 installed gaskets in reverse orientation, causing the valve to leak extemally. 6. A copy of the vendor drawing 7.69-46 could not be located in the 94-3025 work package for EFW-A-340. 7. For identical work: o Step 5.2.1 of Procedure 6-11-1 was N/A'd for valve EFW-A-338 under W.O. 94-3008, whereas signed completed for EFW-A-340 under W.O. 94-3025. Step 5.7.1 was N/A'd completely on W.O. 94-3008, but one sub step signed o as completed on 94-3025. 8. - The required documentation for Step 5.2.2.b of P.wedure 6-11-1 is missing _from - W.O. 94-3025. 9. Step 5.2.2.a of Procedure 6-11-1 for W.O. 94-3025 was signed off as complete. However, there are no manufacturer's position marks identified in the vendor manual / drawings to verir'y. 10. No compSted copy of 6-11-1 could be found for W.O. 92-1746 (EFW-A-338 actuator - removalheinstallation). 11. Design change information from EDCR 83-29 (i.e. flange gasket orientauon and - welded disk key to shaft) was not incorporated. into Maintenance instructions for the 18 EXHIBIT 3 - CASE'NO. 1-96-040 PAGEl.M OFKPAGE(S)

valve, i.e. Procedure 5-55-10, nor into the instrucdos manual C-9-2. A second - C opportunity occurred in 1992 under W.O. 92-1746, DCR 92-170 changed technical manual instruedons (C-9-2), and the vendor dwg. 7.6lM6 to include gasket orientadon instructions, nis information was not inerporated into Maintenance instrucdons. 12. Ink $ equate post modificadon functional testing of EFW lealatim and regulating valves went undetected for at least ten years. 13. EFW isolation and regulating valve leakage was.not identified as a possible safety concern when leakage was noted in Nm h 1991, and May 1994. 14. Post maintenance functional testing did not test essential valva isolation funcdons. 15. In numerous instances steps of procedure 611-1 were NA'd without documented justification. Recommended Action to Prevent Recurrence: Maine Yankee's self assessment process should t e reviewed to identify inadequacies which contributed to the failure to previously identify and correct the de5ciencies identified in this report. (QPD, CED, PED, Maintenance, Operations) h Some specific areas where self assessment appears to have been less than adequate include the following: Worker feedback of field experience to maintenance procedures and W.O.s (Maint.) o Supervisory review of maintenance documentation / procedure adequacy (Maint.) o CED selection / review of EDCR functional test requirements (CED) o PED selection / review of W.O. functional test requiremenu (PED) o Operations review of EDCR/W.O. functional test requirements (OPS). o QPD selection of QC inspection hold points / acceptance criteria (QPD) o QPD oversight of Maine Yankee's Self-Assessment Program (QPD) r o [ EXHlBIT 3 BERL - I. 9 6 0 PAGE85 0FMPAGE(S) 19 i

Causal Factor E: The process used by Maine Yankee to ensure consistency between our C Safer) Analysis and picnt perjJrmance did not ensure that; lhe safetyfunctions of EFW-A 338, 339 & 340 were adequately o ver(fied during post installation testing, The safetyfunc:fon of EFW-A-101, 201, 301, 338, 339 & 340 to o isolateflow were periodically wrified. lhe safetyfunctions of EiW-A-101,338 & 340 were adequately o tested during post maintenance testing. Facts: 1. A basic design function of the valves installed under EDCh 43-29 (EFW-A 338,339, 340) is to ensure isolation of EFW flow to faulted SG(s) under certain accident conditions.. 2. Maine Yankee's Saf:ty Analysis for Main Steam Line Break assumes EFW flow is terminated by cbscre of the EFW isolation / feed regulating valves. 3. EDCR 83 29 did not require either leak testing nor valve stroke timag with the EFW pumps in operation. { 4. The post installation functional testing performed on EFW-A-338,339 & 340 under Implementing Instruction 83-29-5, did not include either leak testing nor valve stroke timing with the EFW pumps in operation. 5. Maine Yankee's IST program classifies EFW-A-101,201,301,338,339 & 340 as Category B valves. 6. Maine Yankee's IST program de. fines Category B valves as, " valves for which seat leakage in the closed position is inconsequential for fulfillment of their safety function." 7. In 199'2 W.O. 92-1746 was initiated to correct EFW-A-338 valve leakage. The FTI associated with this Work Order did not verify the ability of the valve to isolate flow. 8. Tne FTIs associated with W.O. 94-3007 (EFW-A-338 valve leakage), did not venfy the ability of the valve to isolate flow. 9. The FTIs as" ciated with W.O. 94-3075 (EFW-A-340 valve leakage), did not verify the ability of the valve to isolate flow. 10. The FTIs associated with W.O. 91-7584 (EFW-A-101 valve leakage), did not verify ( the ability of the valve to isolate flow. EXHIBIT 3 CASE NO. 1-96-040 20 PAGE/d$_OF/ PAGE(S)

11. 'the FTIs menW with W.O. 94-1997 (EFW-A-101 not completely shut), did not {. verify the ability of the valve to isolJe flow. Recommended Action to Prevent Recurrence: Management shwld continue to emphasize / support the timely completion of the action items identified in the QPD, NSEG report titled " Maine Yankee Safety Analysis Inputs & Assumptions Review Project Final Report", dated July 8,1994. (Managemen', RE) e ( C CASE NO. 1-96-040 EXHlBIT 3 21 PAGE/G OF/f PAGE(S)

== Conclusion:== (. . -11ae disk)>r EFW-A 338.ns mis positioned during performance of W.O. 92-1746 (March 1992) - Facts: 1. 05/14/84 Proper valve assembly and orientation verified by the cognizant field engineer for EDCR 83-29 2. 05-26-87 Stroke time testing performed on EFW-A-301& 340 under conditions representative of a main steam line rupture. Stroke times within acceptance criteria. Test report noted that the valves demonstrated, 'very good shutoff" (0-2 GPM). 3. 11-23 91 Y rahge rate identified on W.O. 91-7585 was 6-12 GPM with EFW A-338 shut. W.O. 92-1746 generated for repairs. 4. 03-30-92, Completed repairs to EFW-A-338 under W.O. 92-1746. Valve body was initially reinstalled upside down but returned to the correct orientation prior to close o of the Wark Order, No seat !:akage test was performed. 5. 08-04-94 1.cakage rate identified on W.O. 94-3007 was 300+ GPM with EFW-(_ A-338 shut. 6. 08 05-94 Completed repairs to EFW-A-338. I.makage test satisfactory. A 7. 08-10-94 Leak test of EFW-A-339& 340 were 17 GPM and 36 GPM respectively. 8. - Review of equipment history (both mechanical and I&C) revealed that EFW-A-338 was the only valve to be disassembled (or have actuator removed). EFW-A-339 has not been disassembled since original installation.- EFW-A-340 was disassembled under W.O. 94-3025, after the leak test performed on 08/10/94. 9. A com'bination of bolting pattern, and mechanical interference makes it impossible to install the actuator on EFW-A-338 with the valve body upside down. [ DSEm y-96-040 ExHisir 3 - PAGE&OF]p__ FIGE(S)

ATTACHMENT A arw+ms erw 6 sel / arw 1st pg -N-- M S.8. w sFw+ ass erw4-m1 P ass

  1. 2 s.a.

ENA 340 EFw A 301 / u w set .. p >H>< -M --tas.t. 4 Test As Found Results As12ft SQ Valve (s) Closed 08/04/94 Q8/10/94 08/11/94 1 EFW-A-101 75 GPM 36 GPM 0GPM EFW-A-338 400 GPM 20 GPM 24 GPM EFW-A4101 & 338 15 GPM 0GPM EFW-A-101,338 & EFW-102 0GPM 0GPM 2 EFW-A-201 0GPM 0GPM EFW-A-339 17 GPM 17 GPM EFW-A 201 & 339 0GPM 0GPM EFW-A-20;,339 & EFW-202 0GPM 0GPM 3 EFW-A-301 O GPM 0GPM EFW-A-340 36 GPM 23 GPM EFW-A-301 & 340 - 4 GPM 0GPM EFW-A-301,340 & EFW-302 0GPM 0GPM ( USEE 1-96-040 EXHIBIT E 23 PAGEdOF PAGE(S)

= w _m ,e aC Q. % Rs ) m mi - t-o. e_. m EFW VALVE LEAKAGE Q. ~ m N ig saio W Sequence Of Events.05/84 -08/94 n<- pasense erass sess ensemamise *"8*"*"8 ed4C te.4aed F***** eest 1 -f-s EpWAsg amane ese an. UWA438 EFWA 338 Ops amee UWA-let &

  1. ~

a" _m 3es aman enume. semesmand ash esmaned as EFWA418.enh E.FW tese.use to 338 esammenned ~ -- gg ,,,n ~ m,..W o.- Wm,,,, e O. 7.SM.L T S. IASO.L 8 .e "y ~ ~~~ an ",,m;~ g; UW - tuuWien me estese een. mese. m.m. g __ spW44ee ammmmm an' EfW44eo E h 3eo N I u.o g .m,,,,,, tenumme espes.pg and W peepeq mammen ses.es inehmes AamlFt* g Y

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~ Proc. No. 20-100-1 Rev. No. 15 Page 6 of 7 ATTACHMENT A REVIEW AND APPROVAL PRCE Submitted by: IM M /~ Author Date Reviewed by PORC on: /O ~ [?" k at Meeting No. PORC Recortaendations: pg - f/[6 con.o M4' & h Y Y Y n /b/0/1f d $ d I f f'E' Approved by: / /b //d /G <-r (- Plant Manager / ' Date I approve this PRCE i lu the recommendations with the exceptions noted below. /P/ 7/9V Approved by: C V.P., Operations tate Exceptions: Return this reoort to the PORC Secretary Assistant for filing and routing to the distribution list (which includes the CMS coordinator). '(- CASC;io. 1. 9 e - 0 4 0 EXHIBIT 3 DAGEg CF$_.PAGE(S)

b i l.a PED-RC-94-014 Page _L of L l .= f C' (ENGINEEK!IIG) ROOT CA43E TITLE: -. [W-A 338/ Erd-A 360 stat LEAKER 3'XJRCE DOCLMENT: _ ID NO: 94-2008/D,4-060 WO8tX ORDER 100R 7tMT SYSTEM: _ EFW TECH FILE No: 35 PLANT coMPOND(T'. 10 NO: N 4-238. 140 VAtvr DESCRIPTION OF PROBLEM ttos ney me w nvereen: F!lt0IMG50 ntm.e n.s co w v e w w - % v cor_ VALIDATION OF IDUlTIFIED Ro0T CAUSE(s): POTENTIAL WLICATIONS cif am

3..a., eetm.,.etMeim:

CONCLUSION (s) cr ,an.s esmettw usw en u, if assunto: EVALUATOR /DATE: h h 6-/J H DISPOSITION: DEPT. MANAGER /DATE:, (Procedure 17-309, Rev. 4, Page 4 of 4)(Attachment A) '( EASENO. 1 9 g. g 4 (, EXHIBIT b // PAGE / OF M PAGE(S) j ,0 'l f f L e m _m.

PED-RC-9a-014 c Page _L of L j CESCAirf!0N OF PROBLDI: Work Order 94 3007 was ger.erated by operations department en 8/4/94. This work order identified excesstve seat leakage of 300 gpa for valve EFW-A-336. The I&C section or maintenance attempted to seat the valve but were unsuccessful and leakage persisted. Work Order 94-3008 was generated by Mechanical Maintenance to disassemble and inspar.t the valve. During the review of "open" work orders on 8I-94, the Perfomance Engineer identified this valve as having Maintenance Aule implicatioy. This information was provided to the Performance Engineering Section Head at the daily sectia meeting. The Perfomance Engineer was assigned the responsibility for completing a departmental root cause, per precedure 17-308. EN-A-335 is one of three valves that isolate emergency feed water to the steam generators. EFV-A-338,339 and 340 were installed under Engineerino Design Change $3-29 during the 1984 refueling outage. The valves were installed to respond to e postulated steam line brcok and prevent feeding the affected steam generator. These valsas are 600 lb, ANSI, carbcn steel, water, butterfly valves, manufactursd, by Contromatics Corporation. The actuators also are manuftetured by Contromatics and are spring' return, piston actuators. They are normally open, and ' fail open" on loss of air to the pisten and/or less of power to the pilot solenoid. FIISINES: August 5,1994 1. EFW-A-138 Discoverad Misnesitioned. In discumston with Pete Mehlhorn, it was identified that EW-A-338 was indicating to be 180* out of position fron "as found' markin{s prior to valve repair. The seat had been removed by Maintenance and provided to PED. It did not show signs of wear. 2. Dine Welded to Shaft Durin$hedisewasfoundweldedtotheshaftforvalveEFW-A-338. inspection of spare parts in W that 3. He nermanent external markimas for EFW A-118 to indienta valva nesitian. Inspection of the a.sambled valve in the Emergency Feedwater Yalve Room revealed no permanent marking visible on the valve shaft,

p. Meh1 horn's pen / pencil mark was the only visible marking.

P. Mahlherr. explained 'as found" markings versus "as left' (See attashed). 4. EFW-A,sta is a ' Failed then? actuat. arm Thi is substantiated both on the vender's drawing 7.79-46 and in EDCR 83-is was further verified by discussions witt F. O'Maal, the 1 & C Tech installed the actue.ter. C b EXHIBIT ~ PAGE - M PFlQ.PAGE(S)

PED-RC-94-014 Page,L 09 L 5. Maintenance instructions elace valve in closed nesitten. Maintenance Proc. s-55-10, reassemoly instructions, step 4.6.7, place the valve in the closed position (see attached). This was further. substantiated by Manufacturer's Instruction Man'ual C-9-2 reassembly ' instructions. 6. The valve aust be rotated to nerait actuator assembly. The valve is in the closed position and the actuator is a ' Fail to Open* actuator. The valve shaft - will not engage with the coupling of the actuator, and is 90* cut of position. Per the manufacturer's instruction manual C-9-2 this is a slightly greater than W rotation actuator. This information was further substantiated in discussions with ths. I & C Valve Tech, T. O'Neal. 7. Correct oositioni9e of the valva is determined by actuatcr direction of rotation. In discussion with the ! & C Tech F. O'Neal, and the Maintenance Mechanic,

5. Greenleaf, It was identified that the valve must M re-positloped to the' nonsal 'Open* position. This norinal "Open' position is 90* oppositt to the actuator rotation.

