ML20203G599
| ML20203G599 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/09/1995 |
| From: | Wetterhahn M Public Service Enterprise Group, WINSTON & STRAWN |
| To: | Logan K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20203G565 | List: |
| References | |
| FOIA-97-325 NUDOCS 9712180166 | |
| Download: ML20203G599 (109) | |
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., 2 Alr. Keith Logan
- United States Nuclear Regulatory Commission Office ofInvestigations 475 Allendale Road King of Prussia, PA 19406
Dear Mr. Logan:
In re.(ponse to your recent request, the following documents are provided:
- 1. Procedure Revision Request R07326, to restrict the number of RCP's in operation while in blode 5 to no more than 2.
- 2. Chart depicting 1993 salary review information for separated individuals.
- 3. Chart depicting 1994 salary review information for separated individuals.
I request that Documents 2 and 3 as identiGed above, be withheld from public disclosure in accordance with 10 C.F.R. V 2.790(a)(4) and (6). 'These documents have been stamped "EXEhlPT FROh! PUBLIC DISCLOSURE PER 10 C.F.R. 2.790." These documents contain information, the disclosure of which would constitute a clearly unwarranted invasion of personal privacy, and information that is con 0dential to PSE&G. The identined - documents contain the type of personal information normally kept confidential. bloreover, . rel ease of such information could increase costs to PSE&G, and subject it to possible CERTAIN ATTACllMENTS CONTAIN INFORMATION TO BE WITilllELD FROM I4 PUBLIC DISCLOSURE PURSUANT 1010 C.F.R. f 2,790 D
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9712190166 971210 PDR FO!A KEENAN97-325 PDR .
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. Mr. Keith Logan:
' August 9[1995; -
Page 2 '. , litigation. These documents'have been treated by PSE&G'as confidential and are strictly controlled to prevent disclosure. . If you require further information in conjunction with this confidentiality - request, or ifI can be of further assistance, please let me know.
- Sincerel ,
Mark J. Wetterhahn .
. Counsel for Public Service Electric & Gas Company
' Enclosure.i s
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' CERTAIN ATTACilalENTS CONTAIN INFORAIATION TO BE WITilllELD FRO'.T!
PUBLIC DISCLOSURE PURSUANT TO 10 C.F.R. s 2.790 _; . .. a
a i t , , i TO: M. Morroni Manager Salem Technical Departmer t FROM: C. Lashkari 8 System Engineer Salem Technical Department
SUBJECT:
CRDM PENETRATION CRACKING ISSUE DATE: March 11,1992 Attached are my trip reports from attending the NUMARK AHAC600 group meeting on February 19,1993 and the NRC meeting on March 3,1993 on this issue. Also, available are handouts from the WOG Material subcommittee meeting in Pittsburgh on March 4th and 5th, which was attended by Dr. Jim Perrin from L:JB. Three plants, DC Cook Unit 2. Point Beach Unit 1 and Oconee have volunteered to inspect their Rx head CRDM penetrations during 1994. As a prudent action and to protect its investment, Salem 1 should plan an inspection during 1R12 (Spring 95) subject to the NRC approval of the NDE acceptance criteria and successful development of NDE and repair technology by the vendors. During presentation to the NRC by N'JMARK, the NRC indicated its disappointment regarding industry position on Rx head leak and boric acid wastage based on Turkey Point, Salem and other plants. Bill Russell, Associate Director-NRR, wants industry to explain how plants operating at pretent are considered safe when they do not have a reactor head leak detection system. Mode 3 walkdown and visualinspection (even with new shroud doors) are considered inadequate because plant may have a good operating record of 300-400 days. NRC wants above position submitted by WOG, CEOG and B&WOG. It then wants to review and document its position why it is safe ! to continue to operate until a plant conducts its inspection. The NRC Senior Management seems to be reacting to the external pressure on this issue. Due to the complexity and long term nature of this project, it is recommended that a l Project Manager be assigned by E&PB, who will prepare and implement a strategic l' plan. This strategic plan could be developed from elements of the attached action plan. Senior Management needs to be briefed regarding resources which may be j required during 94-95. Copies of the handout are not being attached to this letter, but are available if anyone requires it. p\
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- CL/gb) 1 x C: - - General Manager Salem Operations
. Manager Salem Operations:
Maintenance Manager ; Salem
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- Rad. Pro./ Chemistry Manager ' Salem -
_ Manager.;- Statico Quality' Assurance i Onsite Safety _ Review Engineer
- Manager e Licensing'and Regulation -
. Manager - Nuclear Mechanical Engineering - .,
Manager Salem Projects
~ Outage Manager - Unit:1 Outage Manager - Unit 2 1
1 H. Trenka-H.- Berrick w/ attachments > L J. Wiedemann. k. s N-1. t v
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r ABB ASEA BROWN E0VER. January 15,1993 0014-00073 14 Public Service Electric and Gas Company End of Buttonwood Road Mall Code N,27 Ehncocks Bridge, N108038 Attention: Mr. Dilip Bhavnani Senior Suess Engineer Stress Anslysis
Subject:
Copes-Vulcan Valve Opening Time for Pressurtzer PORV/SV Thermal Hydraulic Analysis Gentlemen; Per your telephonc tcquest of Jar.um la,19M, the followingi technical inforrnat on sent to clarify the technical input Jsed for the Pressurizer PORV/SV thermal hydraulic analyds. The original thermal hydraulic analpis was perfctmed by both ABB Imlell and PSE&G 1985. This analysis is the basis that was used by the NRC in 'he:r review and le r:fere in the NRC Safety Ee.luatien Report that we subtruned to PS2&G 10, 1990. on May
; - The 1985 thermal hydraulic analysis for the PORY two-valve case utilized a stroke t 150 ms. The NRC had acknowledged a the SER that the 150 ms value used in was conservative. It should be noted that the 1985 analysis assemed that a loop seal was n pmsect on the charge side of the PORVs.
The analysis that was recently perfonr.ed for the Engmecring Evaluation to demonstra operability (ABB Impell Prcject 0014-000e4) included that effect of a loop seal at the PORVs and utlized an opening stroke time fer the two-valve case of 500 ms (0 . This value is consistent with the EPRI tett data for the Copes-Vulcan PORVs (Re The current analysis (ABB Impell Project 00144X)073) which removes the loop daimag also uses an opening stroke urns of 0.50 second:. Based on current inforTnation received from Copes-Vulcan (Reference 2), Copes-V advised PSE&G that the valve has a " quick-opemng charactenstic," i.e.,90% open the stem travel. I' ting an opemag time in the range of appmximazdy 0.25 seconds to 2 seconds does n sigmficantly affect the rnagnimde est the !oads developed. ABB impell Corporation /g 7 770 Cmone bee M t! 8137? 34to PO Das $316 f ax Ret 3%3Ma Fmewtaede hM 0t F319316
O A._._B..B_ . A88 Impedi Corporation r Mr. Dilip Bhavnaal i Scalor Staff Engineer Public Service Electric and Gas Corppany , January 15,1993 0014 00073 414 ; Page Two : If *.he vahe opening tune is 2 seconds, which was discussed at previous project meetngs is correc, the use of a 0.5 second fully opened condition is reasonable (i.e.,0.3 X 2 seconds = 0.6 seconds). If the valve opening ume is ' second due to mod 1Destions that may have taken place to the diaphragm then the opening nme would be on the order of 0.3 seconds which will still not dgnificantly imroct the magnitude of the loads developed. Based on the current information and EPRI test data, ABB 1mpell believes (nat a value of 0.5 seconds for tne PORVs is a reasonable and justifiable value for the currem vajve configuranon. If you require sddidonal infermadoa please call rac at 508) 370 3403. Very truly yours, M Janus A. Flateny JF/mp cc: Tim Taylor
A_.._B._B.. ABS impeil Corpotatit. Mr. Dilip 3bavne:d Staff Engineer Pablic Service Electric and Oas Company January !$,1993 0014-00073-014 Page Three
Reference:
1. EPRI PWR Safety uid Relief Valve Test Prograrn Guide for Application of VCre Program Results fer Plant Specm Evalcations. Revision 2 Interim Report. July 1962. 2. Copes Vulcan leuer to PSE10 dated December 9,1992 at:ection Chris Zehrer with atta:hed PSE&O Telephane Cor:ference Memorand m dated 11.'682.
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Pubhc Service Electne and Gas Company- P 0. Box 236 Hancocks Bndge New Jersey 08038 .I Nwlear Depannent MEC-93-1021 l Tot J. Wiedemann I Salem Technical Department FROM: Dilip Bhavnani Salem PORV Project
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SUBJEC1s COPES VULCAN VALVE OPENING TIME FOR' PRESSURIZER ! PORV/SV THERMAL HYDRAULICS ANALYSIS j t CATEt- January 15,-1993 Per our discussions yesterday at the Salem PORV DCP meeting, !
- attached letter from ADB Impell provides the rationale for the !
- use of valve opening time of 0.5 seconds for the present i analysis effort. It should be noted that opening time l utilized is consistant with EPRI test data. Discussions with !
valve manufacturer indicates that valve has quick opening . characteristic i.e. 90% open at 30% of stem travel. In t addition, utilizing an opening time of 0.25 to 2'secondr does ; not significantly affect the magnitude of loads developed. If there are any quest.'ons please feel free to contact me. g 4 . i DBaimv j ATTACHMENT C2 J. Ranalli C% % 943stou* T. N. Taylor M. Moronni
- M. Danak' '
D..Longo J..Flaherty (ABB Impell) R. Coward (MPR Associates) MEC File i
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ABB ASE A BROWN BOVERI January 18,1993 0014 00073 018 j [ Public Service Electne and Gas Company End of Buttonwood Road [ ,. y hiail Code N 27 flancocks Bridge, New Jersey 08038 ATTENTION: hir. Timothy Taylor Project hianager f' ' SUDJECT: hiceting hiinutes for meeting held at PSE&G on January ll,1993 for the status of the Salem Uriit 2 Pressurizer Safety / Relief Vt.lve Piping System biodificanon Gentlemen: Enclosed, please find the meeting minutes for the meeting held at PSE&G on hionday, January ll,1993. The primary purpose of the meeting was to discuss the schedule and current action items of the subject project based on restaning the project January 4,1993 for Salem Unit 2. If you have any questions, please contact me at (508) 370 3403. Very truly yours, , Guffdd~ 35 urJ ames F. Flaheny w .e 8%A- W'/ ' e_ Project hianager gy g% ' gf Vnclosure k h , 7 Ji%np da^.L le C ec: D. Bhavnani NfgE4fb C Vd k b' hiwdet J. orga (Falcon Power) 1 W / b O #
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4 TO: T. Taylor, Project Manager-Engineering and Plant Betterment FROM: C. P. Lashkari, System Engineer Salem Technical Department ,
SUBJECT:
PORV/PSV LOOP SEAL DRAINING DCP DATE: January 26,1993 I With reference to yourletter MEC 931025 dated January 21,1993 there are several items which should be included in the scope of DCP. ; 4
- 1. 1PS22 was found to be excessively leaking last outage. 2PS22 was changed out to Yarway FA 130.1PS22 also needs to be changed out to Yarway FA 130 from Rockwell FA 14.
- 2. Several Pressurizer Safety Valve Loop Seal drain valves (1/2 PR 8,9,10, 23) are Rockwell FA 14 valves. These valves have generic problem of stem ;
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detaching from the plug. These FA 14 valves need to be replaced by FA 130.