Tha actuator rotates W clockwise to close and 90* counter clockwise to open. The valve was stroked satisfactorily to verify seat position / actuator rotation. 8. EFM A-133 'As found* condition with disc 110' out from nemal onen Through observations of shaft markings during valve cycling, and -( discussions with l4C/ Maintenance, it was confirmed that the valve had been reinstalled under the previous maintenance activity,180* out of position (See Attached). 9. There is ne seat laakaan evitaria for EFW-A-223. 329 and 240. Review of EDCR 83-and the IST program reveals no leakage criteria. 10. EFW Valves have veauired limited corrective action. A review of equinient history for ETW-A-338 fdentified one corrective maintenance activity that required valve disassembly. Work Onter 92-1746 disassembled and repaired the valve during the 1992 refueling outage. It was noted in the description of work performed 'vlv was in line upside down'. EFW-A-339 and EN-A-340 had not been disassembled since original installation. 11. There are no other similar valves in the MIpPS Daemkana. Review of MIPPS data and vender drawings estabitshed that EFW-A-338, 339 and 340 are unique in their seat / disc to actuator configuration. 12. W.O. 94-200e annetated to refinet actuatar emeratio% keview of W.O. 9+-3006 revealed notes added to step 3 " Valve f tils open and turns clockwise to close, air to close* and " valve in open position as. ( found*. b EXHIBIT PAGE 3 0FM. DAGE(S) _ _ _ _ _ ~ _ _ _

l l PED-R.C-94.-014 p,. 13. ffW-A-338, 339. 360 sent irt one munition or:1v. 1 Review of vendor drawing and spara parts revealed that configuration / construction of seat / disc and shaft crientation is different free more i typical in-line butterfly valses. The disc is more similar to a ball valve than flat type, that can seat in two closed positions (and two open positions, i.e. 90'.) The vertical center-line plane of the disc is the same ar. the seat. For EFV-A-338, 339 and 340, the centerline plane of the - disc shaft is off-set, and parallel to that of the seat. This correlatu to one closed position, and two possible open positions. Of which, only one is i correct, the other orienting the valve disc 180* free the seat on closed signal. August 6, 1994 14. IDd-A-11A emetad esnt lanham. FFW-A-140 failm L W.0. 94-3019 was generated identifying excessive sea + leakage for (FW-A-340, !&C/Ma!ntenance was tne responsible department. 15. Seat cositian chad of IN-A-1do indientad ermeset easitten. W.O. 94-3019 esta611shed that the valve disc / seat were not mispositior.ed 180* out, as found on UW-A-338. Work Order 94-3025 was generated to disassemble the valve. 16. seat deterweration identified for fEW-A-340. Disassembly of valve rstealed seat wear. Valve rebuilt with spare parts. 17. 8 Roc. 5-55-10 TPC Recuired. The disassembly /reasserrely instructions were for 3 pinned ccMection for disc to shaft, disc was found melded to shaft. 18. Yrive erf ente,tfen te oinina incorrect. Yalve reinstalled by Mechanical Maintenance with valve shaft over adjacent piping interference. 11C unable to install actuator. 19. Yalve dise/maat, to actuator missosittened. During post maintenance testing it was discovered that the valve was mispositioned. August 7, 1994 20. UOR 94 040 ear.arated identifyir.e tant leakase of EFW-A-338 and EN a-340. This UOR identified that the preliminary peoblem was discovered en 7/16/94. At that t!ae, 0P3 was unable to idantify source of leakage. The UCR comef ned events for EPW-A-334 and EFW-A-340. EXHIBIT h PAGE [ OFM PAGEfS) )

PED-RC-94-014 Page _1. of i_ / 21. 19JLt.Caus. Analysis ercanded to include UW-A-340._ C~ At PEG mrning meeting, Section Head assigned L. Barron root cause for EfW-A-340. 22. EFW-a-340 normanently scribe marked. During interview with L. Leon and 8. Greenleaf, Maintenance riechanics, scribe mark on outer end of disc shaft identified. This sark was installed by Maintanance while valve was disassembled. There w.is also en existing mark on the coupling end of the disc shaft. This mark is not visible once actuator made up to valve (See attached). Direction of flow had also been stamped on the valve body. 23. UW-A-340 correct valve disc / seat to actuater oosition established. Witnessed stroke to: ting of the valve, and prelisimary saat leakage testing. Seth satisfactory. VALIDATION OF IDENTIFIED; The root cause for saat leakage of EFW.A-338 initially identified on work order 92-3007 was valve seat / disc to actuator mispositioning. Correcting the mispositioning on W.0. 92-3008 resulted in satisfactory' leak testing. Mis-positioning had occurred during reassembly of actuater to valve Wder. Y.0. 92-IT46(completed 4 3-92). The initial root cause for seat leakage of EFV-A-340 identified on W.O. 94-3019 (. was wearing away o/ seat. However, the root cause identified for EFW-A-338 was repeated during reassembly of EFW-A-340. Subsequently, EFW-A-340 correct actuator to valve orientation was established and the valve demonstrated satisfactory seat leakage. P9TENTIAL IMPLICATION 3; There are no cther air-operated butterfly valves with similar seat / disc to actuatcr orientation. EFW-A-334, 339, 340 are of vnique design. ConCLis5!00t3: The maintent.nce procedure (5-410) does not provide sufficient details regarding seat / disc and actuator criantation. This conclusion is based on above findings 1, 3, 4, 6, 7, 8, 12. 13, 19, 11 and 23. RECORREWJ CIMIRECTIVE ACTION 8: 1. Revise Procedure 5-55-10. " Maintenance of EFW Air Operated Trip Valves' (EFW-A-338, 139, 340) te include guidanch for establishing cornet seat / disc to actuator orientation. A sarted up copy is attached. 2. EFW-A-340 has been permanently urked to identify valve position. EFV-A-338 and 119 should be identically. marked (This includes direction of flow). 3. Saat leakage criteria should be established for EFW-A-338, 339 and 340, 4. Post Maintenance functional testing should include verification of seat tightness. EXHIBIT C PAGE 4 0FdD PAGE(S) i

..PGD R.-H - 0 M EN-- A -3 3 8 4'"'^C"*"*'^~ { VALVE POSITION w AcTuhT.c E.t.H.cR.N...N...._ .t.' J i o. 4._T. o... K9.A.. I.R. ~ 'T. PostTio.t. S WTHESSED..S.f. R. 3 i < n _,,.._._'. 6M @ stud ' _.... ~.. _.. I A,,. CPERA,t, V AL.QS Post.TncM on s otPT. SHUT i VAulE BODY ---** s eurap RETAmet l OtsInveT TEMPesAeV ttWcu./PuW j ( MA ret N G s G. 09.ES*.nNs oe.Pr ^~ opts /VAW4.8.Pos;Tip4 -- q } .i... 1 ..?.......... ...)..... l .s s... ....'..... t e t t I i 3 ...........e..+........t. i......... i - PC MTLo*J WTd SS SC. 89 P. M..tst.gog Q.AP..T.g4.RtP.N. t. (... o eva. V A t.v. t. A.o.. s.m. o J.......

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) Fros. Mo. 5-!b 10 z Rev. No. 3 C' Page 5 cf Il 4 l '( 4.8- .REAISEMBLY. INITIAL l n EII Refer to AttacPeent A. VAX.YE A33tM8LY, for referensu t numoers used in reuseaoly section, 4.6.1 - Carefully press bushings (6) into shaft bort of valve body (10) until the bushing is flush with shoulder insico the bore. 4.6.2 Insta11' thrust washers (7) in the body counter bores. = 4.6.3 Install ~ the valve disc (9) and align the stes bore with the body bore. (The disc must be rotated 180' for the closed position and installed from the inlet side of the body.) 4.6.4 Install the shaft (1) with the keywu on *he operator end of the body. 4.8.5 Align the disc ($) and naft (1) and press pins (8) in place. State the pins. C. 4.6.8 Install the 0-ring (11) amund the seat (12). l 6.7 Wib the vaive in the closed position and laying on a flat surface, place the sett assembly over the disc with the G-ring contacting the valve body. (The seat is smaller in diamatar than the dist.) body to avoid pinching the seat /0-ring.) into the valv carefully install the seat retainer (13 4.5,4 4.6.9 Install. saat retainer socket hosd car ::rews. P.93 o There is not 4 torque value for the seat retainer screws. o The seat retainer should be flush to 0.010 aheve the valve body. 4.4.10' Ti hten seat retaiw scrsus per Attachment 5, TOAQUE 31 EMCE PATTERN 3. 4.8.11' Install gland rings (5). t ~. ~~_. PAGE,7 OFJO PAGE(S)

PC.RC 94 C44 A'TT Ac.vadi "C. EFW-A 338 AcrvA. tor RoTATiot4 % VAwE POSmod 1 D <. d 5 SG#t ewe E&s / Posm o.J 4 f, * [ s 9hh [ M a u A toJ O t#0ftb0 TO s g m gp 0 M. _.e.@....'..D' dg Mm N J3 AA I' l Postmo wns toTATED w waceJ6 4 1N'$*OCd b ' w :IO48'TL % c1rmec sa9 CMPJ cn we

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W RC-H - c t 4 AttAC A4et/f "6. Prosecure

Title:

Proc. No. 5-55-10 Class. A j MA:NTENANCE OF EFW AIA UPERATED TRIP YALVES Rev. No. -3 l (EFa*-A-338, 339. AND 340) Issue Date 07-s0-94 ' Review Date 07/96 +fCR USC VITHCUT ADDITIONAL PED /QPD RIVIEW* Page 1 of 11 APPROYED SY PLANT MANAGER /DESIGNE! AT PORC Meli!NG # 94 041 CN 7/14/94 d Cate VO f: Valve No.: Y. v - e d.4 4 rj ';h sI 4 f. ;.: 17 b.f f. h 1.0 08HcTfn The objective of this proced:;re is to provide the necessary information 'cr personnel to perform saintenance on the Esdrgency Feed Water air operatec trip valves EFW-A-336, EfW-A-339, and EFW-A-340 using only a Work Order and this procedure. (CR5-1) 2.0 M CAU?!0NS Z.1 Contro.tatics Instruction Book, (C-9-1A) drawings and/or technical instructions can be reviewed before or during work on the component for assistance or addittenal inforr.ation. 2.2 Cao or c:ver all system openings to prevent foreign saterials (1 ( frore entering the system when the valve is resoved and when the '( [ systeen is left unattended. Z.J Some steps tre intended to be conditional or optional. For these steps, written justification is not recuired to 'N/A' the stap. These steps are delineated by N/A in parenthesis on the right side of the page. I.4 It a piece of safety class equignent having been removed from its indiate area of operation for repair is not reinstalled prior ts the and of the shift. K, 7ATERIAL IDENTIFICATION AND it shall be identified in accarta.ca with Procedure 044 e C3NTRCL, Dy tagging. sarking containers or other st.itable sethods and stored in a designated area until reinstalled. 2.5 If the required work exceeds the scope of this procedura, TI's and FTI's shall be written or revised as necessary and riviewed by PED and QPD in accorcance with Proceeure 0-15=3. WORK ORDER PROCESSING PROCEDURI. '/ 3.0 MIRECU11 ITIS INITIAL 3.1 The Wu number and valve number have toen documented on the cont of this piecedure. 't.L Tite v Avdt. Ac.Tuanog 'to dis 4f ta4T opefMTicw 4 i+ADeerMATM WS9ed No#ied.yAt.wt coce.%cas. TMil - (imetT AcTvwe. 5 A "mismo -ow.c Actuenot.,aue ensuites M22no *A ac.wa6 To' nau.au cuado Posmoo i* E EXHIBIT PAGE /0 0F2[LPAGE(S) p 9 Y

a Proc. No. 5-55-10 Rev. No. 3 Page 2 of 11 C INITIAL 3.2 The PS3 or 105 shall review the F1)NCTIONAL TEST!M forts attached. The PSS/ SOS should consider preparations for the post meintenance testing spesified. Preparatfon should include filling ot.t the Pre-Test Data section of the Valve Test Sheet I from Procedure 3.17.8.2. IST VALVE TEST FCR WORX ORDERS, if ( ann 1k M. ( P33/303 Signature: r 3.3 C.orrect parts necessary fer the repair of the valve have been. i determined Lad are available in stock. 4.0 PeCCEDURE ME Yalve'part locations are shown on Attachsmnt A, VALVE A13DtBLY. 4.1 PREPARATION (. 4.1.2 Contact I1C and request they remove and/or disconnect the air operator. [ Ransoved I&C 4.1.2 Mark valve and piping flanges to enturs that the valve will be ' reinstalled in the same orientation. 4.2 REMOVAL 4,2.1 Loosen the packing gland and check for sigrts of pressure g in the line. 4.2.2 Loosen and imeve flange bolts. 4.2.3 Carafully remove the valve so as not to damage the flange faces. 4.2.4 Remove gaskets and nota siis for future replacasant. 4.2.5 Install protective covers for flange protection and/or foreign atter141 exclusion. t.a.; gasket material, tape etc. s. M EXHIBIT 6 PAGE // OFM PAGE(S)

Proc. No. 5-55-l* Rev. No. 3 page 3 of 11 4.3 O!sA33EMSLY INITIAL 4.3.2 Remove packing nuts, gland retainer and packing gland studs as.iecessary. 4.J.2 Remove seat retainer, asat, and e-ring, as [- Care should be takan to assure the disc sealing edge is [ not niaked er scratchd during extesction of pin. 4.3.3 - Ar.sve pin frove shaft and disc assembly by rotating the dise TO' to the body and pressing the pin out. 4.3.4 Recove va*ve shaft, disc, thrust washers and shgft bushing. 4.3.5 Renovi packing. 4.4 INSPECTION 4.4.1 C1 san all parts using spatcheck Cleanar sxC-5 (MSDS pools) or an approved cleaner. 4.4.1 Inspect all studs / belts and threaded surfaces for: o Nicks o Visual cracks o General thread integrity. 4.4.3 Thercughly inspect the disc and body sent ring areas for: o scratches o Micks o Burrs o Conditions )Aich could affect its ability to function. 4.4.4 Inspect the valve mounting flange faces for: o scratches o Nicks o Surrs. 4.4.5 Inspect the stuffing bcx area for: o scratches o Micks ^ o Burrs o Conditions which could afftet its ability to seal. / C 5 EXHIBIT PAGE /1 OF)C PAGE(S) T

-) Proc.-No. 5-55-10 l Rev. No. 2 Page 4 of 11 ' [ 4.4.'6 Inspect the Shtft fCr: o Scratches l 1, e Nicks o Sending 9 Conditions dich could affect its ability to function. [ 4.4.7

==qc Hotc** { Inspect all parts to be reused during valve reassembly. [. COMMENTS: [ QC INSP. INITIALS /OtP.No./DATE f 4.4.8 Record inspection results noting any repaM perfemed. i This infarnation should be included in the work performed section of t!.i work order. i (. Ins serien Results/Renairs tenuired 4.4.9 If any deficiencies are noted in the valve that may have-

ireviously rendared the valva ineparable, notify Mechanical Supervisor.