- 3. Several of the existing drain valves (1/2 PR 8,9,10, 23) have their reach rods disconnected outside the PZR enclosure. Either the reach rods should be made operable or removed from the field.
- 4. Removal of PZR Safety Valve Loop Sealinsulation box should be delayed until i Salem has had some experience with drained loop seal. Salem would like to retain capability to return to loop sealif PZR Safety Valves have serious leakage problems. A loop seal without insulation would be considered cold loop seal with much higher forcing function. Therefore, retaining the insulation 1 box would allow to return to hot loop seal with lower forcing function.
- 5. Drain line connection to PZR could be provided downstrecm of 1/2 PS22.
Existing PZR liquid space sample could be connected to thermowell in the , bottom of pressurizer. This will require modification of the thermowell. e CL/gb c: D. Bhavnani M 4.Danak Wiedemann , 1
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l - '; PSIEG f Pubhc Service E!ectnc no Gas C:~eam F O. Ban 236 Hancocns Bnage. New Jersev 08038 Salem Generaung Station i Tot John Wiedemann-Technical Engineer-NSSS FROMt Philip P.J. O'Donnell / jf" 2- - Operations Engineer ,_. ;
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SUBJECT PORV Loop Seal Removal ~~" --- i
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DATEt 6 February 1993- . I^ , In January 1993, we provided comment concerning the . design change proposal on the PORV Loop Seal removal. While' the conceptual design of routing these loop seal drains back to-the Pressurizer appears to be the proper method of j addressing the stress loads on the piping, there are a few issues to resolve before pursuing this solution. These issues ! are listed below 1.-How will safety valve leakage be minimized during rormal operation, since the original purpose of the , i loop seal was to minimize' leakage?
- 2. What steps have been taken to preclude the buildup of boron in this drain line? ,
- 3. The medification can only be performed in Mode 6 with the Paactor Head removed or Defueled. How is this :
i goir,q to fit into the outage schedule?
- 4. What will be the impact on RCS leakage with the ;
potential additional safety valve leakage to the PRT? l' If the identified leakage to the PRT increases to greater than 5.0 gpm, what will be the impact on the i. liquid radioactive waste volume and curie count?
- 5. Will-the project pick up the additional cost of processing any additional water waste from the safety
- and relief valves as a guarantee of their work?
Your: assistance in making sure that these issues are addressed' prior to acceptance of the design change will help . expedite the package through SORC. If you have any further ! questions concerning these issues, contact me as soon as. -- possible.- , (RA-i
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l REPORT ON ATTENDANCE AT OFF-PREMISES COURSES , ' SEMINARS AND FORMAL CONFERENCES f flTEE OF PROGRAM PRESENTED DV COST REGISTRATION #EE ,
*!UMARC AHAC Mt g . ( Alloy 600 CRDM) *00 & NUMARC & EPRI to eATES Of Af f ENDAt4CL LOCATIOrd OTHERgncig I,gg-fchtusty 19, 1993 washington. DC $278.07 ,
8 lEF DESCRIPTION OF PROGRAM i This was a working enesting in pr oper ation of a rnesting with the NRC on 3-3-g3. The NAC wants highly susceptable plants in the US to initiate inspections. It is based on ' e new approach whete actions are vnitiated by industry without teguistory actions by the NRC. While this approach may have its advantages because initiotive stays with ; industry, however, disadvantage could be unnecessary requirements which could be im-posed by the NRC without t h e i r m e n a l, <tme n t opproval. Also industry may have problems it implementing letge cost items wethout regulatory teaultements. NUMARC is trying to plo-cote the NRC so that it does not issue e generic letter or bulletin requiring all plants to do inspections. Colem et has to be concerned about future inspections. t
'PLE ASF ATTACH ANY PntNTED OCSCRIPtivE INFORM ATION IF AVAIL ABLD EVALUATION BV PARTICIPANT tinclude strong end ween points of the programi ;
WOG, EPRI made presentations on various segments or program. WOG inforreed the group i that one circumfetential crack has been reported by EDF. This is very significant since all stress emelysis so for in Europe indicate no circumferential cracking would occut, lastead all crackeng will te amiel. A safety evaluation submitted to the NRC wili need to be tevised, The f,RC me11 force inspections with this unfortunate finding. NUMARC & 400 eent all Category 3 plants to voluntarily inspect their teettor head ; during late 1993 and 1994 Setem #1 is e Category 3 plant and cannot inspect during r R11 (Fell '93) since planning for the outage is nest completion. Setem #1 con perfori anspections during 1R12 ef management opproves and the NRC sccepts acceptance criterie and the industry develops the technology adequately. Salem management should not agree to inspection if the above ley parameters are not complete. There is a good possibility ) thet by 1R12, the NRC would have opcided that a problem does not exist in US plants based on inspections et other plants dureng 1993 and 1994 Susceptability inden work by 400 does not consider key factors such as fabrication metnods, elding history & microstructure. Present indau is not a true indication of s eist of PASCC, DC Cook Unit 1 & 2 and a few other plants are prepared to proceed for enspections in 1993/1994., 400 timelene was revised to require NRC acceptance of industry ecceptance criterie before inspections would begin. The meetsng resolved thet the NRC would not be provided with names of any plants and their categories. Generally absolute value for renteng has been revised downwards from previous draft. , Tne CJO presented meterial on outside flows. WOG is nitiating a test program to determine crack growth reise bounding all plants. Additional funding is being i sought and the results would'not be sveilable until the end of 1994. A Ituttutet sub group has reviewed *00-work and finds it acceptable to present it to the NRC,
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X j FUSE NEXT P AGE IF NFCESS ARYt N AME .0F P ARilCAP ANT itTLE DEPARTMENT Chandre-leshkota System Engineer Technical j DEPARTMENT HEAD j SIGNATU fp OJgf ARTIC6 PANT
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e REPORT ON ATTENDANCE AT OFF-PREMISES COURSES l SEMINARS AND FORMAL CONFERENCES j TITLE OF PROGRAM PRESLNTED BY COST REGISTRATION TEE WC Meeting on CRDM C#acking NUMARC
.eAtES OF ATTENDANCE LOCATION OTHER tincluse Travet, Loog- !
ang Materials, etcJ t 3-2-93 & 3-3-93 Rockville, MD $249.02 B',lEF DESCRIPTION OF PROGRAM j There *es an f4RC i.seting on CRDM penetration cracking issue et 10:00 AM on March 3, 1993 in their offices at i White flint Road, Rockville, MD NUMARC also had a working ptemseting on Merch 2, 1993 at 1: 00 PM NUMARC and industry including 400, CEOG, B&WOG, EPRI & other vendors including most i utilities see petticipating in toss program, to keep initiative on this issue with t industty, insteed of NRC isteing new regulatory actions.
'PLE ASC ATTACH ANY ORINTFD DESCRIPTIVE INFORMATION IF AVAll ABLD EVAtVATION DY PARTICIPANT tinclude strong and weaa points of the prograrnt in 16e pre-meeting on 3-2-93 and NRC treeting on 3-3-g3. NUMARC and its AHAC actking grown had doctded to present 3 plants-.hich volunteered to carry out CROM pen, inspections in 1994. Salem ei c4fer to do sospection in 1R12 ISpring 951 was not .
considered eecessary for group presentetion. Therefore now Salem can plan this job to [ cetty out inspection on its o n outing ttR12 to protect its investment in reactor beau. This job is not being proposed cecause of any safety considerations per WOG. NRC Assoseate Director. Bill Allen. congtstulated industry for toking a proactive patosch on this issue. However. he was very forceful and indicated displessire on industry fall owners groups) not submstting a position in short term why industry can continue to operate until inspections are done, Present WOG safety evaluation , has been submitted to the NRC, ehach concludes that up to 6 years plants are in compliance with ASME code with expected teakege in CROM penetration above weld. WOG took a very narrow and indenfensible oppfosch. Most plants are on 18-24 months refuel-ing cycle. Once a plant finds e reek during refueling or shutdoen, it cannot start until leek 4 ropeared since this mould be considered a pressure coundary leekage under Tech. Spets. These comments mere made to AOG by the eriter cut it did not address Salem has defenseple opproach on this concern, in feet, our presentation .ovfd have addressed NRC concern any Salem is safe until inspections in 1995. NUMARC and other utilities did not want to present why Salem is safe in inspect'ng an 1995. Other plants mete reluctant Decause they do not Mave reactor head leak detection system and shroud doofs. Salem also performs a formel reactor head inspection on every shutdown which is quite frequent, it is bad from operational reliab:lity but is considered positive from reactor head inspection frecuency. Three plants planning inspections during 1994 ore DC Cook - Unit 2, Point Beach and Oconee. NRC *ents a CE plant to volunteer and CEAOG indicated providing the same to the NRC as soon as their~ susceptibility tenking *ork is complete. NRC . ants NUMARC to provide ehen the genetic position on safety evaluation of each NSSS o.ners .all ne filed with the NRC. It seems NoR is under pressure from internal and external forces to document its position on CRCM cracking issue at US plants. Since many plants operate
'USE NEXT P AGE IF NECESS ARYl NAME OF PARTICIPANT OEPARTMENT C . t + L tO H VM V lTITLESYST2M GMGil N E c~R sAceM 2 CH sicNATuREf e,ARTICIPA T . OLPARTMEr HEAD BW c JG i M [. /V/MRoM E
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EVALUATION BY PAMTICIPANT (CONTINUED) continuously for 18-24 month cycle and do not htve any leek detection on teactor head, it wlll be difflggl1 for o*nott gfoWP to take e genetic position and satisfy t he tJRC c once r n. f4AC wa n t s a tre e t i ng in June 93 to review EPRI work on tJDE st EPRI tJOE c ent e r . Setem Technical Department will continue to monitor and work with WOO on tiils issue. , tJCRPs f or 1996 inspections et Salem see being repeited and will be submitted to E&PB tesovice ellocation group soon. I I l l l v ., , , . i . 3
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.NESTitCHOUSE PEOPTJETAT,Y CLA03 2 R/V CLOSURE HEAD PENETRATION CRACKING l i'. /
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.. 1 CRDM PENETRATION CRACKING ISSUE -
SALEM ACTION PLAN !
. t l _1993-2R7 , l'
- PRE-INSPECTION WALKDOWN j
- ! - SCOPE MODIFICATIONS OF RX HEAD STAND ,
t I 1993-1R11
-RECORD ON VIDEO & PHOTGRAPH RX HEAD !
PENETRATIONS
-SALEM PROCEDURE & CHECKLIST
-PERFORM-VISUAL INSPECTION
-MEASURE G AP FROM TOP ;
-ME ASURE OVALITY l
-MODIFICATIONS TO RX HEAD STAND
-T-COLD CONVERSION IN RX VE3SEL ,
I 1993-N ON-O U TAG E
-FUNDING WOG PROGRAMS
-FUNDING EPRI PROGRAMS
-CONTINUE TO MONITOR NRC ACTIVITIES
-CONTINUE PARTICIPATION IN WOG/NUMARK/
EPRl/NDE DENTER
-PLANNING FOR 1R12 INSPECTION '
-CONTINGENCY PLANS -
-REVIEW & PLAN MITIG ATION TECHNIQUE
-T-COLD CONVERSION !
-PENETRATION ID PEENING ,
-SLEEVING >
L -WELD THE GAP
--RCS ZINC ADDITION
-NEW RX VESSEL HEAD-PARTICIPATE WITH OTHER UTILITIES
O PSEG PuDhc $erwCe Electnc and Gas Company P O Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department to: P.J. O'uonnell FMS 93 H2 Operanon: Managrr . Ssjem U ,g} Frvm: T.N. Tayler PrWeet Manager . Data: March $,1993
Subject:
PORY and SV L>op Seal Removal No.1 & 2 Unita Ssiem N.O.S. This memo is in respcase to your com:nents outhned in a letter to John Wisdemann (PSE&G) on February 6,1993. The following responses should help answer the quotions and comments tr.iaed in this letter.
- 1. *How will safety valve leakaas be min: mind dunng normal operation, sinn the ornsina!
purpose of the loop ual u as to msn.'rnlu leakaget' ABB Impell recommended draining cf the Pressurtar Safety Valve loop mais based upon uveriences at other nuclear p' ants with sicitar piping eeneerts and staular safety valv*s. Salem's Prueuriser Safety Valm were manufactured by Cresby Valve and Gege Ocupany. Ornby was contacted to dinuas drmning of the protacave water loop seals and any recommended modiflesmans to the safety valves. The erfsting safety valve design has required a water loop seal to protect the saatics of the valvea from steam cutting and non condenemle gases. Crosby has sabeequently douleped a design of the meernals (P'ad. Dine design)(or their safety valves that do not require a water loop seal. N tasse modificacion twolves replacement of the nosale and disc anurt and macnining of the exietics disc ring. The above modifications will take place at Wyle labs under the direction of a Craby npresentative. OVyle Labs has previously nAirtished and tasted the SVs for Salem I'nita 3 and 2 durug past eutages.) Following the modecadons, the SVs will be tuted fer !caksge and set point to the same nquirements o(the past testing. The SVs wi]l function as in:nally dulgned, e.= cept without the requirement for a loop seal. Copu Vulcan, the vendor of the PORVs, was also costaeted to addnse the nmov.sl of the lxp seals. Copes Vulcan roepastded with reecmroendations ta enhance the maling af the PORVs. Copes Valcan recommended 420 stainlus steelint4 mal trim set finished and tested in a 'siave body". N PORVs are butt welded into the system piping and thus, cannot be nir.oved. Copes.Valean will test each new internal trim wt to unt ar esceed the Clus V requireneuu which are the highest quality of leakage requintnants fer metal seating control va5v es. A Copu Vulcan repruantanve will oversee the modificaticas, inspection and tascing c' '.he PORVs. N PORVs were enginal}y designed for steam serdce and do not n$ tire a watte loop seal for opersuon.
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f Mensorandum To: P.J. O'Dov. ell ' Froso: Tal Taylor Date: March 5,1993 Page 2
- 2. 'What stepe have been taken to preclude the busldup of boron in this drain line!"
There will be no build up of boron in the new drain line. First, there is no ersporation taking place in the drain hne to concentrate the baron. Second, during operation, there will exist a continuous flow of" clean' eendensed water in the drain line, essentially continuously purgmg it of any boren which entered it dursar startuvrefueling. Thirdly, the drain line wiD be insulated. Also for comparison, the F8' pressunser liquid space ssapling tubing, which is not maulated or heat traced in containment, has not had problems due to boron solldl8eation. 3. The modifitation con only be performed sn Mode 6 with the Reactor Head removed or Defueled. How is this going tollt insa the outage schedule!" PSE&O has indicated to ABB 1mpell that the modification has been scheduled for this ontage. 4.
'% hat wil. he the impact on RCS leosage uith the potential additional safety valve leakage to the PRTT
- the sdentt,Med laakage to the PRT increasee to greater than 8.0 aprn. what will be the impact an the liquid radioactive watte volume and curie countt' There is no inerasse in potential leakage orthe SVs. The SVs will be tuted at Wyle tabs to the same leakage requinmenta as they currently have. The new Flead Dine internals are designed for a suam environment.
'L
*Will the profcct pick up the addissonal cost ofprocessing any additional water uante fem the safety and erlwf valves as a guarantee of their workf*
This is a regalatory driven O/M duign change. 'Ito best available appropriate technology is being utilised by the project team. If the not result of the modilicatlan is unacceptable,it will be rectified to the extent that is in the best interest of PSE40 by the prqjoet. Operation erpenne will remain to the account or the station. ec: J.A. Ranalli M. Marroni J. Wiedermann
- h C.14ahkart C. hm J.N. I.aech
- 0. Lhrer R. Ketchum D. Bhavnani Scandards Records Coordinator
Salem Generating Station Technical Department Engineering Memo MEMO No: 93-049 REV 1 DATE: APRIL 6,1993 To: R. CORNMAN FROM: C P LASHRARI
SUBJECT:
PCS DFAINDOWN WITHOUT REFUELING RVLIS INDICATION For RCS drain down +1 ring 2R7 after off-loading fuel from the reactor,it is u.derstood that the REFUELING RVLIS indication in the control room is not available.With the core completely otf-loaded to the spent fuel pool, core cooling concerns et NRC Generic Letter 38-17 and INPO SOERs do not apply.With appropriate measures in place to prevent loss or the RHR puroc ,enile in mid-loop,an unnecessary failure of catety equipment (RHR PUMPS) can be avoided.The following precaut:anc snould be taken: 1.The prescurl:er ;old level channel chould be compared to the intermediate (23RC15) leg tygon tubing.If discrepancy is greater than , anches,the tygon tubing ccale reference level chould be verified.Although an initial call to System Engineer is appropriate, the E&PB Civil group support for chooting a level from a reference level outside the biochield area may be necessary. Outage management should alett E&PB Single Point Contact for off-hour coverage. 2.The pressuricor cold level channel vill provide coverage to approx. 104' elevation.from 10 elevation to 99.1' elevation,;nter..ediate leg tygon Jbing local leVol indication anall te recorded evet, hour in the draindown procedure.from elevation 99.1'to 97.5', redundant mid-loop level indication in the control room are still available.Further draini.ng if required, may be accomplished by gravity drain from 23RC14 with RHR pumps tagged. Note that the mid-loop low res level alarm at 97.5' elevation may be utilized in addition to the redundant mid-loop level indication availabic in control room.Mid-loop level indication shall be calibreted !AW applicable procedures.
+ J APPROVALS: ;m A f ,*
INITIATOR--- %/ y >-------
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REVIEUEP 'A - DATE---- M-- ------- t - 4 4 O DEPARTMENTHdk -- - U D TE---
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- 3. Admir.isrativa actions such as a Supervisory Letter, Management awareness training session and fornal training of licensed operators is not necessary since the reactor is nefueled.However, a shift briefing on the revised procedure is recommended. containment coordinators should review Attachment 2 of this procedure with the USS overy shift to prevent unauthorized activities which could result in loss of RCS inventory or loss of RHR pumps 4.When Tygon tubing is the only available RCS indication (from 104' to 99.1' elevation ) local level monitoring REV1 inside the bioshield by an operator overy hour, with direct communication to contral room, shall be provided.
If communication with the control room is lost or malfunctions, the drain down shall be immediately suspended. 5.The tygon tubing installation shall be walked down by the System Engineer and Nuclear Shift Supervisor, 11 Db discrqpancy areu. fir than 9" are recorded. 6.To ensure RCS Jraindown activities are carried out IAW with wr itten instructions,a TEMPORARY procedure with appropriate instructions is recommended. 7.A review of the NRC Ceneric Letter 88-17 and PSE&G responses has been conducted.Dasically, GL 88-17 requirements apply to reduced inventory operations with fuel in the reactor.The GL 88-17 requires two independent level indications in the control room while in reduced inventory operations.The NRC does not approca of Tygon tubing as level indication and states in the GL 88-17 "We also note that ordinary plastic tubing doen not meet our concept of reliable instrumentation, and its use may not be accepted as a component in instrumentation system". Since we are defueled, above concern does not apply. In this evolution, loss of shutdown cooling is not a concern, but protection of the RHR equipment is essential. C:R GALLAHER,Jr., R ANTONOW , J WIEDEMANN , R LEMBERGER REV1
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. MAY 2 61993 NLR-N93085 Westinghouse Eleutric Corporation Box 355 Pittsburgh, PA 15230-0355 nTTENTION: tir. A. Sicari ,
Dear Mr. Sicari:
REACTOR COOLANT PUMP SEAL INJECTION FLOW Salem is evaluating the need for a Technical Specification (TS)
~
change pertaining to Controlled Leakage. Currently Controlled Leakage is dafined as that seal water flow from the reactor coolant pump seals and is limited to 40 gpm with Reactor Coolant System (RCS) pressure at 2230 2 20 psig per LCO 3.4.6.2 (Unit 1) and 3.4.7.2 (Unit 2). Salem Station measures number one seal leakof f to -omply with the '?S. The bases for the specification states that it is to ensure that in the event of a LOCA, the safety injection flow will not be less than that ascumed in the Accident Analyses. As you are aware, seal return is an accurate indicator of the safety injection inventory that would not enter the RCS on a LOCA; however, it is net indicative of the total flow that would be delivered to the seals. For this reason PSE&G is requesting the following informetion:
- 1. What is the specific basus for the 40 gpm criteria for controlled Leakage, (e.g., charging pump runout, flow not delivered to the core via Cold Leg Ir.jection, flow diverted from the RCS, etc.)?
- 2. Does Salem meet the assumptions of the Accident Analyses by measuring seal leakoff flow?
- 3. If number one seal leakoff flow is the correct parameter to be measured, what limiting value should be used; the 20 gpm given in the vendor's manual or the 40 gpm e currently used?
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-Mr. Sicari 2 NLR N93085 t
- 4. If Salem does not meet the assumptions of the Accident Analyses, what affacts does this have on the analyses including assescaent of available margins?
Your response to the above quactions is necessary to ensure Salem ; Stations are consistent with the Accident Analyses. Please ' contact Steve Mannon (609-339-1129) if you have any questions on the above. , Sincerely, *
/ ,
F. X. Thomson, Jr. Manager - Licensing & Regulation SRM/dlb' Ct_ R. T. Brown (N21)
- a. na wi 2- inn @ab H. G. Berrick (N51)
C. P. Lashkari (SO2) C. Timm (S02) J. A. Rowey (N$1) M. R. Danak (N51) S. R. Mannon (N21) '. A. N. Baird (S02)
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O PSEG l Pubhc Service E ectric ana Gas Cy ca y : C 0:n 236 -a,coc s Sr c;e. ' Jew sersey 0B038 Salem Generating Station f I To John Samson i Principal Statf-Administrator - Proc 4 Mati cont : FROMt Jim Webster .
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l Project Mana er aliin Unit'#1 Outages
SUBJECT:
RCP INSULATION ! DATE: June 2, 1993 : OUR RErt 93-028.JTW' ! In response to your lotter dated 5/25/93, I have reviewed the need f j for this insulation with System Engineering, and there arc no current plans for this material. The concern for which this material was purchased was analyzed by Engineering, and their calculations show it is no longer required. I recommend you contact other Westinghouse Nuclear utilities who might-have a need for this material, and thereby give us a better prica than we might get otherwise. JTWicms l C R. Antonow R. Cowles ' R. Graybill
- c. Lashkari ,.