(N/A) l I 4.5 REPAIR ,4.5.1 Replace all parts dich have catensive wear er desage. l 4.5.2 Repair siner defects by stontag, hand polishing or 1 130 ping. 4.8.2 If weld repair er aschining is required.TI's and FT!'s . shall be written er revised as necessory end reviewed by PED and QPO in accordance with Procedure 0-16-3, if0AK ORDER PRCCESSING PROCEDURE.. 6 d EXHIBIT PAGE /3 - 0FM PAGE(S) l :.-

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i ? ::. Mr., 5-35-10 Rev. No. 3 Paga 5 cf 11 ( 4.6 RIASSEMSLY IRITIA!. EII l Refer to Attaehment A, VALVI ASSENSLY, for reference numbers used in reassemoly section. 4.6.1 Carefully press bushings (6) inta shaft bore of valve body (IC) until the bushug 13 flush with shoulder inside the bere. 4.6 2 Install thrust vashers (7) in the body sounter bores, i 4.6.3 Install the valve dise (f) and align the stem bcre with the body bere. (The disc aust be rotated 180' for the closed position and installed frsa the inist side of the 5cdy.) 4.6.4 Install the shaft (1) with the *:eyway on the operator and of the body.. 4.6.5' Align the dise (9) ar.d shaft (1) and press pins (4) in place. Stake the pins. 4.4.6 Install the 0-ring (11) around the seat (12). ' 4.8.7 With the valve in tr.e closed position and laying on a flat l. surface. place the seat assembly over the disc with the 0 ring coetacting the valve body. (The seat is smaller in dienster than the aisc.) 4.5.5 Carefully install t*e seat retainer (13) into the valve body to avoid pinching the seat /0-ring. 4.6.9 Install seat retainer sacket head can screws. M o There is not a torque value for the seat rotatner screws. e The sete retainer should be flush to 0.010 above the valve body. 4.6.10 Tighten seat retainer screws per Attachment 8, TOROJC SEQUtNCE PATTDuts. 4.6.11 Install gland rings (5). EXHIBIT 5 PAGE // OFd4 PAGE(S)

proc. No. I-55-10 Rev. No. 3 Page 6 of 11 INITIAL 4.6.12 Install chevren packing (4) 3/C 29916. o Male packing piaca first, flat side down. o Install remaining packing with the ' cud" facing the inside of the valve. o Install the fosale pactiny piece last. 2F.19H evertightening the gla,'.d retainer nut could result in bending of the air operster. 4.6.13 Install gland estainer studs (!$), gland ring (3), and gland retainer (2). Secure lightly with hax nuts (;4), 4.6.14 Repeat Steps 4.6.11, 4.6.22, and 4.16.13 for the opposite end of the valve. 4.7 INSTA1.LATION 4.7.1 Inspect systes piping for cleanliness prior to installing the valve. @!. E 1m TM Yht.st we vasaser desww1mp tiesettw # ASO i 4.7.2 Ap ly a thin film of nuclear grade anti-satze to the be ting and to the bottom of the nuts to be torquad. '[ 4.7.) Install bottos four studs and gaskets. 4.7.4 Install rwtaining studs and nuts. 4.7.5 obtain a torque wrench with the correct size socket (s) and that has a torcus range in the middle half (15-75% scale) [ that corresponds to the required torque value u; 30 t ft/lbs. Torque Wrench #: Cal Due Date: I / j h EXHIBIT l PAGE M OF.)O PAGE(S) v 4

Noc. No. 5-55-30 Rev. 10. 3-Pt.ge 7 of 11 c. 2ND

INITIA1, VERIF.-

HEl It f: extremely important to feller the proper toreuing squence. If this squence is not fellowed, ehe flange can Decome cacked. Then regardfess of the amount of subsequent torquing, it cannot se brought bacr to parallel. Refer to Attachent I. TORQUE SEQUDICE PATTERNS, for proper torque sequence. I 4.7.6 Torque flange bolts to 200 ft.-lbs. using the three pass process. Fftst Pass 9 70 ft/lbs - Second Pass # 140 ft/lbs .l Final Torque t 200'Ft/lbs [- 70AQUE SEQUEN;E SAT / UNSAT [ COMMENTS: JC INSP. INITIAL 5/EMP.No./ SATE: 201RTti N #pd DME/8detT44' CAreMCC-4.7p. [ Contact I&C and request they insts 1 and set uq,uJtadf.Nm Qtwg g A Q p the 41r operator. 14C 4.7.8 Valve identification tag.is in place. If notify Operations to install an rsisstagIcation tag. L identif 5.0 FINAL ccN01TIONS ELTI 2 during the seurse of the work, this procedure proved nadequate out of sequente, or otherwise difficult to use, I)g unrt Ug a steen copy and put it In the Maintenance ' red file for consideratton during the next review of this precedure. 5.1 SUPERVISORY REVIEW 5.1.1 Werk order has been completed. detailing work i accouplished and is resty to' send to Operations for functional t,esting in accordance with Attachment C, FlANCT10NAl, TUTIME. Satisfactory Completion Maintenance superviser Data - [IM$t2T {""'ISu49eeggY.l.S.4,8^. f48#64 49AmiotS.t.1e 4C.19478f ed.:....... ....., '% gNf867 #M M 'ar'** - a n. \\. d EXH1817 i DAGE /d OFdPAGE(S) Y

Prec. No. 5-55-10 Rev No. 3 Page 5 of 11 6.0 RETERfM f5 6.1 SOURCES 6.J.] Centromatics Instnction Manual, ' Operating & Maintenance Instruction v:s Ptrts Catalog for 3 Inch Class 800 Butterfly Valves and Operating povtcas", Doctanent Muscer EX-QA87965 (IB t-9 ]A). 6.2 CITATIONS d.2.1 Procedun 3-15-3, WORK ORDER PROCES$!NG PROCEDURE. 6.2.2 Precedurs 3.17.3.2, IST VALYC TEST 3 FOR WORK ORDERS. l 6.2.3 Precadura 0-08-I. MATERIAL IDENTIFICATION AND CONTROL, 6.3 CCMMITMENTS ( 8.3.1 CAS-1, CMS 28-3-4. %velopment of New Maintenance Procedures *. ( d ( EXHIBIT L PAGEff 0FMPAGE(S) i I

e Proc. No. 5 13-10 Rev. No. 3 C hge 9 of 11 ATTAC2f.RT_4 VALVE A!!DtBLY I %,+-- .g e l t --E) --ii.. a.- g m 4 f M e.N n 3 s. [ g (.) /Cl t_. J --3p d l g /* 9 r'f 3 (j] N;, kf N f l 3 e- \\ 4 T d-[., M x \\ l 1_I [ 'W m t f EXHIBIT I PAGE[0Fl.P._.FAGE(S, E _ ____ __.:_ _ --____ _ _ ________ _ _.,__ _____-.___

Proc. No. 5 5510 C.... e. Rev. No. 3 Page 10 of 31 l A mat 99ENT 3 TCR"'JE IEGUENCI 74TTIMS Thart art two baste ter:::: sa:::ence patterns; cr.: Nr circular arrangresents sad ens f:r eval or alongatzd arra.twents. For t:rque pattarns net addressed in this peccadurs, contact Plant bgineering. l' l2 8 'l I E E 5 .5 4 g 3 8. I 4 a 4 l (' ~~ l 3 6 10 H' I 7 4 i '3 1: il 3 7 g 16 UI I i 23 ! ! I in s 5 .S.C ,7 17 g *19 t-e y a m is 8 M 2! E 3 to it. 4, 7 7, 4 o o a g' U' 43 3 8 14 7 22 ft 2. . t 0 I g g it 14 l 4 d E llf IC E l '1 I4 4 I e 3 4 7 11 / 8 o a

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g a Proc. No. 5 55-10 Rev. No. 3 C Page 11 of 11'

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FUNCTIONAL TEs ING WC f: Neinnet4hIa Seeartaant 1. Geerstfons A. Check the vaive and ass',ciated tubing connections for Zero external 1eaksge. + . NEE The Yalve should be leak ti@t at full systes pre 5 Nm and temperature. However, ' n some cases (such as returning a systas to operation after a refueling or extanded outage) full system pressure and taperature may only be gradually achieved ever a period of time. ~( s' satisfactory: 095 Q S. Perform a pre-opers'tional check from the Auxiliary Shutdown Panel to insure that the valve cycles preserly. !atisfactory: C. Perform testing as required by Procedure 3.17.2.2, IST VALVE TE5TS FOR WCRK ORDEA3. If there is any question which test (s) should be perforsed, then call PED for resolution. Satisfactory: 095 D. Yalve verified acceptable by: PS3/305 Date Timm s { EXHIBIT ~ PAGEJO OF)O PAGE(S)

, Cs0 -lf a n a MaineYankee

2) /. 4' 7. I ACLia 6L E Etf C7a* cit V StNCE + 972 329 EATH ROAo
  • 8 AUNSWICK MAINE odo11 * (207) W100 i

RESPONSI5!LT/ \\/cilIcm RESPOND BY N4 NRC DUE DATE M4 September 1, 1994 MN-94-95 JRH-94-215 UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555

Reference:

(a)- License No OPR-36 (Docket No. 50-309)

Subject:

Maine Yankee Licensee Event Report 94-016, Emergency Feedwater Isolation Valve Leakage (- Please find enclosed Maine Yankee Licensee Event Report 94-016. This report is submitted in accordance with 10CFR50.73(a)(2)(ii). Please contact us should you have any questions regarding this matter. Very truly yours, ames R. Hebert, Manager Licensing & Engineering Support Department JRH I Enclosure l l c: Mr. Thomas T. Martin l Mr. J. T. Yerokun Mr. E. H. Trot'ior Mr. Patrick J. Dostie i RCtfriNG ADDt I l CASENO. 1-96-040

8 ETs :"M-=

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q/ l (. EXHIBIT _7 j) kW f = PAGE[0F1PAGE(S) A-23-19

e Imc FORM 366 U.S. NUCLEM REGut.ATORT CDMN15310N APPROVED 87 000 No. 3150-0104 ($ 92; EXP!RE5 5/31/95 C- -- EST!NATED BUR 0Ei PER RE!PON!E TO COMPLY VITH : -l TH13 INFORMAT! N COLLECTICM REQUEST: 50.0 MRS. LICENSEE EVENT REPORT (LER) rCRwARO CoM*ENTs REGARotNG - suRetN EsTIMATt re THE INFORMATION AND RECOR S MANAGEMENT BRANCH I (MN66 7714). U.S. NUCLEAR REGULATORT COMMI55!CN. i (See reverse for reewtrec numee* of digits / characters for each block) WASHINGTON. DC 2 555-0001. AND TO THE PAPERWCRK REDUCTIDM PRC.ECT (3150-0104). OFFICI 0F uaNAG M NT AND M GET waswtNATON DC 20503 FACILITY NAfE (1) 00(XET 3RMBER (2) PAGE (3) Maine Yankee Atomic Power comoany 50-309 10F 4 TITLE (4) Emeroency Feedwater isolation Valve leakaoe EVFNT DATE f5) tER 4tmBER (5) RE*0RT DATE (71 DTNER Facitf 7IES fMVOLVED f 91 $E0VENTIO. REVI$10N FACILITY NAME DOCKET NUMBER MONTM DAT TEAR YEAR g NUM8ER NUMBER NA 00 09 02 94 b'#3'3II"*"I (A 08 04 94 94 016 3 Tut 3 RE*0RT f f stmMITTED PtastlANT TO THE REQUIREMENTS OF 10 CFR E ' fChen we e* acent fill (PERATING le0E (9) 20.402ft) 20.40$fc) 50.73(a )(2)(< c) 73.11(b) pgygg 20.405faH11(t) 50.36(c)(1) 50.73(a H2Hv) 73.11(c) LEVEL (10) 20.405(alf1Htt) 50.36fc H2) 50.73(a H2)(vit)

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g. y w w w.- n 20.405faH1Htit) 50.73(a H2)(1) 50.73(a H2 Hvitt)(A)

(50ectfy in gjY $ U 20.405f a H1 H t v) X 50.73(a)(2)(st) 50.73(a H2)(vitt HB) Abstract below p.,.g,M4 20.405fa HI)fv) 50.73faH2)(tt1) 50.73(alf2)(s)

  • "! " _*b.