N. Leech Shadlock [M. Wiedemann File /WP51/OMU1 i 1
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- The Energy People: [
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NUCLEAR DEPARTMENT WORK ORDER
- TASK RETEST ACTIVITY PL W/O: 930304071 ACT: 02
_ ,_. ORIGINAL , UNIT ASSIGNED PLANNER SPTY RLTD RETEST MODE RESP DEPT /GRP S1 HUNT 2098 SR SOD / OPS COMP /PEO NO: ICH216-AO SYSTEM CH 1 CHL WTH EMERG CONT AIR COMP SPLY LINE GLB VLV AIR OPR i LOCATION: - ' W/O DESC ICH216/EMER CNTRL AIR CMPRSR CHL WTR SUPPLY VALVE-EXCEEDED PREVIOUS STROKE TIME BY GREATER THAN 2dt. PLEASE TROUBLESHOOT AND CORRECT PROBLEM AS REQUIRED. PMT STROKE TIME TEST. WORK PERFOrJ4ANCE DEVIATIONS: RECOMMENDED EXERCISE TEST, STROKE TIME TEST, FAIL SAFE ACTUATOR TEST RETEST RETEST PERFORNED/RESULTS/ DEVIATIONS: (JPS UFY REASON POR NOT PERFORMING REr'OMMENDED RETEST) gjg us, igu ) S t. o P - S T c H - x o "^' UNSAT g.g,g M7 N h 3 5'c
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N/A (ATTACH COMPLETED RETEST PROCEDURE TO THIS ACTIVITY) MOTE
.IQUIP USED _
CLOSEOUT - (SGS - SIGNATURES REQUIRED FOR SAFETY RELATED WORK ONLY) q SAFETY RELATED: NON SAFETY RELATED: , PEASON PERFORMING RETEST: BADGE NO. DATE
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.IVITY CLOSEOUT SHEET RT NO. TASK 000000 PL W/O: 930304071 ACT: 01 PAGE 2 OF 2 M+TE v)A EQOITMENT r USED/NEEDED _
DEFICIENCY REPORTS INITI ATED A//A COMPONENT SERIAL i d,fa. SERIAL i UPDATE / / DESCRIPTION OF WORK PERPORMED: / //8[1 mfd khLuld SM.-% Auo GbS6RVtfD 0PS fd$ $ $ A1 A NSWIC 't / if Tl?Sf $ UA j< ' Sf*/ L fA t iK D TI5K . nit *VI t a m' K n A J SA~c w D2 -f*K M* G AJ .2 3 Ele .
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Rn. / PAII. ORE CADSE: N 6WU PMT PERPORMED: -
1 ___________________..------WORKREQUESTDESCRIPTION-------------------h2 COMMAND INPUT ===> init jddE EATD moi >E , PAGE 1 OF 2 i WORK REQ 110 : 930615112 AR STATUS-DATE: /06/15/93 DATE INITIATED: 06/15/93 LAST UPDATE : 06/[/93 CPL INIT DE'T r : STD_ SE_ INITIATOR : ;SHKARI , RESP DEPT : SSP _ PRIORITY A PLANNED START: 06/15/93 AR TYPE : CM _ C.R. INST /IND : . MUC: SYSTEM: CS UNIT-CMPNT ID : 51 11CS49 FG NO: F/ CMP DESCR: UNIT 1,11CS HDR LOCATION : INSIDE CONT ANNULUS : EMIS HUNG? : N ACTION STATEMENT DATE-TIME : 3.v.2.1 06/15/93 08:52 EMIS TAG: REF. WO/ACT-
SUMMARY
- DRAIN & FLUSH 11 CS HDR !
A DESC: DRAIN & FLUSH 11 CS IIDR, ENTER TECH SPEC, TAG 11CS PUMP BRKR_,,_ . CONNECT A 3/4" DRAIN HOSE TO 11CS67,OPEN 11CS 67,OPEN VENT 11CS46 WHEN THE ?lCS IIDR IS COMPLETELY DRAINED,CLOSE 11CS67 AND 11CS46, OPEN 11CS2_ _ FILLING !!DR COMPLETELY.HDR LEVEL SHOULD DE APPROXIMATELY SAME AS RWST LEVEL _ AND CAN BE VERIFIED BY Tile TYGON TUBING ON 11CS67 AND RUNNING IT ABOVE 130' _ ELEVATION. _ TAKE ANOTHER SAMPLE AND HAVE IT ANALYZED BY CHEMISTRY.I* NOT SAT,REPERFORM
~ DRAIN AND FILL.IF SAT,THEN REMOVE DRAIN HOSE AND RETURN SYSTEM TO NORMAL
_ LINEUP. EXIT TECH SPEC. ; MESSAGE: PF4 TO COMMIT; 'CAN' TO CANCEL; 'NEXT' FOR NEXT SCREEN
- B MY JOB LU #28 i
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-SALEM; GENERATING STATION TECMNICAL DEPARTMENT f i
ENGINEERING MEMO j lo [--------------- DATE 6-3 0-9 3 MEMO NUMBER - $ 3 ------ ; FROM: C P Lashkari, Systen Engineer j i To: F Kaminski, IST C'oordinator ,
SUBJECT:
-4.0.SV Testing of 1CH216-A0 REFERENCE W 0 930304071 IAW S1.0P-ST.CH-0003-(Q) ,
Several times in the past, valve ICH 216-A0 has exhibited , greater than 25? increase in stroke time as tested per above referenced procedure.This has resultat in placing this valve in alert range and increase in test filoquency from 92 days to 46 days.Since this is a 20 second valve and there-As no ', added value in increased test frequency,a new baseline can be established using test data dated 06-04-93 (2.31 seconds) which was performed as a result of maintenance on the valve. j l If any additional information is required,please contact me i i C R Gallaher P O'Donnell J Weidemann ' t i APPROVALS-(signature an date) ; INITIATOR - p, d% 7/' - REVIEWGR ; 1 6( 14 - DEPARTMENT HEAD' # ll/ h W 7443 >\ \ (~ 4,\ '
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Scic: GCn rcting StOtiCn TCchnic31 D p3rtC:nt ENGINEERING HEMO MEMO NO: 93- l oo DATE: June 30, 1993 TO: Frank Kaminski, IST Coordinator - Salem Technical FROM: Charlie Lashkari, System Engineer - Salem Technical
SUBJECT:
4.0.5 P-CS.11
REFERENCE:
TEST DATED 6/2/93 The purpose of this memo is t) revise the existing baseline dated 6/26/91 for 11 CS pump, based on rev' s 01 the test data during last 2 years. Review of the trended test data clearly indicates that the scatter of test data is too close to the high alert line (1.02x). This basically is a problem with the Code (Table IWP-3100-2 of ASME Section X1). The Code allows lower side acceptable range of 71 but only 21 for higher side. The acceptable range is (0.93-1.02)x D.P.(reference). The Code reflects the basic presumption that pumps do degrade but do not improve over time. However, the Code did not tako into account the readability error of the operator, specifically when the pressure gauge needle may be fluctuating bi approximately 5 psig. This fluctuation and operato1.- readability errors may be responsible for putting 11CS pump in alert range in the past. - To correct this situation, it is recommended to change existing baseline table with new baseline prepared from test dated 6/2/93. This would minimize higher value alert while also avoiding thu lower range alert as well. Please centact me if any additional information is required. CL/gb cc: R. Gallaher P. O'Donnell J. Wiedemann INITIATO 14 EnA.f -l- Y 3 REVIEWER d DEP m nENT <, mdfN
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___ _ . _ . ___m. .. . - - . _ _ . . . _ . _ _._- . . . - _ _ . _ l l i TO E J Gailbraith-Chemistry Engineer + C P Lashkari- System Engineer r* FROMt , SUBJECT ICS49 DEAD LEG SAMPLE CHLORIDES DATE: July 1,19*3 I i With reference to your letter dated Juno 15,1993 and
- Chemistry Specification Deviation Notice No. CH93-258 -
~
please initiate - procedure change to initially flush prior to taking sample for chlorides from CS header. i It is System Engineer's opinion that' elevated chlorides ; in samples taken from 1CS49 during June 93 were due to l contamination from chlorides in containment air.It is possible for containment air to condense insida the threaded cap.There was no contamination of containment ' spray fluid.inside containment spray header.In addition, RWST inventory was well within specification during this j period. ; C:J'W Wiedemann , i i b
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O y. TO: K. Piko Manager - Salem Technical (Acting) FROM: C. Lashkari System Engincor Salom Technical SUCJECT: CRDM PENETRATION CRACKING ISSUE DATE: July 7.1993 This momo presents curront status of CRDM Ponotration Cracking issue. O&M Funding is required for planning and implomonting a multi year and a multi discipline action plan that encompasses reactor head inspection and repair technology for both Units. My memo dated March 11.1993 briofod the status of tho issue as presented to the NRC on Maich 3,1993 by NUMARK ADHOC Advisory Committoo on Alloy 600 (AHAC 600) group. Notes from the NRC mooting on July 15,1993 are being circulated separately. During this mooting, WOG presented to the NRC acceptanco critoria for axial and circumferential cracking in CRDM tubos. Acceptanco of this ':ritoria is ossential for inspections to begin and is expected by the and of 1993. Listou below are different olomonts of our acticn plan and proposed responsibility assignments.
- 1. Salign 1 Insnection On March 3,1993, NUMARK modo a presentation to the NRC when 3 U.S. plants (D.C. Cook-2. Point Beach 1 and Oconoo) committed to inspections during 1994. WOG had classified all oporating Westinghouso plants based on operating timo and RCS temperaturo. A Relative Susceptibility index (RSI) was calculated which comparos U.S. plants to Ringhals 2. The susceptibility index for Salem 1 indicates a potential for Alloy 600 cracking based on comparison to Ringhals 2 operating history. Salem 2 is relatively Ic.w on RSI and therefore will need inspection based on Unit 1 results. Salom 1 was found to be in the samo category ar. DC Cook 2, Point Beach 1 and Trojan. Although Salem 1 has not made a specific commitment to the NRC, Salem 1 is planning inspection during 1R12 as a planned and prudent action. The inspection is anticipated to be performed through the ISI Group. The Toch. Dept, and E&PB will be responsible for selecting the NDE method and vendor.
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- 2. NDE Techniques f
EPRI is developing various NDE' techniques at the NDE conter. Thc mockups will be complete by the end of the year. Demonstration and qualification of operators / vendors will be carried out to meet the - - inspection echedule in 1994. A vandor will need to be selected to j support 1R12 inspection requirements. RESPONSIBill1Y Salem ISI/ Salem Technical /E&PB 1
- 3. REPAIRS /VSE AS IS A NSSS vendor from West? .1 house, CE or B&W needs to be selected for carrys.'q out a USE AS-IS safety evaluation or repairs to CRDM tubes, if required. All tooling and techniques need to be evaluated on EPRl/NSSS mockup as a contingency. The piesent acceptance criteria does include a use as is criteria that will allow repair at a later date. ;
RESPONSIBILITY Salem Technical
- 4. YlQEQ RECORD OF RX HEAD _ UNDERSIDE A video recording of Rx Head underside is proposed which will be useful as planning tool prior to 1R12. This recording can be n.ade by Brooks Associates for approximately $40,000 during 1R11 with approximate dose of 100 MR. The job can be completed in one shift and requires approximately 1 more shift of prep ~ation time inside containment. ,
it should be noted that this video recording has no potential to detect any defect which could delay the outage. Out Of Budget funding is needed so that this activity can be carried out during 1R11. Work Order
#930702183 has been initiated.
S. MITIGATION TECHNIQUES The following 3 options are presently _WOG and EPRI recommendations. _ Additional- options may develop as technology and industry data increases . .The final selection at Salem will be dependant on life extension efforts, outage impact, technical feasibility and economic impact, No action is anticipated until after 1R11. ;
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'A, Rx Head T Cold Conversion-WOG has recommended modifications internal to the reactor which c- could increase leakage flow. This increased flow to the Rx Head is_ expected to reduce the Rx Head temperature, thus decreasing the potential for CRDM cracking.
t B. Zinc Addition WOG and EPRI have reco,:. mended Zine addition to the RCS which will reduce the potential to cracking in the Alloy 600 CRDM tubes - while generally lowering the radiation fields around RCS piping and components. This would reduce the outage dose while reducing
- the potential to cracking in Alloy 600 CRDM tubes. This
, modification will not be available until 1995 and therefore, will be implemented during 1R13 and 2R9, if considered effective and safe. Since WOG general meeting did not approve this item, a sub group of approximately 15 plants has been foi. >3d to continue to fund the research. PSE&G is participating in this effort. C. Shot Peenina CRDM Tubes - WOG and EPRI have recommended shot poening the CRDM tubes. This project is similar to shot peening the S.G. alloy 600 tubes which is being carried out during 1R11. Shot poening does not help with existing crack initiation. However, it will prevent further crack growth by removing residual stresses at the crack tip.
- 6. RECURRING PM TASK After initialinspections are complets on both Salem Jnits, RTs will have to be created to carry out this inspection approximately every 10 years along with- 10 years ISI of RV. This will depend upon WOG recommendations.
SUMMARY
/ CONCLUSIONS CRDM Alicy 600 defect acceptance criteria is expected to be approved by the NRC by the end of 1993.
Inspection will be required on U/1 based or$ similer history to Rinhals-2. Salem U/2 may be required later based on U!1 results.
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An inspection technique will need to be selected and scheduled for 1R12. Itis expected to involve robotics to minimize doso. Repair techriology is under development. Proventivo options are under develop nont. O&M budget allocations for insg9etioni anatyeis and repair will be needed for support of 1R12. CL/gb C: General Manager - Salem Operations Operations Manager - Salem Maintenanco Manager - Salem Radiation Protection / Chemistry Manager Salem Manager - Nuclear Mechanical Engineering Manager Radiation Protection / Chemistry Services Manager - Reliability & Assessment l l l
To: K. Mathur, Project Engineer Revitalization Project, E&PB From: C'P Lashkari, System Engineer dl Salem Technical Department
Subject:
Review of DCP 1EC-3249 Haplac4 ment of PS 1&3-Internals DATE: July 22,1993 A review of DCP 1EC-3249 indicates a potential nuclear safety concern regarding safety Injection during the special test per DCP Section 10.3.4.1 used to determine pressurizer spray valve capacity.This capacity should be determined either by calculation or test by the vendor.The test does not predict the valve flow capacity nor does it provide adequate acceptance criteria. My concerns are:
- 1. Increased-probability of Safety Injection during the test as a result.of rapid RCS cooldown.
- 2. The test does not determine the capacity of the valve quantitatively.
- 3. The special test in the DCP is a test or experiment not described in the SAR. However, safety evaluation per 10 CFR 50.59 does not address the test.If this special test needs to be run, then safety evaluation should be-revised.
In addition, the package is deficient with respect to CDs on changes to the Maintenance Department Procedure SC.MD-CM.RC-0002(Q) and CBD. Marked-up changes should be provided in the DCP. C: K Pike J Wiedemann / k
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- I TO: !H. Trenka, Project Managerf Design Revitalizatien Project, E&PB' 1 1 FROMi J. Wiedemann, Technical Engineer 2 Salem Technical Department - i
SUBJECT:
REVIEW OF DCP 1EC 3249L REPLACEMENT OF PS 1&3 INTERNALS . DATE: July 23,1993' : The following concerns have been identified during the DCP review. These issues will - require a resolution before station acceptance and System Engineer's signature.
~'
.1. Technical Specification 3.4.9.2.b cooldown rate will apply during functional testing.- The maximum anticipated cooldown during the valve testing must be below the Technical Specification maximum of 200'/hr.
- 2. PRZ Surge Line transients will be mi.nimized during testing to minimize nozzle loadings.
- 3. The'50.59 must address the catential for an S.I. For the test, 50.59 Safety- 1 Evaluation section 2C shouid be answered yes.
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TO: W Grau, Station Licensing Engineer Nuclear Licensing & Regulation Dept. 2 FROMt _ C P Lashkari, System Engineer , qgh42[_ , Salem' Technical Department
SUBJECT:
Salem' Tech' Spec Change DATE: August 5,1993 , With respect to Tech spec change for Containment Air Temperature,-following is my comment. S R 3.6.5.1 . The containment average air temperature shall be the arithmetical average of the 5 primary area temperaturesons provided in the basis section and shall be determined at least once per 24 hours. If the primary thermocouple is
. not available, the secondary thermocouple location shall be used.
THERMOCOUPLE LOCATIONS (IN BASIS SECTION) , PRIMARY SECONDARY
- a. Elevation 84'-North Elevation 84'-South
- b. Elevation 78'-East Elevation 84'-East
- c. Elevation 106'-North Elevation 106'-Sou*hwest
- d. Elevation 106'-East Elevation 121'-West
-e. Elevation 136'-West Elevation 136'-South
- Ct J K Wiedemann
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i RX HEAD VENT LEAKAGE l i SALEM UNIT 2 , LEAKAGE IN GPM ! 2.5 - f 2 ! t 1 1.5 - l 1 l < 0.5 -- - O l l ! -- l i i I (l l ! i l l l l 1 l+ 1 -l- 1-1 1 -l -l 666666666666666666666666666666 4
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- PLohc Service Electric 'anc _ Gas Company P O Scr 236 Hancotas Br.cGe, h Jersey 08038 f
- Nuclear Department SCI-93-0682 ,
-.TO: J. Ba!. ley _
f Nuclear Engineering Sciences Manager, FROM: J. Perrin Technical consultant - Materials
SUBJECT:
WOG MATERIALS SUBCOMMITTEE TRIP REPORT-DATE: August 17,-1993 I attended the-August /10-11, 1993 WOG Materials Subcommittee . Meeting held in Pittsburgh, PA. The meeting was chaired by Sid Butler of Southern' Nuclear. Highlights of the meeting are listed below.. REACTOR VESSEL CLOSURE HEAD-PENETRATION ALLOY 600 PWSCC PROJECT. This WOG project is being conducted as a result of Alloy 600 penetration leakage in an EdF PWR, Inspections indicated the presence of axial cracks on.the inside diameter of the .illoy 600 tube. Cracks extended above and below the' head attachment veld. The EdF leakage nas been attributed to primary water stress corrosion cracking (PWSCC). Inspections at other EdF plants
! revealed the presence of_ cracking in additional penetrations. (
Phase 1 of the WOG project confirmed PWSCC as the cause of the
~
cracking, and concluded that Westinghouse vessels may be susceptible to head-penetration PWSCC. _As part of Phase-II, a 10CFR50.59 safety evaluation for operation with_ cracked penetrations was completed. Also, a technical basis was developed forLa susceptibility screening methodology. The 4 current phase of the' project includes the-assessment of an outside diameter crack.that has been found-in an EdF reactor, the continued support for developing:an industry strategy, the
- performance cf additional microstructural assessments, and the continuation of crack. growth tests. The preliminary results of the in-progress crack-growth tests is that_the growth rate is similar to the rates found:in earlier testing-for-Alloy 600-steam generatorLtubing. The_NRC is reviewing the technical basAs of
-flaw = acceptance criteria ~for flaws _that may be found in U.S.
plants.- In addition, a proposed WOG project authorization for
-developing a-cost-effective management strategy _ relative to potential repairs and mitigation actions.is under review.
/
TEMh People-n ne v w : n
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i [ J. Bailoy 2 08/17/93 PROJECT AUTHORIZATION FOR NEW FATIGUE TEST DATA. A project authorization for a program to develop new fatigue test data was considered. The S-N fatigue curves for carbon and low alloy steels defined in ASME Section III were based on laboratory tests performed in air many years ago, while the_-_ curves for austenitic steels-include more data in the area of high cycle fatigue. The da/dN-fatigue crack growth curves in Section XI were based on data from more current work, but today there are improved analytical methods and a better appreciation of the effects of variables. In some cases, da/dN data exceed the Section XI curves. As a result, the NRC has been considering the appropriatoness of the fatigue curves in Sections-III and XI. Another concern is whether or not the ASME approaches described in the respective sections reflect the environmental effects on fatigun behavior. After extensive discussion, a project authorization for a reduced program in the amounc of $25,100 was approved. LOW UPPER SHELF ENERGY LEVELS. A WOG project is underway to assess upper shelf toughness for prussure vessels in Westinghouse plants. A draft report, WCAP 13587 " Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors", is in the process of being completed. SURVEILLANCE CAPSULE FLUENCE PROJECT. Since the early 1970s, a large number of Westinghouse pressure vessel surveillance capsules have been examined, with fluence values determined for each capsule. A WOG project is being conducted to reevaluate fluence values that have been reported during capsule examinations. This is being done using improved techniques for determining fluences that are now available. It is important that fluences be known as accurately as possible since the NRC uses this information as part of the input for establishing trend curves incorporated into regulatory guides. It is expected that this project will be completed by the end of 1994. REACTOR VESSEL ANNEALING. A project authorization for an annealing demonstration project was presented. The purpose for annealing a reactor pressure vessel is to remove neutron embrittlement that occuru during plant operation. Although many Russian reactor vessels have been annealed, no U.S. vessels have yet been annealed. An economic analysis has been performed to evaluate the cost benefit of vessel annealing. Westinghouse believes that development of an annealing option would allow some PWRs to save significant money in the course of operating until the end of the 40-year license period. In addition, annealing may be necessary for some plants in order to go beyond 40 years into a license' renewal period. A proposed project authorization is expected to be faxed to subcommittes' members later this month. Please contact me at X1655 if you have any questions or would liku additional information. JP:bjt SCI 93307
s j J. B311cy 3 08/17/93 C M. Morroni J. Wiedemann ' C. Lashkari J. Ranalli D. Longo J. Wilson H. Berrick Standards Records Coordinator SCI 93307
J TO: .R Morgan ,Impell Corporation
- FRON: C P Lashkari, Salem Technical Department-DATE:~ September 2,1993
SUBJECT:
Part B' Closeout of DCP-2EC-3190 DCP was presented to System Engineer on the last day of-closure.This DCP as-presented can not be closed out as detailed belows. , 1.New components 2PS58,2PR67,2PR68,2PR66,2PR64,2PR65, 2PR69,2PR70 have open work orders in MMIS.to cover t inservice exam. by SQA 2.DCP. Team / Vendor Crosby must provide recommendation on valve refurbishment / resetting frequency for PZR Safety Valve which has undergone-flexi-disk modification.
~
3.CDL in the package is not signed-ofr. Vendor Manual PSDP 10G418 (PZR Safety-Valve) And Vendor DWG 315269 (PORVs). were checked from TDR.These documents have not been revised by E&PB.
- 4. Revised BOM No. 996Q for PZR Safety Valve is not in MMIS.Therefore, Station would have difficulty in ordering the. parts for next job.
- 5. Salem Maintenance Dept Proc. TSC.MD-GP.ZZ-0001(Q) has not been revised incorporating DCP 2EC-3190.