e LICENSEE CDNTACT FOR THIS LER fif) NAME TELEPHONE NUMBER (Incluce Area Coce) George Stowers, Nuclear Safety Specialist (207) 882-6321 M 7 ONE LINE FOR EACH CDIPONENT FAf t tRE DESCRIBFD IN TFIS REPORT tit) I 'CAUSE SY3 TEM COMPONENT MANUFACTURER h CAUSE $737EM COMPONENT MANUFACTURER ORiA I qpg w ,D 7.p ig sr I SUP*LMNT AL RE*0RT EXPECTED fla) EXPECTED 404 N DAY vf30 TES SUBMISSION (If yes. c =eier. EXPECTED sueMissicN c4TE). X No NA NA NA DATE (15) AssTRAct (Limt to 14c0 so ces. i.... aceronimately 15 singi.. oaceo typ ritten itnes) (1s> At 1220, on August 4, 1994 with the reactor in a cold shutdown condition, plant operators determined that an Emergeray Feedwater isolation valve for #1 Steam Generator was leaking by.- Further investigation identified similar leakage in the Emergency Feedwater supplies to #2 1 #3 Steam Generators. Subsequently it was detemined that under af.cident conditions which require isolation of Emergency Feedwater, isolation valve leakage could exceed Safety Analysis assumptions. Maintenance activities have been performed to reduce Emergency Feedwater valve leakage. In addition, administrative controls have been implemented to ensure Emergency Feedwater leakage _is maintained within the bounds of Safety Analysis assumptions during accident conditions. 7 EXHIBIT CASE ND. }_nng4O l PAGE_1 0F y PAGE(El%) e. ..m.. -,-o 4, n*-i%,, - - =,, - -, - ,w-w - - -i sw.i. - - - -. + - - - - - -- -- < =----

e LC F0b 366A u.5, NUCLEAR RE6utATORT CCml15!CN ' APPROVED BY WB No. 3150-0104 t n (5 92) EXP!RE5 $/31/95 j EITIMATED BUROEM PER RESPONSI TO COMPLY W!rm C, Th!$ INFORMATION 00LLECTION REQUEST: 50.0 HRS. ~ICENSEE EVENT REPORT (LER) U,"r0 E $ 3 M EtO$5 ENEE N " sRANN t TEXT CONTINUATION (Mass n!4), u.5. NuctEAR :'EGULATORT COMMI5510N. VASHINGTON, DC 20555-0001. AND TO THE PAPERVCRK REDUCTICN PROJE:7 (3150-0104). OFFICE CF manaGEutwT AND av0GET vaswfNGTON OC 20503 FACILITY NAME (1) 1 00CrE7 pe9 f f) L D U BER f6' P4ff f31 SEQUENTIAL REV!$10N Maine Yankee Atomic Power Company 2 0F 4 50-309 94 016 -- 00 ' TUT f If Oe sence 's eeu'*ee. vse seest+eani eeeies of woc r a 3f6Al (17) e On August 4, 1994 Maine Yankee was in a cold shutdown condition making preparations to restart the plant fcilowing a maintenance outage to correct Steam Generator tube leakage. At approximately 1220 while performing a leak test of the Emergency Feedwater (BA) Isolation Valve (ISV) for #1 Steam Generator, it was determined that EFW-A-338 leaked by at a rate of approximately 400 GPM. Leakage past #1 Steam Generator Emergency Feedwater Regulating Valve (FCV) EFW-A-101 at 75 GPM was also identified. These leakrates were determined using normal system instrumentation. Subsequent investigation determined that the actuator for EFW-A-338 was coupled to the disk approximately 180 degrees out of alignment. This misalignment is believed to have occurred during valve maintenance performed during our 1992 Refueling Outage. Maintenance activity was initiated to reposition the actuator / disk to the correct orientation. At 1700 on August 5, 1994 it was determined that Maine Yankee's Steam Line Break Safety ( Analysis assumes zero leakage past Emergency Feedwater isolation and regulating valvos. 'herefore, at 1957 on August 5, the NRC was appraised of the situation via the Emergency w.otification System in accordance with the provisions of 10 CFR 50.72 (b)(2)i. A follow up notification was made at 1151 on August 6, 1994 after leakage was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW-A-340. On August 10, 1994 comprehensive testing was performed on the Emergency Feedwater isolation and regulating valves for all three Steam Generators using precision test equipment. These tests revealed leakrates for various individual valves and combinations of valves from 0 to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studies conducted in parallel with these tests established the following revised acceptance crite-ia: 1) Leakage past individual isolation / regulating valves should be less than 40 GPH. The purpose of this limit is to prevent excessive cooldown following a Main Steam Line Rupture with failure of either the isolation valve or regulating valve to close. 2) Leakage past both the isolation and regulating valve to each Steam Generator should be less than 10 GPM. The purpose of this limit is to provide at least 30 minutes for operators to ensure EFW is isolated to a faulted Steam Generator following a Steam Generator Tube Rupture. Such action is taken to ensure SG safety valves will not be lifted due to overfill. 3) Zero leakage to a Steam Generator with a tube rupture within 30 minutes. The purpose of this limit is to prevent lifting SG safety valves due to overfill following a Steam Generator Tube Rupture. EXHIBIT ~7 ordertomeettherevisedacceptancecriteria,EFW-A-101 wha 8 bcehakage. ' past EFW-A-101 & 338 to zero GPM and administrative controls were implemented to ensure termination of any potential future leakage to a ruptured Steam Generator within 30 cinute g igg m nual i oja i n valves (V).

ry (( ], UMTto STATES I ,.,., ? p NUCLFAR REGULATORY COMMISSION e es RtoioN - 475 ALLENDALE ROAD LI

o

- KING OF PRUSSIA. PENNSYLVANIA 1H06141$ - .y... f ( - September.- 22 1994

Mr. Charles D. Frizzle-
President;
Maine. Yankee Atomic Power Company-329 Bath Roa6 Brunswick,' Maine 040112

DearMr.IFrizzle:

SUBJECT:

NRC INSPECTION NO. 50-309/94-15 This refert 'a the resident-inspection activities conducted by Mr. J.-Yerokun Land others of this office from July 17 through September 6,1994. The inspection includ.d observation and review of activities authorized for the Maine. Yankee Atomic: Power Plant, Wiscasset, Maine.; The findings were -. discussed with members of your staff during the inspection period and with Messrs. G..Leitch,- D. Whittier, R. Blackmore and others of your staff at the exit meeting held on September 8,1994. t l The inspection consisted of document reviews, interviews, and observation of activities 1important to public health.and safety. The purpose of the inspection was to determine whether activities authorized by the license were cc,nducted safely _ and in accordance with NRC requirements. Areas examined 'during the inspection are identified in the report. Based on'the results of this inspection, un apparent violation was identified and is being considered for escalated enforcement action in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions", 10 CFR'Part 2, Appendix C (1994). The apparent violation is the failure of l Maine Yankee's test program to assure that all testing required to demonstrate-components will perform satisfactorily in service is identified and performed -as'reguired by 10 CFR Part 50, Criterion XI Test control. This failure of your engineering organization to determine leakage criteria for the emergency feedwater isolation valves-and translate it into appropriate testing requirements 1 contributed to a situation in which these valves would have been unable to isolete feedwater to a faulted steam generator. Accordingly, no Notice of Violation is presently being issued for this inspection finding. In c l-addition, please be advised that the numbo and characterization of the apparent violation described in the enclosed inspection report may change as a -resulttof further NRC review. An enforcement conference to1 discuss this apparent violation has been. l' scheduled for October 1_4, 1994. The purpose of this conference is to discuss the apparent violation, Lits causes and safety significance; to provide you the opportunity to point ouc any errers in our inspection report; and to= provide .a'a opportunity for you to present your proposed corrective actions. In l particular, we expect you to address the_ issues concerning the excessive valve h leakage. These concerns.are (1) the ' apparent f ailure of several barriers fincluding maintenance work control, quality controls, and post maintenance l 4

testing;:and, (2)- the corrective actions from similar events which may have CA8 N 1 96'-040 EXHIBIT PAGE 7 OF_)J__PAGE(S)

M4{di '{. g .=.

l c Mr. Charles D. Frizzle 2 prevented this problem, including the identification of an emergency feedwater isolation valve installed backwards in 1992, and the excessive leakage identifled in a secondary component cooling and a primary component cooling isolation valve. In addition, this is an opportunity for you to provide any information concerning your perspective on 1) the severity of the issue, 2) the factors that the NRC considert when it determines the amount of a civil penalty that may be assessed in accordance with Section VI.B.2 of the Enforcement Policy, and 3) the possible basis for exercising discretion ili accordance with Section VIII of the Enforcement Policy. You will be advised by separate correspondence of the results of our deliberations on this matter. No response regarding the apparent violation is required at this time. Tnis enforcement conference will be open to public observation in accordance with the Comission's continuing trial program as discussed in the enclosed Federal Register notices (Enclosure 2). Although not required, we encourage you to provide your comments on how you believe holding this conference open to public observation affected your presentation and your communication with ,the NRC. We also noted two. instances, in which your staff failed to adhere to NRC requirements. In one case, your staff failed to return a high pressure safety injection (HPSI) pump control switch to the " pull-to-lock" position as required by your surveillance procedure. However, this error was quickly identified and corrected in approximately 90 minutes, well within the ^ technical specification remedial action time of six hours to return all but one HHSI pump switch to the " pull-to-lock" position. In the other case, your (' staff identified that one of fourteen smoke detectors in zone 46 of containment was not included in the surveillance procedure. Technical Specification 4.12. A required testing of containment smoke detectors. In accordance with 10 CFR Part 2, Appendix C, these violations are not being cited because the safety significance was low, your staff identified and promptly corrected the discrepancies, and the violatiors were neither repetitious nor preventable by corrective actions to previous violations. ~ In acErdance with 10 CFR 2.790 of the NRC's " Rules of Practice", a copy of this letter and its enclosures will be placed in the NRC Public Document Room. Sincerely, t Richard W. Cooper, I.. Director Division of Reactor Projects Docket No. 50-309 EA No. 94-183

Enclosures:

1. NRC Region I Inspection Report 50-309/94-15 2. Federal Reaister Notices ~ CASE NO. 1-96I04h EXHBIT PAGE } OF_Al_PAGE(S)

t Mr. Charles D. Frizzle 3' cc w/encis: C: G. Leitch, Vice President, Operations P. Anderson, Project Manager (Yankee Atomic Electric Company) R. Blackmore, Plant Manager L. Diehl,-Manager of Public and Governmental Affairs J. Ritsher, Attorney (Ropes and Gray) P. Dostie, State Nuclear Safety Inspector P. Brann, Assistant Attorney General U Vanags, Maine State Planning Office C. Brinkman, Combustion _ Engineering, Inc. First Selectmen of Wiscasset Maine State Planning Officer Public Document Room (PDR) Local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) K. Abraham, PA0 (2 copies) NRC Resident Inspector . State of Maine, SLO Designee (- { CASE NO. 1-96-040 ;. EXHIBIT PAGE_ 3 0Fl3_.PAGE(S)

- Mr. Charles D. Frizzle-- 4 h- -' - Region ! Docket Room _(with concurrences) bec w/encis:.- D. Holody,--- EO - W. Lazarus, DRP - D. Lew, DRP - J.'Lieberman, OE-J._Linville, DRP- . bec w/ enc 1s (VIA E-MAIL): W. Dean, OEDO E. Trottier, LPM, NRR W.-Butler, NRR N. Shannon, ILPB. ..O e e 4 (CASENO. 1 - 9 6 - 04 di - r + Exsisir PAGE_,Y OF >J PAGE(S)

U.S. NUCLEAR REGULATORY COMMISSION REGION I Report Number: 50-309/94-15 Docket Number: 50-309. License Number: DRP-36 Licensee: Maine Yankee Atomic Power Company 329 Bath Road Brunswick, Maine 04011 Facility: Maine Yankee Atomic Power Station Inspection Dates: July 17 to September 6,1994 Inspectors: J. Yerokun, Senior Resident Inspector W. Olsen, Resident Inspector D. Lew, Project Engineer Approved By: JM': 9/MNjV W. V17zarus(pief Dite ' Readttir Projects Section 3B Scope: Resident inspection and safety assessment of plant activities -(.. including operations, maintenance, engineering, and overall plant support. Overview: See executive sumary. WlEWi. 1-96'-04 0 EXHlBIT l - PAGE_8__OF.)_2.PAGE(S) L. n ' a, o, cy { e!(p! I n ,v pjp lq g,, v ( L

9 HS' EXEClTTIVE SlfMMARY Operations Operators demonstrated excellent safety perspective in conducting plant evolutions during normal power operations and shutdowns and in identifying and correcting problems. However, some instances of inadequate attention to details were observed when a hose was improperly left attached to an emergency feedwater pressure gauge and when operators left a high pressure safety injection pump control switch in an incorrect position. Appropriate attention and monitoring continued to be placed on reactor coolant system leakage. Maintenante Maintenance and surveillanen activities were conducted safely and in accordance with procedures. Sone weaknesses were idc1tified with the details and sequence of work instruction steps and personnel not performing steps ir. the most appropriate order. These instances were indicative of inattention to detail and performing to less than management expectations.

  • Enoineerina Engineering organization maintained good safety perspective and provided good support to the plant. The steam generator tube repair outage was conducted in a safe manner supported by knowledgeable staff.