4 Please resolve above items before presenting the package for signatu'se.If you need any assistance,please call me at X-2754. C:T' Taylor
,,/'J Wiedemann D
Salem Generating Station Technical Department Engineering Memo MEMO _NO: 93-145 DATE: October 14,1993 . t TO :- P O'Donnell, Operating Engineer Salem Operations Department FROM: C P Lashkari, System Engineer Salem Technical Department
SUBJECT:
Salem 1 Mode 6 Pressure Relief at Greater Than 0.3 PSIG Procedure S1'.OP-SO.CBV-0002(Q) requires System Engineering evaluation at containment pressures between +1.0 PSIG and +3.0 PSIG. This memorandum provides guidance on this subject. Please install a hand-sender un IVC 5. While IVC 6 is fully open, crack open-1VCS to establish airflow from containment. This will slowly reduce the containment pressure yet ensure the pressure relief system components are not damaged. When containment pressure is reduced to 0.3 PSIG, 1VCS may be fully opened. Contact system engineering when the pressure relief is complete, so that the system visual inspection may be completed in accordance with procedure step number 5.1.2. N f MD DATE /0 - / 4 -- REVIEWER ,l DATE / O- /4.- 93 ~ DEPT HEAD L!Uierlo.ian lw am DATE IO /4-93
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mghouse units dunng sheduled sprmg outsges. No Com.
"{'
basoon Engmeerug P VR has been idenufied for the in. specuons. The Wesunghouse uruts are Wisconsin Elecmc { f C: JME DN" Power Co.'s Point Beach l (NW. 25 Feb.1) and Amencan
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Elecme Power Co.'s Cook.2: ce B&W PWR is Duke Pou C.IW Wi er Co.'s Oconee.2 (NW, ;3 May,6).
/* MW- De 75% wall thickness limitanon on axial cracks at or above the weld, Numarc says, is consistent with Amencan i
Society of Mechanical Engineers (ASME) Boiler & Pres. sure Vessel Code, Section XI, and provides additional mat.
.g gm agamst complete penetration througa the wall. Calcula.
glO dons have been done to show mat all penetration geome. g0i g mes can support a continuous circumferential crack with a g/~ , depth of 75% of the wall,Numarc's Alex Manon wrote p(2- . . Russell. I' The determination of the length of future service wnh ig cracks is plant specific and dependent on crack geometry s ON ' ' 'ad ia'd=' 'a"d "' ""* ' "ad'- Because Westlaghouse calculations show almost no l&*11W of a crack geanng big enough to rupture without leaking first, the cruaru are aimed more at limiting leakage dunng service. Lankage of borated coolant can corrode me vessel head. The approach for the proposed acceptance cnteria differs from the ASME Section XI methodology, which is calculated by puning a margin on the entical crack sias. In this case, Numarc says, the entical crack sias is far OWNERS PROPOSE DEPTH LIMITS FOR tm large-abat 20 inches according to Wesungham analyms Gnside ,22 Feb, N aHow a pamcal REPAIR OF CRDM PENETRATION CRACKS The U.S. PWR owners groups have proposed to NRC applicanon of that approach, so protection against leakage is the key element. that PWR operators repair cracks in reacsor vessel head alloy 600 penetracons for control rod drive mechanisms The proposal states that cracks must be charactertzed by (CRDM) when axial cracks, at or above &c weld, reach length and " preferably" by depth, but if only length is deter. , 75 % of the penecrauon wall thickness. mined then depth should be assumed to be " half the length The owners groups proposed that circumferential cracks based on expenance with the shape of flaws reponed." at or above the weld os limited to a measured length of 50*.a' -David Srsifox, Warhingros of the circumference regardless of depdt to protect against rupture dunng operanon. However, circumferential e below the weld "are acceptable regardless or their depth, RINGHALS MANAGERS EYE MEPLACEMENT pronded the length is less than 75% of the cucumference." OF HEAD WITH PENETRATION CRACKS Axial cracks below the weld "are acceptable regardless of depth as long as their upper sanomity does not reach the Opersaars of Sweden's Ringhala 2 are considenng re. placing the vessel head now that more crack inh arinne bottom of the weld dunng the pened of service until the nex inspecnon,"se owners groupcommoded. Axialcrack: have been found in t sting of vessel head penetranon welds, which extend through saWor above the weld ase not limited A cucumfematial crack in a weld near um top of the i in length but are limised so a depdt of 75% of the wall thick- vessel head was found in June. As oflast week,10 mes ness, the owners groupe geoposed. mdicanons of circumferential crackmg had surfaced in welds connecting penetranons to the vessel head (NW,16 he proposed crack accapanes criesria for CRDM pen. Sept.,5). etraucms were transmined to NRC's William Russell, assor Ringhals apakmanaa Goesia Larsen said me first alaer. ciale duector of the Office of Nuclear Reactor Regulanon native is to repair the anenny vessel head. But he said that for inspection and rachnical assessment, through the Nucle. replacement with either a new or a ready.made vessel head at Management & Resources Council (Numart) in late July. is also being discussed. '"the ,;ost isn't that big if you com. NRC has not ys t decided whether to accept the crite. ( ria. However, with FRC concurrence, the proposed cQria pare it to what you gain runnmg at full speed in the winner," Larwn said. A new vesse! twad cosung 100-million kronor will be used for the l'ust time in the first quaner of 1994 (U.S.512.5.million) could pay for itself in two momhs, he when U.S. uulities, following the lead of the French and said. Soedes, begin inspecung for CRDM penetration cracking _ Al.eady manufactured vessel heans, cosang "much en vessel heads. " Pilot inspecuans" are currently scheduled less," may be available in the U.S. and France, Lttsen add-to take ptsce at one Babcock & Wilcox umt and two West. 4- ed. Such a vessel head would have to be inspected and NtlCLEONICS MIEK - Septemtw 23.199s
. m c .m,z -
. _ _ _-,m, _ . _ _ ~ . . - - . - .
3- 1 s ,. l
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modified before being insta!!ed but that would probably take less ame than havtng a custom order filled. i More than a thed of the Ringhals.2 vessel head penetra.
' uons have been inspected with ultrasound, boetfor the 4 circumferenual craebng and axial craebng that was discov. ;
cred on the penetrauons themselves last year. A sample e from the most recent circumferendal cracking indicanons is - scheduled to be taken next week and sent to the Studsvik laboratory for analysis. Even if the vessel head isn't replaced now,it may soon need to be. A!Lhough the axial cracks were repatted last year, the stress which caused the cracking remains, which could lead to new cracking. Eventually, especially if more circumferential cracks surface, repaars could become less economic than replacement.-Arunne Sabu, SeckAolm 1
i l l / HANDLED BY b 7M DATE COPIEDI 6bl DATE SENT
/kh/
OTHER TO: M. P. Morroni Technical Manager - Salem FROM: F. X. Thomson, Jr. [ Manager - Nuclear Licensing and Regulation
SUBJECT:
REACTOR COOLANT PUMP SEAL INJECTION FI4W DATE: OCT 215G3 REF: h R-I93561 As requested by your department, Licensing has been investigating a change to the Technical Specifications which would redefine Controlled Leakage as seal water 12 the RCP's. To assist in the justification for this change, Licensing requested information f rom Westinghouse concerning the bases of the LCO, assessment of the accident analyses assumptions, and resolution of values to be used, (Re f. NLR-N9 3 08 5, attached). Westinghouse has verbally responded to our request by stating that it is their belief that the Contro11od Leakage should be measured to the RCP sesis rather than seal leak-off as currently stated. Westinghouse will be assessing the affect on the accident analyses based on recently acquired data and formally documenting their response. At this time, it appears unlikely that their response will be received prior to the end of the Unit 1 eleventh refueling outage. It is Licensing's opinion that this initial assessment could impact the manner in which the seal injection flows are set. If 40 gpm proves to be the naximum total seal injection flow allowable at normal operating pressure, (equating to 82 gpm accident flow used in Westinghouse analyses and cold Leg Injection Cl'w o testing), then the CV98 valves should ce adjusted such that Cinaximum of 10 gpm/ pump is all that can be achieved with the @ full open and no normal charging flow; the actual LOCA configuration during injection and that which was used in the analyses. Achieving the configuration when setting the RCP seal injection flows coming out of the outage would avoid the risks of adjusting flows during power operation should Westinghouse's finalized conclusion necessitate such en adjustment.
%)
~
M. Morroni. 2 NLR-I93561 This-flow limitation might necessitate a change to-operating procedure S1(2).OP-SO.CVC-0001(Q), " Establishing, Charging,. Letdown, and Seal Injection," but other procedure changes, (surveillance, RCP cperation, and alarm response) would not require revision until approval of the_ proposed License Change Request to the Technical Specifications since they would still 4 bound the lower seal flows. 1 Since Westinghouse's conclusions are preliminary, adjustment of the Unit 2 RCP seal flows at this time is not warranted. However, should the opportunity arise such that Unit 2 enters a Mode when adjustments can be made with reduced risk, consideration for similar adjustments should be given. This item has been discussed with C. Lashkari and C. Timm of your staff. If there are-any questions or assistance is needed in any way, please contact S. Mannon at ext. 1129. pRM/ C (w/o attachment) D. Smith (N21) J. Weideman (S02) C. Timm . (S02) C. Lashkari (SO2) H. Berrick (N51) M. Danak (N51) J. Roway (N51) L. Catalfomo (S01)
4 1
\
To: L K ogara, Nuclear Licensing Engineering & Plant Betterment FROM: C P Lashkari, System Engineer %A/hdi-Salem-Technical Department
SUBJECT:
- Review of Dre.ft LCR 91-06-Rev.2 : 1 DATE: November 2,1993 .
-A review of above referenced LCR has been completed and following are my comments:
- 1. Include. Tech Spec # 3.1.2.1.b and place a-* on charging pump and add a footnote explaining charging or safety injection
- 2. Delete "In addition to the requirements of specification 4.0.5" 4
- 3. Add a new surveillance prior to proposed 4.4.5.1 to-verify-manual cycling capability of PORVs after block valves have been closed.
- 4. Add a word " excessive" before." seat leakage" in insert A for Bases section.
- 5. Revise insert-A "All safety injection pumps and charging pumps whether centrifugal or positive displacement type -------- "
- 6. Revise insert-B "A maximum of one safety injection pump or charging pump whether centrifugal or positive displacement type --- "
- 7. All above comments apply to both Salem Units.
C: P O'Donnell-(Sol) S Gillespie (N38) v'J Wiedemann R Lamberger_ R Villar . -(N21) D connell* - (SOS)
/
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31 1 PSEG
- ublic Serece Electne anc Gas C: noanyu o O Scx 236- Pance:rs Er.cge, N J. - 08038 604/339 1100 q
Steven E. Matent>erger vice F'et cent anc CNel Nuciea< Officer , p 2: -. j
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- FROM: . S. E. Miltenberger m $hgk Vice President & Chief Nuclear Officer
SUBJECT:
FOLLOW UP TO DECEMBER 3,1993 MANAGERS' DIALOGUE *
- DATE: December 8,1993 '
Because of the importance of the subject of matter of the last two Managers'
- Dialogues (PERFORMANCE), a follow up meeting has been scheduled. We will meet ori Friday, Janonrv 7 199d nt the Procnecinn Ennter Annme 1 & 1_ This meeting replaced the January 26th Dialocue outlined on the 1994 Manacers' Dialogue schedule you received on Friday. Before January 7,1994 it is expected that you will regroup with your small group from the December 3,1993 meeting and review the.following:
PROSLEM STATEMENT: " POOR OR BELOW STANDARD PERFORMANCE IS NOT EFFECTIVELY RECOGNIZED OR DEALT WITH"
.1 . Catn s of the problem 2, Sory 1 cossible solutions
- 3. The ES ,,ossible solution
'4 The specific action necessary to support implementation of the best possible solution.
On January 7th, each group will report out their best possible solution and the-
. specific action necessary to support its implementation. We will discuss each,
- vote and together decide what action will be implemented.
Attached (for your information) are Nuclear Department roll - up numbers e regarding Performance Appraisal ratings,- OSR and Merit Level. If you have any questions, please either contact your small group leader or Lynne Newman at extension 1156. n La i M
- Attachment /
M-4924e'
. , , . - _:,..;~_,J.. --:-.~.- _ _ = m ..,- <--_ - .- ,
r 1 . PERFORMANCE A'PPRAISALS 1 OVERALL RATINGS 2,000 l 1,532/665 ^ l 1,500 - 1 l 1,000 -- t 658 /2n
- 500 '
O ~ S M N B c E TOTAL NUMBER OF EMPLOYEES =2313
l - . MERIT LEVELS 1,200 1,000 - l l bi4 /6u 800 t ! 600 - i l 400 345 un " 200 - 154 m> - 1,4 m l ML1 ML2 ML3 ML4 i i TOTAL NUMBER OF EMPLOYEES = 1427
-, _....~,..._....c... ., - - ~ . = , ,
e FORM WO.MR-AP.22-0006-1 e NUCLEAR DEPARTMENT INCIDENT REPORT FORM l I [.8 - $~D [ COMMITMENT NUMBER b*t'#[ b/ INCIDENT REPORT NO. USE CONTINUATION SHEETS IF NECESSARY SN "SInbECU%ifEkER (WrcM ord /Of/57 P/PE AllPPL E UNIT (S1,S2,S3,HC): 8] DATE OF INCIDENT: /2 / 1 / D TIME: OdM.
SUMMARY
OF EVENT (IF ESF ACTUATION, INCLUDE SOE. PRINTOUT) :- -- - -
.2^) retini'G 70 L.oe,4re 7ka()dM.t wAi< otJ ll4 eutLL ED iJAr28 PJPItCG, iT uAs :D/s awERED 7xAr 77 tere is MPRo.x_p
.bo AITC RUbSER PATc2 HEL.D &f Hose CLAMP /N 7xgy 70 STOP THE CE4E.YHlY/E $$Nd&PrABL E AER!;9 T'8 ASM E P/P/tdG l'A/Uc-L EAR CLMS 3' )
A)OTE 1W' Pl Pit \CY: m,qu ucr yA\n=// THex) tJeqt L_ LGA K d^LD UT kGA5dRE~MSJVS ftRE GE/AfG &AN#D, REPORTED BY: [# M DEPT:[MMY788MHONE h.4T: 2-75 4 SECTION II (SNSS/ OPS MG) RX PWR AT TIME OF EVENT: -
% UNIT LOAD: -
MWe Op Con / Mode 3C REPORT MADE PER ECG7 (Yh (IF YES, ATTACH ECG COPY) LCO #:J. 9'/0. / A/S #: C'. DATE IN:/ME/f3 TIME IN: //Jo E.R.#: INITIAL CAUSE DETERMINATION: EQUIP / DESIGN PERSONNEL PROCEDURAL OTHER: ~~ REPORTABLE: YES h REASON SNSS/NSS SIGNATURE: I e ed DATE: / L / /o / 9.3 COMMENTS: A . OPERATIONS MANAGER REVIEW :/ [ D?TE: /2 /// /
& t?]OE
- vuoi..r Common Pag. I or- 4 '
R.ir. s j l DE 15 I l
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j. g TO -: .The Technical-Department Manager.
'FROM: aC P Lashkari, Systom" Engineer SUBJ ECT: 4 POPS Setpoint impact on-PTS Concern DATE: January 30, 1994-Vice President Stanley LaBruna required Salem. Generating Station to investigate possible. violation of Tech Spec on RCS Heatup-and cooldown at-Diablo Canyon 1/2 as reported in Industry Briefs Volume II, Number 29,( copy attached ).
An EWR was written to E&PB to evaluate this concern.
- Additionally, Westinghouse identified in PSE-93-204 a .
Potential Issue (PI) regarding non-conservatism in the POPS setpoint development. The pressure difference from the vide range pressure transmitter to the reactor vesnel midplane . (where the Tech. Spec. heatup and cooldown pressure / temperature limits are defined) was not considered. In OE 5691, Comanche Peak SES ( 4-loop Westinghouse PWR) alsa reported that indicated wide range pressure could be lower than actual pressure at reactor vessel / core midplane elevation by approximately 50 psi.
~
Based on above concerns, the NRC issued INFO NOTICE IN-93-56 which shows similar concern has been reported to the NRC by Byron', Zion, Diablo Canyon, Kewanee,Sqquoyah and Point Beach , Salem analysis provided by E&PB letter MEC-93-917 dated December 30,1993 confirmed that calculated pressure at core midplane would violate the Tech Spec Heatup & Cooldown curves. Cooldown Curve at 20 F/HR was considered'in determining the POPS setpoint. E&PB has takea credit for ASME Code Case:N514 which provides 10% margin over Appendix G of ASME Code. Appendix G is also the bases of-Tech Spec
. heatup and cooldown curves. The Code Case is not approved
- by the NRC and its use to defend a Tech Spec violation is not allowed. Licensing Department needs to submit an emergency. revision-to the. Tech. Spec. on heatup and cooldown curves. If.. limitation-on number of reactor coolant pumps in
-operation during-POPS applicability temperature range >
is implemented as a result of this concern evaluation, a 10CFR21 report may still have to be made to the NRC
- based on' calculation error discovered by Westinghouse and confirmed by-the Licensee /PSE&G.
1 l
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! SALT:M UNIT 2 5th REFUELING OUTAGE O&M FINANCIAL PLAN - R/C 069 ROLLUP MILLIONS ! 10 t BUDGET = $9.1 MILLION _._ g _ .... , ,....... .. ... . . .. . .. ... ... . .. . .. . . I 6 ^ 4" , . .'. . . b. . . SP.ENT_.+..COMM!TTED T.HRU 3/18..= 44 MIL. 3/31 s j - - . . . . . . . 4
~
O 3/183/25 4/2 4/B 4/154/224/29 5/6 5/135/205/27 6/3 6/106/17 WEEKS BUDGET ACTUALS + ACCRUALS * ' SPENT + COMMITED , e
,R %
l l
- SALEM UNIT 2 5th REFUELING OUTAGE O&M FINANCIAL PLAN - ALL R/C's MILLIONS _ __
14
- BUDGET = $12.8 MIL - .
3g - . . . ..... ... .... .... .. . .. . . .. .. . . . . 10 - - - - --- -- -- - ---- - --- --- ---- - - - - - - ----- - g .. ............. ....... . .. .... .. . .. . . . . .. . . . . . .... . . . . . . i BEGIN OUTAGE g - . ... ........ . . . .. .. .. . . . . . . . . . . .. . . . G/31 4m. .. . . ... .
; .SPE NT. f..COM MIT.E D..TH.RU. 3/.18..=. 44 MIL... .
Y 2 -- -- -- - -- O~ 3/183/25 4/1 4/8 4/154/224/29 5/6 5/135/205/27 6/3 6/106/17 WEEKS __ BUDGET ACTUALS + ACCRUALS
- SPENT + COMMITED
COMMITMENT NUMBER ' INCIDENT REPORT NO. USE CONTINUATION - EMEETS IF NECESSARY - SECTION I (Initiator) - - - REPORT
SUBJECT:
UNIT , (S1,32, S3, HC) : '$ h DATE OF INCIDENT
, / /3// N TIME: ddD [N
SUMMARY
OF EVENT (IF EST ACTUATION, ' INCLUDE 'SOE PRINTOUT) : RGvIEv) OF elm t.ETTER MEC '93 'll7DA7FD DEC.3c),/993
/ND/Cd7ES THAT PORS SETPOINr AIAs set EcrED NoA1-caisevmww m .
(A)EsTihlGH0USE .DID NOT CoMCGER 17/E C?Ari: GwvAnch) DFFEREwc2
&EnEEGu TRAMSmtrTER TAP LO647/otJ ksb 2EAC7DR/a9E MiDKAIJM AND MIAtvilc &FmcTS hA= To RCP GPER AT/0td. NEsnsrGM6nic HAS tJsb ULCUL ATED' DWJAr4 It' Gfmcr c?F 3/ P3Z' PbR 'M cRcf CPERATIord.IdHEO THls 3lPSE I& AbbED 70 At46PJZ~ Af///CN/3 /ERK ACE i
~ PR,ESSLRE 2XX 7D POPS SErPOIMTprJ7SPST AWD A TRA^JJ/EAJ t CA35nCG RC S (WE5.5L]RE 2McR GASE .f7 i.s agius Anum +1ED 17/RT SftLEM 2ns1=s h/b7 GCE A?cRE 7747M 0AQ 09cP$6L6 2cTDF.
Or/LL U17ECHJPEC HEGrvP/tst,%3rd CdBVESluMirl SRE V/ Ole 7ED. REPORTED BY: 4%rd d'A -O DEPT: TECM _ PHONE EXT: 2759 SECTION II (SNSS/ OPS MG) l RX PWR AT TIME OF EVENT: % UNIT LOAD: MWe Op Con / Mode RIPORT MADE PER ECG7 (Y/N): (IT YES, ATTACH ICG COPY) LCO #: A/S # DATE IN: TIME IN: W.R.#: INITIAL CAUSE DETERMINATION: EQUIP DESIGN PERSONNEL PROCEDURAL OTHER: < REPORTABLE: YES/NO, REASON SNSS/NSS SIGNATURE: DATE: .
/ /
l COMMENTS: l 1
- @PERATIONS MANAGER REVIEW DATE: / /
l Nuclear Counton Page 1 of-4 Rev. S l
. . .x O
+9
i' ! I LTO: R. Everha'm , Sale.m Statio Planner. Salem Maintenance and Planning Department; i
- FROM:- C P Lashkarl , System Enginee Salem Technical Department
SUBJECT:
Suggestion on CRDM Ventilation Fan Repairs 4 E DATE: March 8,1994
+
Your Pace Idea regarding improvements in the process of repairs to the ; CRDM ventilation fans was reviewed within Salem Technical Department using standards and guidelines for such processing. The suggested idea is already covered by the present action plan for these fans based on past job , review and continuous improvemc its in the efficient use of the Salem Station resources. This is evidenen Sy the action plan provided to the Outage Manager on or about August 13. This was before System Er ;ineer became aware of your Pace Idea. Pace ideas must involve innovation, original and unique thinking in l suggesting improvements over and above what is expected by the company in our normal duties as good employees. A questioning attitude aoout past ,
-jobs and practices is e: sected from all good employees as a normal management function.
Processing of Pace Ideas must be handled professionally, judiciously and _ anonymously. Your efforts are appreciated by the program ' If you are not - E satisfied by the disposition, appropriate procedures are available for appeal to the higher management if you so desire. A dispositioned idea can not be
-reopened but a new idea can be submitted if any additional information is to be added to the original Pace Idea.
u
- C:' J K Wiedemann l 1
- R Settle 1
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PSEG ; Pubhc Serece Ewetric ec Gas Company P O Boa 226 Hancocks' Bridge New Jersey 08038 t Nuclear Department Tot J. Wiedemann Technical-Engineer,- NSSS q FROMt E. H.-Villar Station Licensing Engineer - Salem-
SUBJECT:
- PORV POSITION INDICATION REF: NLR-I94239 DATEt. May 24,.1994-
'This. memo documents the result of our meeting on May 19, 1994, and my conversation with Mr'. Jim Stone (NRC Project Manager) regarding the PORVs position le31 cation and the requirements of Generic Letter (GL) 90-06.