A failure of the engineering organization to determine app-opriate leakage criteria for emergency feedwater isolation valves contributed to a situation ( in which these valves would have been unable to isolate a faulted steam generator. This situation also indicated weaknesses in work control, post maintenance testing, quality controls, and corrective actions taken in 1992. The event is being considered for escalated enforcement. Plant Support Piant support activities in the areas of radiological controls, security, and emergency preparedness were conducted safely during this period. Good work practices were evident as well as conformance with requirements and procedures. Safety Assessment /Ouality Verification Management kept abreast with ongoing activities and demonstrated excellent safety perspective. Maine Yankee performed well in identifying and correcting deficiencies. The Quality Assurance Department identified an inadequate fire detection surveillance procedure, which caused a violation of technical specifications. Operators also identified a mispositioned high pressure safety injection pump control switch. il f Cut. 4 i, i. PAGE_ f OF_13_PAGE(S)

TABLE OF CONTENTS ii EXECUTIVE-$UMMARY............................ ( iii TABLE OF CONTENTS 1 i 1.0 OPERATIONS ............1 I 1.1 Steam Genwrator Tube Repair Outage............. 2 1,2 Reactor Coolant System (RCS) Leakage 1.3 High Pressure _ Safety Injection (HPSI) Pumps and RCS Low 2 Temperature Overpressure Protection (LT0P).......... 3. 1.4 Hurricane Readiness..................... 4 2.0 MAINTENANCE............................ 4 2.1 Maintenance Observation................... 5 2.2 Surveillance Observations.................. 2.3 Secondary Component Cooling (SCC) Valves SCC-A-460 and 461 6 7 3.0 ENGINEERING................. 7 3.1 Steam Generator (S/G) Tube Repairs .............= 3.2 Emergency Foodwater-(EFW) Valve Leakage outside Design 8 Bases............................. 4.0 ' PLANT SUPPORT........................... 10 4.1 Radiological Controls.................... 10 11 4.2 Security.......................... 11 -4.3 Emergency Preparedness ( 11 5.0 SAFETY ASSESSMENT / QUALITY VERIFICATION 12 6.0 ADMINISTRATIVE 6.1 Persons contacted...................... 12 6.2 Summary of NRC Activities.................. 12 6.3 Interface with the State of Maine.............. 12 12 6.4 Exit Meeting bi L-p- = CASE ND. 1 96-040' iii EXHIBIT - I L PAGE,7 0F A3 PAGE(S) 1o ! ~ . ~. .c

\\ ) o DETAILS -1.0 OPERATIONS At the beginning of the inspection period, the plant was; shutdown for steam generator tube inspection and repairs.- Power operations resumed on August 14, 1994.: On a daily basis, inspectors verified adequate staffing, appropriate -i access control, adherence to procedures and technical specifications limiting conditions-for operation, and operability of-protective systems, including emergency power sources. The inspectors also verified operability of selected ' Engineered Safety Features (ESF) trains and assessed the condition of the plant equipment, radiological controls, security ar.d safety. The inspectors monitored the status of control room annunciators and radiation monitors to ascertain that they were being maintained adequately. The inspectors evaluated plant ecuipment material condition to ascertain that it was well maintained and hat no detrimental effect on plaat safety. On August 14, 1994, during a plant walkdown, the inspector observed a clear plastic hose attached to EFW pump 25C suction pressure gage, PI 1203C. There was no pump testing or-other work ongoing end the attachment was not , designated as being temporary. This discrepancy was innediately brought to -the attention of the plant shift supervisor (PSS). The licensee investigation revealed that the hose was installed during the outage in accordance with procedures 1-104-14.1 (.2,.3), Placing Steam Generator No. 1 (No. 2, No. 3) in Wet Layup & Steam Generator No. 1 (No. 2, No. 3)-Pressure Tube Leak Test. However, the procedures did not appropriately equire removal of the hose. ( The hose was removed and the PSS ir.itiated a procedure change suggestion to _A rectify this inadequacy. 1.1 Steam Generator Tube Repair Outage From July 16 to August 12, 1994, the plant was shutdown to inspect the steam generator tubes and to plug defective tubes. This activity was performed because the primary to secondary leakage had increased enough to necessitate the shutdown for repairs. The results of the inspection and repair activities are-digcussed in section 3.1. After the defective tubes were plugged and other shutdown activities completed,.the plant was restarted. - Plant heatup was initially delayed due to high total undissolved solids (TUS) in the reactor coolant system (RCS). The initial concentration on August 9, 1994, was approximately 7,500 parts per 4 billion (ppb). The TUS appeared to be evenly distributed in the three RCS. loops. After several cycles through combinations of letdown purification, the - TUS was reduced to less than 500 ppb and plant her. tup was restaaed in accordance with operating procedure (0P) 1-1, Plant Heatup. The licensee documented the issue of undissolved solids in the RCS in unusual occurrence i report (UOR) No. 94-064. The source of the undissolved solids was attributed to the steam generator tube inspection and plugging activities that occurred during the shutdown. The inspector observed portions of the reactor startup per Procedure. 1, Reactor-Startup and afterwards reviewed the cnpleted procedure cad arified . ' {. _ tnat it was-completed satisfactorily. The inspector verified that portions of the prerequisites (section 5.0), such as the opening, locking out and white d.. EXJHIT . C#3E M PAGE % OF }3 PAGE(S) 2 c i

2 ( tagging (safety tag) of the breakers for valves RC-M-11, 12, 21, 22, 31, and 32 and SIA-M-11, 21, and 31 were completed prior to commencing the reactor startup. The plant was critical at 12:33 a.m. on Saturday, August 13, 1994. The power operation mode was resumed at 5:45 a.m. on August 14, and the generator was synchronized onto the grid at about 10% power at 10:07 a.m. At about 50% power, operators safely transferred main feedwater supp1'y to the steam generators f rom the electric motor driven feed pump, P-2A. to the steam driven feed pump, P-2C, The inspector concluded that the transfer of pumps was performed safely and in accordance with procedure 1-104-5, and that operators demonstrated good safety perspective. Primary to secondary leakage has remained below 0.251 gallons per h y since return to power operations. 1.2 Reactor Coolant System (RCS) Leakage Total RCS leakrate remained at about 0.38 gallons per minute (gpm) at the end 'of this period. RCS identified leakage is limited to 10 gpm in Maine Yankee technical specification section 3.14, Primary System Leakage. About 0.14 pgm of this leakage was attributed to RCS Loop 1 Het Leg Isolation Valve, RC-M-11, packing gland leak. The valve packing leak prior to the mid-July shutdown was about 0.04 gpm. This increase in leakage rate was evalurted by the licensee and determined to be minor, and that it did not indicate aru significant degradation in the valve's packing. The inspector reviewed the data maintained by the licenses to trend the valve leakage and also discussed the ( issue with the licensee. The licensee stated that the valve packing will be repaired (repacked) during the next efueling outage after RCS decontamination is performed. The licensee also stated that along with the fact that the valv2 packing had not deteriorated, ALARA consideration was a major factor in deferring the repacking until the next refueling outage. The inspector concluded that the current leakage was minor, adequately contitined (diverted by permanent piping te the pressurizer qJench tank) and that Maiintt Yankee continued to adequately monitor the leakage. 1.3 High Pressure Safety Injection (HPSI) Pumps and RCS Low Temperature Overpressure Protection (LTOP). During a shift turnover meeting on August 11, 1994, the oncoming control room c#erator (CRO) noticed that neither of the handswitches for the high pressure tifety injection (HPSI) pumps was in the " pull-to-lock" position. With the plant in technical specifications operation mode 4 (reactor coolant system temperature between 210 and 500 "F.), the reactor coolant system (RCS) low temperature overpressure protection (LTOP) necessitates that one of the HPSI pumps be in the " pull-to-lock" position. Technical Specification 3 A.D.3.C states that in that condition, only one HPSI pump shall be aligned to provide charging flow and be available for HPSI service. Further, it requires that the control switches for the other HPSI pump must be in " pull-to-lock". ( CASE ND. 1.-1)6- 04 0 b EXHIBIT PAGE_ /9 0F_))__ PAGE(S)

3 (D -. 'At_that-time, punp P-146 was in operation but the handswitch for.P-14A was in-. the.' Auto' position. The licensee determined that following the completion of procedure 3-1-2.4 Energency Core Cooling System Routine Testing, earlier, pump P-14A was misti.kenly left in the " Auto" position. .The int.pector disrdssed this issue with plant operators, reviewed the completed plant r,rocedure (3-1-2.4), reviewed plant conditions. and applicable-technical specifications. Procedure 3-1-2.4, Step-4.4 warned that if the RCS temperature is 'less than the temperatura criteria for running two HPSI pumps, then only one HPSI pump control switch may be out of the " pull-to-lock' position. Additionally, Step 6.6.3 of the procedure requires that if the RCS temperature is less than the temperature criteria for running two HPSI pumps, then the control. switch of the secured pump must be placed.in the " pull-to-

1ock" position. However, the operator performing the procedure' erroneously entered 'N/A' for the step.

The licensee documented this event in an unusual occurrence report (UOR No. 94-068).- Imediate r;orrective action was taken to return the switch back to " the " pull-to-lock" position. The switch was in the incorrect position for approximately 90 minutes. 'The licensee's review of the procedure indicated = that the procedure was adequate; however, the operator erred in making the step 'N/A". Technical Specification 3.4.D states that only one HPSI pump be aligned for 1 l operation, and-that the control switch e for all other HPSI pumps be in epu11-to-lock". An exception is provided for up to fives minutes to allow a second HPSI. pump to be out of " pull-to-lock' for the purpose of rotating operating equipment. The technical specification remedial actions require that if this l condition is not met, the licensee has six hours to restore the pumps to the required conditions, cased on the allowances provided by technical specifications and the relatively short period in which the second pump was out of " pull-to-locL", the inspector concluded that the event had low safety significance. InacIordancewiththeNRC'sEnforcementPolicyin10CFRPart2,AppendixC, Section VII.8(1), this violation of technical specifications is not being cited because, the licensee identified the violation and took prompt and appropriate corrective actions including placing the switch in the correct position. The safety significance was low since the switch was not in the correct position for a short duration of time. Additionally, the violation was neither repetitious nor preventable by corrective actions to a previous-violation. 1.4 Hurricane Readiness To ascertain that Maine Yankee was adequately prepared to deal with a hurricane, the inspector reviewed the licensee's hurricane readiness. -There

have been no advance-directives issued requiring a plant shutdown should a hurricane approach the site.- However, plant abnormal operating procedure I

(A0P) 2-40. High Wind, Hurricane or Tornado, contains contingencies for a plant shutdown. The procedure provides guidance for preparations in the event EXHIBIT I EWL. 1 l PAGE[OFM PAGE(S)

a 4 (- of a possible high wind, hurricane,.or tornad'o condition-and:for_ placing the: . plant in a-safe condition. ;The plant' shift supervisor- (PSS)_ has the option of initiating a plant shutdown. The inspector verifled that in addition to the normal-telephone lines there f -were other connunication capabilities such as separate phone lines, state police radios,' microwave phone. lines and walkie talkies that would be i .available in an emergency. The inspector discussed the use ol' these other connunication capabilities with the licensee to ascertain that the licensee staff is trained on the use of the equipment in an emergency. l u A review of the plant flooding history showed that no safety related or important-to-safety equipment are susceptible to room flooding during periods of intense rainfall. No such equipment would therefore become inoperable or inaccessible to operators due to direct exposure to severe weather. 2.0 MAINTENANCE

  • Overall, maintepance and surveillance activities continue to be performed well, except as discussed in section 3.2.

The inspectors ascertained that activities were performed safely and in accordance with approved plant procedures. Maintenance activities were conducted safely during the steam generator tube repair outage. L.~ 2.1 Maintenance observation The inspectors observed and reviewed-selected maintenance activities to assure i that the activities were conducted safely; complied with technical specifications and work order (WO) requirements;-that required approvals and releases were obtained prior to commencing work; that the work procedures were appropriately detailed and followed; and that equipment was properly tested and returned to service. The inspectors observed portions of the following .naintenance activities: e WO g4-03037, Diesel Generator (DG) IB Relays f e WO 94-02097, Inspection and Repair of General Electric AK-25 Reactor Trip Breakers, TCB-1 eL WO 94-02098, Inspection and Repair of General Electric AK-25 Reactor Trip Breakers, TCB-2 i e' WO 94-03313, Secondary Component Cooling (SCC) Heat Exchanger E-5A cleaning overall, the maintenance activities were performed satisfectorily. The e . troubleshooting _ effort to resolve anomalies associated with diesel generator 18, reflected excellent engineering effort.- personnel involved in the . troubleshooting activities-effectively determined the potential causes, and subsequently. -identified a failed relay coil. :The inspector noted, however, that the human factors considerations associated with the work instructions to test and replace,-as necessary, DG 18 relays were weak. Specifically, several steps and actions were not in the appropriate sequence. A step in the procedure directed installing the relay before directing the weighing of the C relay coil for seismic concerns. Additionally, the_ functional testing of the CA8N 1-96-040 EXHlBIT f PAGE // OF_1f_PAGE(S) 1 -... ~

4 5 { relay coil required that one wire from each contact be disconnected before performing continuity checks. However, this step was several pages after the wires had already been connected to the relays. As a result, maintenance persor.nel had to connect and verify wire installation twice. The inspector noted some minor deficiencies concerning Procedure No. 5-77-2, Revision 4, Inspection and Repair of General Electric AK-25 Reactor Trip Breakers. One step involved verifying that the breaker was one of two configurations; however, the step listed three subsequent items. The third item was in fact not a description of a configuration but should have been a subsequent step. Inconsistencies were also noted in the marking of steps not applicible (N/A). A note in the procedure indicated that the if a step directed going directly to another step, the steps which were skipped need not be marked N/A. Personnel were marking all the skipped steps N/A because of confusion in interpreting the directing step. However, all actions were performed properly. The inspector noted that maintenance personnel have alt,o marked several ambiguous wording and potential improvements for incorporation into the next revision. The inspector observed a weakness in procedural adherence during the cleaning of the SCC Heat Exchanger E-5A. Although the licensee allows deviating from the sequence in a maintenance procedure provided that supervisory approval is provided, the inspector nr.ted that the procedure could have been performed in the sequence without affecting the quality of the work. For example, action ^ to remove, clean and inspect the strainer was performed before the outlet water box was opened, contrary to the written sequence of the procedure. The { temperature el Jent was removed subsequent to several other actions, contrary to the written sequence in the procedure. Additionally, some steps were signed off prior to completion and one step on the cover of the WO was signed after completing and signing off several steps in the procedure. Although these instances did not reduce the effectiveness of the maintenance activity, the inspector concluded that this was indicative of inattention to detai.l.ard performing to less than management expectation for adherence to procedores. 2.2 Surveillance Observations The inspectors observed and reviewed selected survelliance activities to assure that the activities satisfied technical specification requirements; that personnel adhered to Administrative and surveillance procedures; that test instruments were calibrated; and that test results satisfied the acceptance criteria and when they did not, that the licensee took appropriate actions. The inspectors observed portions of the following surveill?nces: 6-03-4-2, I&C Preventive Maintenance Activity Procedure, e 3-1-4.B DG-1B Surveillance Testing The inspector observed calibration activities on RCS Loops 1 and 2 T-cold and T-hot temperature indicators (TI-111X-1, TI-111Y-1, TI-121X-1, and TI-121Y-1) at the alternate shutdown panel. The I&C technician performed the instrument { calibrations in accordance with the procedure. The inspector ncted an error f mm w ~ PAGE_ [kOF}_.3_PAGE(S)

7 6 (I in the new calibration stickers placed on the loop 1 indicators that had just been calibrated.: The stated dates-of completion were inaccurate-since they reflected 1995 instead of 1994. When this was pointed out to the technician, he made the proper corrections.- However, this incident was similar to other instances in the past when the inspector had observed and discussed inaccurate Jates on calibration stickers, and was an indication of inattention to details. The~ issue was.further discussed with licensee management. There is some confusion regarding whether the calibration dates on the stickers are the ' official" record of calibration. Therefore, the use of these stickers is being re-evaluated by the licensee. The inspector observed the performance of DG 18 surveillance after the completion of maintenance activities. The procedure was conducted appropriately and the diesel operated satisfactorily. There were no discrepancies observed. 2.3 Secondary Component Cooling'(SCC) Valves SCC-A-460 and 461 During the performance of a 24 hour leak test on the air supply to the solenoids for SCC non-safeguards isolation air operated valves (A0V) SCC-A-460 and SCC-A-461, a control room operator (CRO) noted_that one of the green indicating lights for SCC-A-461 was extinguished. Valves SCC-A-460 and 461 are 16" valves that isolate the supply to and return from non-safety related SCC loads in an accident to ensure adequate SCC water supply to residual heat v. removal (RHR) and emergency diesel generator (D/G) 18 heat exchangers and to the control room air conditioning unit. The licensee inspected valve SCC-A-C- 461 and found it open with one of its two solenoids inoperable. The valve was imediately declared inoperable. The leak test was to demonstrate that the backup accumulator will provide sufficient air to keep valves SCC-A-460 and SCC-A-4.51 closed to isolate non-safeguard loads upon a loss of instrument air. Because this valve isolates Scc flow to non-safety related loads, the licensee entered the remedial action of Technical Specifications 3.8 since the SCC subsystem was required to be operable to provide cooling for RHR train B. The valvemas closed and work order No. 94-02997 was issued to perform troubleshooting activities in the valve solenoid wiring circuit. Subsequent investigation revealed that the malfunction was the result of loose + leads on the solenoid. There was some concern that the leads had become loose because of a design deficiency of the solenoid. The licensee's insediate corrective actions involved closing the valve and then installing.a temporary jumper in~the faulty solenoid switch to allow proper operation. The planned long term corrective action is to replace the faulty 34 position switches with-a d fferent design. This replacement was deferred until-the next cold shutdown-following receipt of_ a replacement switch which will not be available until mid-October 1994. -This event was reported to the NRC by Licensee Event-

Report No. 94-015,: Secondary Component Cooling System Outside Design Basit due to _ an Inoperable Non-safeguard Isolation Trip Valve.

{ cAsm 9 6 - 0 4 0 EXHIBIT PAGE_/,3'_0FJj_ PAGE(S)

7 3.0 ENGINEERING The engineering organization provided good support to the plant. Good safety perspective was evident. The inspectors observed preparations for the steam generator tube inspection and repair outage and attended'a training session for contractor personnel who were to be involved in the efforts. The training was well conducted and appeared appropriate. 3.1 Steam Generator (S/G) Tube Repairs After the plant was shutdown, Maine Yankee performed a 1cw pressure hydrostatic leak test, with test pressure on the secondary side, on S/G No. 2 since the primary to secondary leakage had been determined to be in S/G No. 2. The test showed four leaking tubes with leakage ranging from one drop per second in the worst leak, to just a wet spot on the fourth tube. The licensee then performed eddy current test (ECT) using the motorized rotating pancake coil (MRPC) method on these tubes. The results _ indicated that the cracks (defects) were circumferential (cire) cracks initiated from the inside surface , in the hot leg expansion transition region near the top of the tube sheet. The mechanism for the cracks appeared to be primary water stress corrosion cracking (PWSCC).' Maine Yankee then chose to perform ECT of the expansion transition region above the tube sheets in all the tubes in all three S/Gs. The plugging criterion established was that any tube with an identified circumferential crack would be plugged. At the completion of the ECT, all tubes identified cs having cire cracks were (- staked and plugged. The licensee's evaluation performed by the S/G manufacturer's representative (ABB-CE) concluded that because of the lower cold leg temperatures and industry operating experience, an P.RPC examination of the cold leg expansion transition regions was not needed. Following the ECT, Maine Yankee performed in-situ hydrostatic pressure tests with test pressure on the primary side on 6 tubes in steam S/G No. 2 which included the 4 that were initially identified as leaking tubes, and 4 tubes in S/G No.' 1. The purpose of thest tests was to try to understand how the tubes would respond to various differential pressure conditions including the worst case. The test pressures were based on the normal differential pressure across the tubes (1450 psid); 1.4 times the differential pressure produced during a main steam line break (322 psid); and 3 times the maximum operating differential pressure (4800 psid). The test pump had a capacity of 0.496 gallons per minute at 2800 psig. In seven of the ten tubes, the licensee was able to attain 4800 psid. Of the remaining three tubes, two attained a pressure of 2700 psid and the third attained 2600 psid. Two of these three were previously identified as leaking in S/G No. 2. However, the result' of these tests were not conclusive, since they did not simulate an actual pressure transient that could occur during a main steam line break. Maine Yankee agreed to submit a report to the NRC sumarizing their conclusions from the review and analysis of the results of 3he ECT and the plans for the next refueling outage by October 15, 1994. Following the inspection and plugging of defective tubes, the licensee determined that the plant could safely be returned to power operations because, all the tubes were IMUG I EXHlBIT PAGE M OF8 PAGE(S) /

.= 8-JE ' inspected for circ cracks; each tube with an indicated cire crack had been. L staked and plugged; preliminary evaluations showed that'cire cracks are not expected to grow from undetectable to a size of saf_ty concern in one e operating cycle; and administrative controls were in place to require strict. leakage monitoring and evaluation to provide reasonable assurance that the plant would be shut down well before technical specifications.leakaga limits are reached. After determining that the plant might hee been operated with degraded RCS pressure boundary, Maine Yankee appropriately made a report to the NRC in accordance with-10 CFR 50,72, with a followup licensee event report (LER) -Number 94-012, Steam Generator Tube Inspection in accordance with 10 CFR 50.73(a)(2)(ii). During the outage, Maine Yankee agreed to present the results of the inspection, steam generator operating history, actions taken and those planned prior to return ng the unit to power. This meeting was held on August _9, 1994, at NRC's headquarters in Rockville, Maryland. The NRC staff _ found acceptable that the unit could be safely returned to power operations until shutdown for the next refueling outage (scheduled for

  • February 1995) with reasonable assurance of protecting public health and safety.. A NRC specialist inspection (50-304/g4-16) was performed to assess

.the ECT data analysis methods. Overall, 126 tubes were plugged in S/G No.1, 27 tubes in S/G No.2, and 152 in S/G No.3. The total number of tubes plugged to date is 231 in S/G No. 1 (4.05%), 95 in No. 2 (1.67%), and 244 in No. 3 (4.28%) for a combined total c,f 570 (3.3%). Maine Yantee's current plugging analysis with regard to heat _ (- transfer and-power limits conside.s a maximum of 250 tubes plugged in each steam generator. l - In Maine Yankee's Final Safety Analysis Report (FSAR) section 14.12, steam Generator. Tube Rupture Incident, the potential for steam generator tube rupture is addressed with the assumption of a single tube rupture. The licensee pay have operated in an unanalyzed condition during the period prior to the shutdown fo. tube inspection and repairs, however, this is uncertain until-detailed analysis of the inspection results are performed. Meanwhile, the NRC found that Maine Yankee's determination, that the plant can be safely i operated and within its design basis for the remainder of the operating cycle, was acceptable._ The NRC's decision was, in part, based on the licensee's actions to plug all tubes with circ crack indications. L 3.2 Emergency Feedwater (EFW) Valve Leakage outside Design Bases L On August 4, 1994, Maine Yankee identified excessive seat leakage in the steam - generator No.1 EFW isolation valve (EFW-A-338) and EFW control valve (EFW-A-101). Leakage was quantified at 400 gpa and 75 gps, respectively. On the following day, Maine Yankee determined that EFW-A-101 and EFW-A-338 leakage exceeded the-design bases assumptions for their isolation function during the main steam line rupture (MSLR) and steam generator tube rupture (SGTR) events. . Subsequent testing indicated that the steam generator No. 3 EFW isolation-valve (EFW-A-340) also exceeded the design bases assumptions for its; isolation function. I CASE NO. 96-040 EXHIBIT p L PAGE./I_OF k3 PAGF(S) L .-.. --- - -.u n.- .-.a

. - - - - - -. =. l- - ) + l 9 ( Maine _. Yankee identified th'e excessive leakage, while investigating.an anomaly, which occurred on July-16,_1994. While the plant was in cold shutdown, plant Loperators _ observed -leakage into steam genwator No.- 1, but were unable to- -determine the source of the leakage or. subs 6quently recreate the condition. Upon disassembly of EFW-A-338, Maine Yankee found that the disc had been coupled to its air operated actuator _180 degrees-out of alignment. EFW-A-338 is a butterfly valve manufactured by contromatics Corporation. The valve design is unique in that-its disc is asynnetrical and hts a seating surface Consequently, the valve disc - which is similar to that found iri a ball valve. can seat in only one d:tection; and, with the disc being 180 degrees out of alignment, the valve was essentialiy opened. Maine Yankee disassembled EFW-A-101 and EFW-A-340 to determine the cause for the excessive leakage. Maine Yankee found that. EFW-A-101 was not fully seating, and that both EFW-A-101 and-EFW-A-340 were degraded. The original analyses for the MSLR and SGTR events assumed no leakage for the EFW control and isolation valves. On August 11, 1994, Yankee Nuclear Service Division completed an analysis (TAG-NY-g4-042) that established a flow limit through one valve at 40 gpm, and a flow limit through-one line at 10 gpm (based on a thirty minute isolation response time). The 40 gpm limit through any one valve was-based on MSLR considerations for return to criticality, containment environment, and EFW pump runout. The 10 gpm limit was based on the SGTR considerations for increased off-site release and long term cooling. ^ -Maine Yankee's immediate corrective actions were appropriate. Maine Yankee initiated unusual occurrence report (UOR) No. 94-060 and safety issue concern f (SIC) No. 94-006 to document and address the problem. The allowable leakages The past the EFW control and isolation valves were quantified by engineering. valves.were repaired and tested satisfactorily. Interim actions were implemented to ensure that operators appropriately isolate feedwater to a faulted steam generator. Maine Yankee Operational Information Notice Nos. 94-21-1, 94-21-2 and 94-21-3 directed operators to close manual isolation valves EFW-A-102, 202 and 302, respectively, when directed by emergency operating procedures-(EOPs) to isolate feedwater to a steam generator. The bases for this guidance was provided to operators in a memorandum issued on August 11, 1994. After identifying that the plant had been in an unanalyzed condition, Maine Yankee appropriately made a four-hour report to the NRC in accordance with 10 CFR Part 50.72. Maine Yankee had identified several. issues in the closeout plan for $1C g4-006. The issues-included performing a root cause analysis-for the improper assembly of EFW-A-338,- identifying assumptions implied or stated in the plant's safety analysis and determining if these assumptions were validated 4 through testing,: developing procedural controls to test EFW valves and -specifying_a testing frequency, determining the adequacy of EFW isolation valve design, and assessing the need to revise E0Ps to clarify directions to isoitte-feedwater. The inspector reviewed the preliminary root cause determination performed by the plant engineering department (PED). The conclusion from this review was 4 that the maintenance procedure 5-55-10, Maintenance of EFW Air Operated Trip Valve:(EFW-A-338, 339, and 340), did not provide sufficient detail regarding-f fA U EXHIBIT 0 ' PAGE /6 0F11_PAGE(S) i- ~.

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10 the seat / disc and actuator orientation. Recommended corrective actions {- included revising the maintenance procedure, permanently marking valve position, establishing seat leakage criteria, and verifying seat tightness as part-of post maintenance testing. The inspector independently reviewed the maintenance work document associated with the incorrect assembly of EFW-A-338. During the 1992 refueling outage, EFW-A-338 was disassembled under work order No. 92-01746, bc:ause the valve leaked by its seat. The work package indicated that the valve was " upside down." Based on discussions with the PED persennel, the valve was found with the seat facing EFW flow contrary to design drawings. The valve was i reoriented, however, the inspector could not determine whether additional corrective actions were taken. With the valve oriented in the wrong direction, EFW flow would tend to unseat the valve. The inspector was not able to determine if Maine Yankee evaluated this incorrec'; configuration to determine if the as-found condition placed-the plant outside design bases back in 1992, or if other actions, such as confi..ng that tha other similar valves . were not installed backwards, were taken. A significant inadequacy was identified in the test program, which did not detect excessive valve leakage due to the improper installation of one emergency feedwater isolation valve and the degradation of two other emergency feedwater valves. As a result, the leakage through these valves exceeded the assumed leakage in the documented design bases. This inadequacy appears to be a vi !ation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control Program, and is being considered for escalated enforcement. This event which resulted in the plant being outside its design analyses indicated several other weaknesses. The corrective actions were inappropriate with respect to the problem identified in 1992 when the valve was found to be installed backwards. The maintenance procedures did not provide assurance of the proper reassembly of the valve, and post maintenance testing was not provided to ensure that the valve was installed correctly. 4.0 PLANT SUPPORT Plant support activities in the areas of radiological controls, security, and emergency preparedness were conducted safely during this period. The inspectors monitored work practices, and conformance to requirements and i procedt.res. 4.1 Radiological controls Ir spectors-routinely reviewed radiological controls including organization and management, external radiation exposure control and contamination control. The inspectors also monitored the licensee's radiological work practices, and conformance with radiological control procedures and 10 CFR 20 requirements. l Activities to support the S/G tube inspections and the repair outage were well managed and performed. -( CASEND. 1-96 040,, EXHIBIT PAGE,/7 0F M PAGE(S) ~~ ^'

R I I -- 4.2l Security The inspectors verified:that security conditions met-regulatory requirements, the requirements of the physic al security plan, and complied with approved procedures..The inspectors observed set:urity st'affing; protected and vital area-barriers,; vehicle searches and_ personnel identification; access control, badging and verified that they were in accordance with requirements and that appropriate compensatory measures were taken when required. i 4.3 Emergency Preparedness The' inspectors verified that the emergency response facilities wors well maintained and kept ready for use in emergency situations. 5.0 SAFETY ASSESSMENT / QUALITY VERIFICATION Missed Technical Specification Surveillance r During the course of a QA/QC surveillance (Number g45-033) on the fire j ' detection system; the QC inspector identified a d screpancy between the i surveillance procedure, 3-1g-3, smoke and Heat Detector Testing, and the detectcr location drawing (FE-51A). The print showed 14 detectors in zone 46 of containment at elevation 44' 6"_while the ;urveillance procedure only i-included 13 detectors in the zone. Field verification by the licensee showed ( - that there were actually 14 detectors in the zone and that detector 46-2 was not tested by the procedure. 'he Quality Assurance Program department appropriately issued a corrective action request (CAR No. 94-030-5) to document this. discrepancy as well as the recessended and implemented corrective actions. This missed surveillance of the smoke detector in zone 46 of the containment is a violation of technical specification (TS) 4-12-A which requires that all smoke detectors in the plant be tested on a periodic-basis. The licensee was L unahl.a_to immediately determine when the detector was last tested. However, since TS 3-23 which specifies what detectors are required to ba functional during plant operations does not include the missed detector, the safety significance was low. In accordance with the NRC's Enforcement Policy in_10 CFR Part 2, Appendix C, Section VII.B(1), this violation of technical specification 4.12.A is not being cited because the-safety significance was low and the violation was identified by the licensee. - Additionally, the. violation was not repetitious and could not have been prevented by corrective - actions to a previous violation..The licensee immediately tested the detector upon discovery and initiated actions to revise the surveillance procedure to ~ include the detector. The inspector reviewed the licensea's actions regarding_this issue. The identification of_the problem by the QA department demonstrated good performance. The licensee took prompt actions:to rectify the deficiency by -testing the detector and initiating actions to revise the procedure. Additionally, the licensee initiated licensee event report (LER) No. 94-014 ito repori the missed technical specification surveillance to the NRC. The 1 ~ inspector expressed no other concerns regarding this issue. "F CAEN0r. l*9 0 TO4 0 ' PAGE /6 0F_2).PAGE(S) b-aw-g----+rgee .+mw+- e-.ww.-i.+r-i--,,*w--eemsed a.,ise s +,,, m oi-e,--aii er----,- ,s.- ,p--y-g-s as- ,w g-

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l 12 ( 6.0 ADMINISTRATIVE 6.1 Persons Contacted During this report period, inspectors conducted interviews and discussions with various licensee personnel, ir..luding plant operators, maintenance and surveillance technicians and the licensee management. 6.2 Summary of MRC Activities othar inspections conducted during this inspection period include: Motor Operated Yalves (MOV) Phase Two (50-309/94-14), steam Generator Tube Outage (50-309/94-16), and Engineering (50-309/94-18). During the inspection period the inspectors conducted backshift inspection on July 18, 22, August 1, 9, 15, 18, 22, 25, and 26, and deep backshift inspection on July 23, 30, August 8, 14, 20, 23, 28, and September 3.

  • T. Martin, Regional Administrator, Region I visited the site on August 17, 1994.

J. Linvi11e, Chief, Region I Projects Branch 3 visited the site August 1-2, 1994. v,ng these visits, Region I managers toured the site, and held discussions *.th inspectcrs and licensee management. 6.3 Interface with the State of Maine ( Periodically, the resident inspectors and the onsite representative of the State of Maine aiscussed findings and activities of their corresponding organizations. On August 11, 1994, the inspector attended a presentation given by Maine Yankee to representatives from the State of Maine. The presentation, which us focused on the steam generator tube inspection and repairs, was similar to that presented to the NRC on August 9, 1994. The information presmted is provided in the NRC public document room with NRC letter dated August 19, 1994. 6.4 Exit Meeting Inspectors perfodically held meetings with senior facility management to discuss the inspection scope and findings. At the conclusion of the inspection, the inspectors also presented a sumary of findings for the report period. EXHIBIT PAGE /7 OQJ_ PAGE!S)

I DICLOSURE 2 FDERAL REGISTDt NOTICES .O ( CASEND. { ]. 9 0. O s f' f EXHIBIT PAGEJO OF >3 PAGE(S)

= O suPewsorfAmy promasAnuw: N f Commission pubushed a policy statement On b implementatino of a two year trial program to eBow selected enforcement conferences to be open ro public cAmarvadon on July 10.1Mt2 (57 FR 307621. De purpose of b trial program was to duermine whahar to maintain the current policy stated in Section V of b "Geners! Statanant ed Policy and Procedure for Enforcement Action." (Entcuoement Policy) 10 CFR Part 2. Appendix C that. "enlottament conferences wiu not normelh be open to b public."or to odopt a new policy that would allow most soforceawat conferences to be open to attendance by all members of b public. C-mts were required to be provided to the Comrnission on or before the compleuon date of b trial program. A correction to the original nodcs was issued on July 17.1992 (57 FR 31754) to correctly identify the scheduled ( compledon of the trial program as July 11.1994. On May 13.1994 the Executive Director for Operedons directed a reexaminadon of the NRC enforcement Two Year Titel wn for b Conctucting Cpen nh staff.y a Review Team of senica dR As part of this com rebensive Conhos; Continuadon of Trial ,, view of the Enforcornent P ,the "+ C NRC intends to consider b issue of AcceCv: Nuclou Regulatory whethee the Commisalon abc 4)d Commluion. establish open enforcement confomacas Acncer: Su lement to Policy as the normal practics, la the interim. Statement; nunuaden of Trial the NRC is continuing b o PN8'*8"- ,og,,,,,,,,,,,,,,,,,,,,g p,,,,,, pending the outomne of b suesaARVIDe Nuclear bgulatory Enforcement Policy Review. The Review Commission (NRC)isissuing a Team intends to com IMe its review of supplement to its two year trial pro 6 ram the Enforcement Poll in early 1995. for cond open enfortement As pan ofits miew of the gg, conferences. e p rpose of this Enforcement Policy, the NRC intends to lement is to in rm b public t. Issue a Federal Register nodes solidting g su th C's continuaden of the trail public comments to assist the Review g program until the commission acts upon Team. This nodce wsilincaude the NRC staffs recommendations solidtin cornments on the issue of regarding open enfortement OPen en orcement confnences. Dated at Rochille. MD, this 13th day cd I-* g confersnope, FOR PVRTNE A edCMADON CcwTAC7: 'N # James 1Jeterman. Director. Office of For the.%deai Replaton Commassen n Enforcenent. U S. Nuclear Regulaton I'2*" U'6*rtu* g w Commitilon. Washiripon. DC 20555 Derector. Offxv of frifortement Q (301-504-2741). lFR Doc 94-17,00 Filed 7-ira e 45 aznl cL CASE ND. ] - 911 - g 4 g ~ - - --.,.-...__.__.-.-.4 em-,--e w e-r-

W-Ateauassa Sand esamendter m of6erum u.g, Nedase rm Wa6bingtse,DCapesa. ATTN:

Dedettes and Seretse Bronsh. Hand dehver esameshim Om White FlintNort,11888!teekville pne, A*dville, bel between P.48 aJa. Io (13 pm.Federalwor6deye, copies of commenta may be emersined at the NRC pubbs Docuneat koon. 2120 L Sent, NW.flawer149,1), Washinglen DC mensmusesoammeneenrace: ansaIJebssmaa,Otroener.OSee of M ~ U.S.NasinerRegulatory Commissism,Washingene.DC anus (305486#e). suenammmmeseemeanom i Ansflyensd i l The NRCs serventpebsy as esferemment aseforemose le addrueed in sesehen y er the latest sevisten to the

  • GeneselSteassentof and preaden dw asfuesseset (tatuesmaatpubsy11e CFR part 2, t

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  • eefuemment wulnot asemallybeapento the public?

However,to Commission has decided toimplementa estalpseyan to ,(- sumetpabeysettisagesd14 estamesewheestto meestaan se seismemanaemaiseness er to adopt a l newpaksy thatweeld alisw amot l Two VeerTMuipvoyssa ser enfanumsatenminemensto be open to CeneusemsOpen Enforumment ensadnero by aR esehere of the public. CenterenesesPoesy timesament-pousy h AansenNesleer Regulatory Camaissima, m asnesspebsy sestament, TheNRCleM e two yur l'telpeupon to enew manmsmTheNesleer ehesseensaaf esfutement Cserassies(NRCIleissesseeh esmAusnesa.TheNRCwGImonitor the stetnesseento W %two year Wiel propen to nuew selossed a estebhnhapennenamepehey for enfeessment samlumesse to be opento osedsetens enfusament atasadanosby aR usemebers of the emelsesse en as amemment of penne,niepoucyet.une t the6diourlagmeessia: the ertalpreyam (1)Whsiker tolastthat the andladernethe of how to get eseismasswee span kapested the taformatlesenageoadas open NRCs andMy to esadast a meaninsful enfesoament:- a coalamass and/etWt the NRC's s&Testhis trial lae5ectiveen 'N'""""8 p"#'3" July 18,1958, comments sa the (1)Whether the spea osafemace propram ass being Mved. Submit lapeceed the llamasee's participation in gpw U.3 T - O((*iG'I commaste se er before the c:=ly b coefsmem of the trialpreysm scheduled for Ju (3)Whs6er the NRC expended a 11.1983. Ceaunects received after this sisnaicant emeint of resources n / d*te wt11 be considee.d if it te practical mHne tbs ceafervoce public and EXHTBli 0 to do so, but the Comninion is,sble to H)* " wet of paue intuuon '"""**"Id*'*6"'*'.**" opening the enforcement confetence PAGE WOF 23 PAGE(S) received on or before thle date "C

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-r- - -= was es mome=== m se est,ese a pausemel 8igen,bessere,poens % h l Eedessemees Casemmes esme m aa! renseese. ee.eH est larger i Ealersement osadoremese wul met be h*'P64ala, et k whisde em 18" he pomened, and shes I i ,. w she puhts (m admeenen au==** d *e renamies erp d owepsee,mmene ma, no mm ed, t mes esmesspistee-besasem. god reglemal eSee esel asensus w i (1) W be tahamegehst as n, Aamendes Open Redwesomene esednes the esfeseemana eenlarense j males against as to aseles though nel e rovideaL at if the Cemfuesses p is amendense me ressemel vndual. tune es Pressen. emisesmens senderence i l wbether asinevident has commened As seen as itla delannemed that as enferoament confersace wGl be spas to wtB emedese to be a weles between mm b NRC and the h Whos 6e I ORavelves alsidasant pereennel ["%g yCdaraDF, aderessoa esahreene to epse for - i not failures whose the NRC 4 as uasted eg Pubus obearveten,N la met open for "8e 98848880808 I pubha thei theinevideal,[e)lesdyed

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.ed,re , no.e,m m ne aee Uthe be med a he esteirag =h wulbe provided es apparamily to i ceaferseas wtB be esadeseed et a i retetively smau beansee's feathty. (1) poseed a he pubts rutett wness asumente enemymme}y l Plaally, with te ofthe Deeemania,em, to se reWenaleSee.These esaments i r.nece eve Dtresser opereneen. in Tes+= taneshame w med wul embeseeensir be forwarded is the eaforosseset eenkreases wdlmet be h)Tou free electroans beMons board Direcaer of the Ottee of Baaernement for

  • " in to the pakte to spesial easse rev6ew and seasiderseen.

are emese has been aboue aber estabilekseest af the ts54ee Deted as assiness, nm, she rib der of Jair the bemeAt ofpakke mesesse systeman, the ymbale eher eng

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, ( ce the she ponendelhopest (sot)e speam.nsk eas es aheets a recordlag of Forabe NamiestS h # - e emisseemaat asemais a g open endmeamese sament3. chew. I Pareenlar esse-eraferessen. The NRC wtD issue emother L _, efete cessemassa The NRC wul otrtve to esaduct opse Federal Regleser mense eRetthe neuese. [pHm. 54Mm Fund FM h) erderoement eenferesses dartes thee mesesse eyeesmo are estabilehedsi m,m, i two. year Iront poespas la asserdanes To endes the $5tCla makteg-1 wie ihe f.newing er genne-e e,,e. gem e,,an 1 Appe===ana6 as passent elaR .u(o ) abeere afani F . sed,,.,me..ss.iem.d. a,mesmees. M 54 ea-f- s.o. sed e*adsened by sheNBC will be ope ter atuadhe e prseenlarendorsem P^th de- = =s senhe== shmed ase me m.ame adeel Corr 8CtlOnS hd=8 **'- (2) At Imat ano apes andersamm8 ideme18ed to to mestes sedes vel sr. No. He conference w6ft be sendseesd to ensh of ameesmates the eN enforcement h esal Whose; and semisemos as Aster than tre bestness Petery. My 17. test (3) e enforcement asedeemese doye preeres e n eage,ee m en

  • nu te condussed with a voetely of to eseguemen, trPeo af beemesse.

88UCLEARIWGLA.ATORT To avoid poiestialbias k abe IIL C880'et of Opse Enferoament 00madesses 6elecean process and to attempe to med C**8 se"* the three goals stated aheve,every b econdense wt6 earrest prootsen, Two. Year Ttted propen ter fourth elebie seducessment semieresse enfersencat seafweseos wdl centems Cenemsig Open Enferoement tavolvtag one of three setegenee of - to aanneDy be hand at the NRC Conferensees poesy sensament beonaees wel marmeDy be open toIbo etBces, b4 embers of b pubbe be Pubbe darteg the anal prepen. aUowed access to b NRC regional 3 - However. la sness where therm is en efBees no anand spee saloromaaat Is notice deceaset 9116233 beginnLng ortsotas edludicatory proceedlag with cederences la accordamos wrth the en pawe Mrst in the taene of Priday, one er more notorveners, enfertammet " Standard Opereung Proceduree For Idy.10. llet, os page scra2. la the mdersocos tavolving laseos reisted to LM., Se:ertty Scypert For NRC W eolana, mader oaves. bestanins wbject maner af the ongoing Heanage And Meennss* pubhshed in the Bhb he. *Ny 11. les: shorJd Acetion snay also be opened. For Novendier 1. tart 136 m 56::31). The se road *)sh1L1est". e pciposee of LMe irsal program. the prwedurve provide tbat Msifxs may be suan come essee ) CASE ND. 1 9n- 04 0 EXHlBIT I PAGE ch3 OF13 PAGE(S) w-d -%e + e < emew-r e eeew-n-w,wn+-=,--=me,-_-

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~,. -. _ n) my - ML z)1.4.9.i MaineYankee afuasti f tCTaiC4Tv liNC E ' 6 72 go p,Yg,Upg 329 BATH ACAo a BAUNsWICK. MAtNE 04o11 * (2",ll% j P.2SPOND B.Y N4 NRC DUE dAfI N4 October 28, 1994 MN-94-109 JRH-94-253 UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555

Reference:

(a) License No. DPR-36 (Docket No. 50-309)

Subject:

Maine Yankee Licensee Event Report 94-016-01, Emergency Feedwater Isolation Valve Leakage Gentlemen: Please find enclosed Maine Yankee Licensee Event Report 94-016-01. This report [ is submitted for information only. k-Please contact us should you have any questions regarding this matter. Very truly yours, Q.hs James R. Hebert, Manager Licensing & Engineering Support Department JRH/ jag Enclosure c: Mr. Thomas T. Martin ( Mr. J. T. Yarokun Mr. E. H. Trottier 7 Rob ' ADDU l y Mr. Patrick J. Oostie I.E3 ' $$m " i m .emme ( l :n

  • u lT l

CASE NO. 1 9 ti - 04 0 \\7 EXHIBIT /d La\\94mn\\t4109 PAGE / OF 6 PAGE(S) dr). q 4n .q-4/Gtppgd& %?' 4

WC FCAM M6 U.S. autLLAR REWLAIORT C38811310m APPR0vtD 87 CMS 40. 3150-0104 (5 12) CIFI's*1 5/31/95 C, Elf!MAf!D SUA0t4 81 ,tP0hlt T: C* DLT WITM: J LICENSEE EVENT REPORT (LER) rEa'o"l'e".'$ $m ' U*E'u'isYlAftt96i ' 3 ( THE INFORMAflCM AhD AtLORCS MahaCtkiNI BRANCM (>=68 1714). u.3. 4pCLEAA ttGULAfCRY COMM1111CN. (See reverse for reewtrea necer of digits /enefacters for eacn bloca) WA1MlhGTCN. DC 20$55+a001. AND TO THE 8APERWCAC i REDUCTION PtoJECT (3150 0104), OFFICE of a.anactutet a=D sco,tt wasu m t?= Se refe3 FACILITY 4AfE (1) 00CIIT sp ett (2) PAGE (3) Maine Yankee Atomic Power Company 50-309 10F 4 TITLI (4) l t Emercenev Feedwater isolation Valve leaklee i fYFNT DAff f5) LER W!a8BER f61 RE?TT DATE f 71 GTHrf rAf f t f tf E1 INVOLVED f R1 I h N M alon TH GAf TLAA ?tAA ronfN oat itAR 01 10 28 94 $ 3" "' 7 ""*" 08 04 94 94 016 (FtRAflh6 TF11 t[ Pet 11 Stapf 7TED DWillaaf TD TE tr0UIREMNY10F 111 CTI l? f C4ev eae ee acae) fill 3 Im0f (9) 20.402(b) 20.405(c) 50.73(a)(21(tvl 13.11(bl 20.405(alfilft) 50.36(el(1) 50.73(alt!)(vi 73.flic) pagA l LfWL (10) 20.aC5fal(11(1#1 50.36(cit!) 50.73(a)(2)(vit) X OfMLN g pc 3 #; 20.405(a)(1)(Itt) 50.13(a)(21(*) 50.73(a)(2)(vtit)(At ($pecify in @M N[b y a'1 M T U; 20.405(a)(11(tv) 50.73(a)(2)(til 50.73(al(2)(vittl(B) Abstract below c y-20.405(a)(11(vl 50.13(al(2)ftet) 50.13(a)(2)(a) ' N De tittt$rt contact rar Tw!1 tra rif t wt IELEPMoht 9uM8tR (Inclues Area Coce) George Stowers, Nuclear Safety Specialist (207) 882-6321 [ con,tnr o=r of =r enn tAcw c ner=r rariar aritniero in Tur$ arret fin ( .Aust a A sysTra Comeent=f mAmurACTunta CAust systtw compontNT mAnurACTURER 0 BA ISV C630 Y X BA FCV F130 Y X BA ISV C630 Y g SUP#t EDf 4 T AL R E PM T D Pf C'ID f la ) ggpgg7gg e tw Sav VEa# fts Stanission y,a NA NA NA (!f yes caiglete (IPtCitD $USM1311CN CAft). DATE (1s) ABSTEXT (Limit to 1400 spaces, t.e., approximately 15 stegle soaced typewritten lines) (16) At 1220, on August 4,1994 with the reactor in a cold shutdown condition, plant operators determined that the Emergency Feedwater isolation and regulating valves for #1 Steam Generator were leaking by. Subsequently it was determined that under accident conditions which require isolation of Emergency Feedwater, valve leakage could exceed Safety Analysis assumptions. Maintenance activities were performed to reduce Emergency Feedwater valve leakage to within acceptable limits. In addition, administrative controls were implemented to ensure Emergency Feedwater leakage is maintained within the bounds of Safety Analysis assumptions during accident conditions. A root cause investigation identified inadequate maintenance procedures, and inadequate post maintenance testing as the main causal factors for this event. The safety significance of this ' event was not initially known, therefore this event was 'triginally reported under the provisions of 10CFR 50.73(a)(2)(ii)(A), an unanalyzed indition that significantly compromised plant safety. However, recently completed .aalyses show that plant safety was not significantly compromised by this event. Therefore, this event is now being reported for information SNf5lT /4 2953 E CE 'IO. )..g,, > - 1 0 PAGE k OF 6 PAGE(S)

  • C M" 36 (5 H)

oc rga au e novc g g so oio4 u.s. acwn itsurant camission Nif!@oni!$t$n" celt $a"Uou[sT!#[oYo EI LICENSEE EVENT REPORT (LER) U!raimNE0iN[c$os'EEtN"Na5 l g g. g s g y g assutA,i,on g j g TEXT CONTINVATION g W J % a % M W e % lif FAfftffY maasr ft) enrvrf masars f f) 11R -a*4 fa' PA&f f31 SEQUENTIAL REVllloM y Maine Yankee Atomic Power Company 50-309 2 0F 4 016 -- 01 94 Tm t r e w. -... mau, me. u.. sm anni e e.. or we r. isan on on August 4, 1994 Maine Yankee was in a cold shutdown condition making preparations to ) restart the plant following a maintenance outage to correct Steam Generator tube leakage. At approximately 1220 while perfoming a leak test of the Emergency Feedwater (BA) Isolation Valve (ISV) for #1 Steam Generator, it was determined that EFW-A-338 allowed about 400 GPM flow when in the closed position. Leakage past #1 Steam Generator Emergency Feedwater Regulating Valve (FCV) EFW-A-101 at a rate of 75 GPM was also identified. EFW-A-101 & 338 are installed in series and both receive a signal to close on low Steam Generator Pressure. Subsequent investigation determined that the actuator for EFW-A-338 eas coupled to the disk 180 degrees out of alignment. Maintenance activity was initiated to reposition the actuator / disk for EFW-A-338. Leakage past EFW-A-101 was corrected by external adjustments. At 1700 on August 5, 1994 it was determined that Maine )ankee's Steam Line Break, and i r I ' team Generator Tube Rupture Safety Analyses assume no leakage past Emergency Feedwater solation and regulating valves. Therefore, at 1957 the NRC was appraised of the ( situation via the Emergency Notification System in accordance with the provisions of 10CFR 50.72(b)(2)i. A follow up notification was mado at 1151 on August 6, 1994 after leakage of approximately 30 GPM was identified past the Emergency Feedwater Isolation Valve for #3 Steam Generator, EFW-A-340. This leakage was later determined to be caused by erosion and the seat subsequently replaced. I On August 11, 1994 comprehensive leak testing was performed on the Emergency Feedwater isolation and regulating valves for all three Steam Generators using portable test equipment. These tests revealed leak rates for various individual valves and combinations of valves from 0 to 36 GPM per Steam Generator (See Attachment A). Safety Analysis sensitivity studies conducted in parallel with these tests established the following revised acceptance criteria: 1) Leakage N individual isolation / regulating valves less than 40 GPM; 2) Leakage past both the isc. ' ion and regulating valve to each Steam Generator less than 10 GPM; and 3) Zero leamage to a Steam Generator with a tube rupture eithin 30 minutes. 4 The following immediate actions were taken to ensure conformance to the revised acceptance criteria: 1. EFW-A-101 was adjusted to reduce leakage past EFW-A-101 & 338 to less than 10 GPM. 2. Plant Emergency Operating Procedures (EOPs) contained adequate guidance for identifying and isolating EFW valve leakage prior to this event,(E-0, Emergency f Shutdown From Power or Safety Injection, Rev. 5; E-3, Steam Generator Tube Rupture, L Rev. 10; and FR-H-3, Steam Generator High Level, Rev. 15). However, as an added \\ precaution, additional administrative controls were implemented to reinforce the need for operators to ensure any leakage to a ruptured Steam Generator is terminated within 30 minutes. "8" CASE M. 1-96-040 2054 canc ? ne < ~ ~., ~ .1

MC FORM 3464 U.S. mELEAA RLEUTORT C3ellL31Du APPSOVED If Weg no. 3150-0104 ($.lt) CIPltt! $/31/95 C f!T!*TED staDEN PCR RE$PONSE TO COMPLT VIT) { THIS INf0RMAT104 COLLtcTION A(0U[57: 50.0 HR$. { [ LICENSEE EVD(T REPORT (LER) ' U 0FOR N ' A j T [CD 0$ AN i TEXT CONTINUATION (ames me). v.s. sucttAn awuuToer comisstem. WAsMih6 TON DC 10555-0001. AND TO TW PAPLAWORK U.*a E r,a E ' Err. M I N nc E $ # u ruu m enn m c-- - - > - -a m tn -w ts Paar m $10u[nflAL a[vi3IDN Maine Yankee Atomic Power Company 50-30g 3 0F 4 94 - 016 -- 01 Ttrf (nr mere e,ece se r wirw use emitional seeies of me rer. au) (17) In order to assess the potential safety implications of this event, detailed studies were performed of Main Steam 1.ine Break, and Steam Generator Tube Rupture events using as-found EFW leak rates. These studies were conducted using the same NRC approved methods normally, used for our core reload Safety Analyses. The results of these studies indicate that the i consequences remain bounded by analyses reported in the FSAR, and the impact on the health and safety of the public would have been negligible. A root cause investigation of this event revealed the following causal factors: Maintenance instructions lack sufficient guidance to ensure consistent proper assembly o of the EFW isolation valves, An erroneous belief that specifying the functional testing required by the IST program o(. for a given component, ensures adequate post maintenance functional testing of all the component's safety functions. k Actions to reduce the probability of recurrence which are planned or have already been completed include the following: Maintenance procedures have been revised to provide additional guidance for assembling o the EFW isolation valves to ensure consistent, proper orientation of the valve disk, body, seat, and actuator. Verbal and written guidance has been provided to increase awareness among those who o develop, review, and approve functional tests, that the functional testing required by our IST program should not be relied upon to ensure adequate post maintenance functional testing. cur". 1 - 0 " ". G 4 0 EXHlBIT /d PAGE #7 _OF 6 PAGE(S) i a 2955

Attachment A Emergency Feedwater System (SimplifiedDiagram) rW4M e5+1M F w lin -M--#1S.S. p r e ans w w + art 3,, g-N-- en s.s. p a, ww+sas rw+ art"" . P.ase -N-- 83 8.1. f Test As Found Results As Left k. Jia Valve (s) Closed 08/04/94 08/10/94 08/11/94 1 EFV-A-101 75 GPM 36 GPM 0 GPM EFW-A-338 400 GPM 20 GPM 24 GPM EFW-A-101 1 338 15 GPM 0 GPM EFW-A-101, 338 & EFW-102 0 GPM 0 GPM 2 EFW-A-201 0 GPM 0 GPM EFW-A-339 17 GPM 17 GPM EFW-A 201 & 339 0 GPM 0 GPM EFW-A-201, 339 & EFW-202 0 GPM 0 GPM 3 EFW-A-301 0 GPM 0 GPM EFW-A-340 36 GPM 23 GPM EFW-A-301 & 340 4 GPM 0 GPM EFW-A-301, 340 1 EFW-302 0 GPM 0 GPM CASE NO. 1-96-040 EXHIBIT /# PAGE I 0F.$__.PAGE(S) c-C 2056 .}}

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