As previously discussed, GL_90-06 recommended the PORVs not be cycled in Modes 1 and 2. However, the GL does not unilaterally grant relief from the requirements of Technical Specifications (T/S); it is up'to the licensees to identify and request the appropriate T/S changes. PSE&G submitted its final GL 90-06 application to the NRC on December 8, 1993, and NRC approved the request on April 7, 1994 (Amendment 150/130) with an implementation date of June 4, 1964. However, our submittal failed to recognize that Channel Functional Chr ck (CFC) of Table 4.3-11 needed to be changed. This CFC is-tequired every 92 days and verifies circuitry continuity through the open and close limit switches. This-surveillance has been historically performed by cycling the PORVs through one full cycle. Consequently, the CFC for the PORVs position indication as established by Table.4.3-11 nust be conducted at the required frequency to remain in compliance with the T/S requirements. At'your request, Licensing pursued a technical specification
-interpretation of this . issued with our NRC project manager.
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, -.. _ ~ _ _
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' J. Wiedemann
' NLR-I94239 The result of the discussion between Licensing, and our NRC project manager is presented below;-
-(A) The NRC.(NRR)_will not be pleased: knowing that the PORVs are to be cycled-in-Modes 1 or 2; however, they recognize that a T/S surveillance' requirement must be complied with, and failure to do so will. result in a T/S violation.
In conclusion, GL 90-06 does not prohibit the cycling of the PORVs at power in order to satisfy the requirements of Table 4.3-11 of T/S 3.3.3.7. The required CFC must be performed within the required frequency to avoid a possible T/s violation. ' Note
-that cycling of the PORVs may not be the only appropriate means of complying with the surveillance requirement. Licensing is presently developing a license change request to modify Table 4.3-11 in crder to avoid having to cycle the PORVs in Modes 1 and 2.
Should you have any questions regarding this memo, please do not hesitate to call me at ext. 5429. EHV/ CC F.- Thomson, Jr. J. Morrison P.-O'Donnell P. Ott D.-Smith file-1.4.1 l l l-o
O t TO: 51 PASTAVA, LER COORDINATOR
~
FROht: C P LASIIKARI, SYSTEhi ENGINEER SUBJECT. MISSED SURVEILLANCE TABLE 4.311 ITEM 12 DATE: JUNE 2.1994 BASED ON REVEW OF AVAILABLE PROCEDURES FOR STROKING TIE PRI&2 AND DOCUMENTING THE SURVEILLANCE FOR CHANNEL FUNCTIONAL EVERY QUARTER PER TECH SPEC TABLE 4.3-11 ITEM 12 ON BOTH SALEM UNITS 1&2 IT CAN BE DEFINITELY CONCLUDED THAT THIS SURVEILLANCE HAS BEEN MISSED MORE THAN ONCE ON EACH UNIT. TIE ROOT CAUSE IS IMPLEhfENTATION OF NRC GENEPJC LETTER 90-06. THIS GENERIC LETTER REQUIRED IMPLEhiENTATION OF NRC POSITION ON XOT STROKING TIE PORVs IN MODES l&2. THIS WAS BfPLEhENTED BY SALEM AS OF JANUARY 1,1991, JUST AS REQUIRED BY GENERIC LETTER. TIE PROCEDURE CHANGE REQUIRED OPERATING ENGINEER / OPS MANAGEF. PERA11SSION FOR STROKING PORVs IN MODES 1&2. THERE WERE SEVERAL TIMES WIEN TECHNICAL DEPAR5fENT REQUESTED STROKING THE PORVs TO VERIFY RELI ABILITY OF VARIOUS DIAPHRAGM MATERIAL IN PORV AIR ACTUATOR. ADDITIONAL STROKING ARE RELATED TO DCP OR POST-MAINTENANCE TESTING. PER APPLICABLE PROCEDURE AT VARIOUS TmiES DURING LAST 3 YEARS TIE STROKING OF PORVs WAS MARKED NOT-APPLICABLE BECAUSE PLANT WAS IN MOOF 1 OR 2. THE OPERATING SHIFT WERE IN COMPLIANCE WITH TIE PROCECURE AND NRC GENERIC LETTER. WHAT WAS NOT REALIZED BY VARIOUS DEPARTMENTS IN SALEM STATION THAT STROKING OF PORVs WAS ALSO NEEDED TO COMPLY WITH ANOTHER TECH SPEC ON POSITION INDICATION PER TABLE 4.3-11 ITEM 12. THIS CONFLICT WAS ALSO NOT REALIZED BY LICENSING AND VARIOUS REVEWERS OF THE LCR (AND VARIOUS REVISIONS ) SU3MITTED IN RESPONSE TO GENERIC LETTER 90-06. IT ALSO WAS NEVER PICKED UP BY TECH SPEC ADMINISTRATOR IN REVIEWING THE LCRs AND PROCEDURE REVISIONS WITH RESPECT TO TECH SPEC M/2tlX MAINTAINED BY HIM. THE ROOT CAUSE AS STATED EARLER IS AMENCE OF TECH SPEC REFERENCE AND INADEQUATE PURPOSE IN PROCEDURES IN EFFECT IN 1990 TBE FRAME. PERFORMANCE OF STROKING OF PORVs & TIMING, MET THE POSITION INDICATION SURVEILLANCE IN MIND BUT NOT ON PAPER. THIS ABSENCE OF DOCUNENTATION LED TO THE MISSING OF OBVIOUS CONFLICT IN TECH SPECS BY VARIOUS PERSONNEL IN THEIR REVIEW OF LCR FOR y GENERIC LETTER 90 06. 4
- t I
f t I
- 12, CURRENT l UNTIL PROPOSED LCR REVISES TECll SPEC TABLE 4.3 11 IT'
- PROCEDURE S1/2 OP.ST.PZR. 0002(Q) REV 4 WILL STROKE Tile PORVs AT 92 DAY IhYERVAL TO C0htPLY WITII TECli SPEC TABLE 4.3 11 ITEh! 12.
ATTACilED IS UPDATED COhtPLETED PROCEDURE TABLE AND LIST PROCEDURE REVISIONS C: J WlEDEh! ANN l 4
I SALEM UNITS 1/ 2 PR1/ 2 TESTING ; 4.0.5V / TABLE 4-3-11 ITEM 12 1PRI ! 1PR2 l 2PRI I 2PR2 _ DATE SEC O/X l DATE SEC O/X i DATE SEC O/X l DATE i SEC O/X i i I l 2/21/901.8/.6 i 2/21/90i.7/.6 5/23/90 . 6/.7 l S/23/9dI~9T6' 5/5/90:1.45//1.9 5/5/9011.07/.82 8/23/90 .9/.8 i 8/23/90il./ 6 9/13/90 1./.5 9/13/9011./.7 8/31/9011.13/.8 8/31/9011.01/.81 i 12/23/90 1.2 /.5 12/23/901.8/.7 3/29/911.9/.8 1 3/29/911.7/.7 i ! . 4/12/911.9/.85 4/12/91 [7/.8_ I 4/15/911 NOTE 3 1 4/15/91 tNOTE 3 9/20/9111.1/.4 9/20/91 1.1/.4 5/15/9111.2/.4 1 5/15/9117777 9/22/91I1.2/.6 9/22/91 i l ./.8 i 7/26/91iNOTE 3 I 7/26/911 NOTE 3 7 0/19/911.9/.8 i l 10/3/91 j 1.56/1.25 1 10/3/91!1.43/73 _10/30/91l NOTE 3 10/30/91 NOTE 3 i l l l 11/13/911.97/.8 11/7/91!1.12/.68 i ! ! 12/18/91i 9/.7 i i I l 1/8/9211.12/.81 1/8/9211.05/.25 ; } i 1/29/92'.91/.45 1/29/92I.81/.3 I i j 2/26/92; NOTE 3 2/26/92iNOTE 3 i 5/12/92 . 9 /. 5 i 5/12/92!.88/.56 6/17/92 il ./.6 I 6/7/92 il .2/.7 i 6/16/92._1./.7 I 6/16/92 II ./.8 7/26/9211.1/.7
- 7/27/92!1.31/.63 i 7/11/9211./.7 7/12/921.8/.6 9/14/92iNOT DN i 9/14/9211.1/.2 i 8/20/92 l.6/.6 8/20/921.6/.5 10/29/9211.F 3 : 10/29/92T1.3/.6 i 10/7/92iBLANK I 10/7/92iBLANK 1/21/9311.1/.5 1/21/9311.4/.6 >
1/31/931.9/.6 i 1/31/931.8/.6 5/15/93lN/A i 5/15/93!N/A j 4/29/931.9/.7 1 4/29/931.85/.7 8/16/93!N/A 8/16/931N/A 7/13/93 i t .2/.7 7/13/93 i t .4 /.6 i ! i l 8/26/93lN/A 8/26/93iN/A 9/13/93i?.1/.5 i 9/13/9311.3/.4 i ! i i 10/3/93i.d9/.6 > 10/3/93i1.4/.6 1 10/15/931.96/.66 ' 10/15/93!.95/.57 11/23/93il.5/.75 11/23/93il.65/.77 i i I 12/3/93lN/A i 12/3/93!N/A i 11/14/93lN/A l 11/14/931N/A 4/19/94i 9/.6 i 4/19/94i1.4/.8 ! 12/15/93l1.0/.63 ! 12/15/931.89/.5 l I i l 3/2/94i.8/.46 1 3/2/94i.99/.35
+
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i INCI DENT REPORT 94137 TIME LINE OF PROCEDURE Cl{ANGES l
/
t 03/09/87 1-4.0.$V hilSC 1 REV8 I 03/27/91 REV9-02/28/92 REV10 02/01/93 REVi1 I i 02/17/93 REV12 REVl3 ! ,: 06/04/93 JAN. 91 IMPLEMENTED NRC POSITION ON NOT STROKING PORVs IN h10 DES 1&2. j PER GL 90-06, MISSED T S 4.3.3._7 1 TABLE 4.3.11 St.OP ST.ZZ 003 REVO St.OP ST.ZZ 003 REV1 ; 09/11/92- St.OP PT PZR 002 REVO TlHS PROVEDURE COMBINED ALL PZR VALVE TESTING IN ONE PROC. ' 01/30/93 REYl 06/21/93 REV2 t St.OP ST PZR 002 REVO 06/21/93 09/17/93 REVi 12/03/93 REV2 ! 03/09/94 REV3 , 06/01/94 REV4 AMMEND. IMPLEMENTING GL 90-06 . I r l i 6 i 6 [ nn.- -,,,.--n.,,,,, ,-,--,nw.-,,----,-~,. ,- -,.n..,.,- ,+ ----v.-,,-., - - ..,+--,--.n---,---- ,,--- - - ..n----- ---- ---n
SALEM GENERATING STATION SALEM TECHNICAL DEPARTMENT ENGINEERING MEMO: 94 103 TO: P J Ott, Operating Engineer Salem Operations Department FROM: C P Lashkan, system Engineer Salem Technical Depanment
SUBJECT:
CRDM Ventii. - n On Salem 1& 2 DATE: June 10,1994 Salem Operations Dept. Procedure Sl/S2.OP-AR.ZZ-00ll(Q) alarm respon3e to ber.el alann for high temperature at the fan outlet requires consideration of Unit shutdown by NSS/SNSS if temperamres can not be lowered. This Engineering Memo documents why Unit need not be shutdown. He procedure will be revised reflecting the Engineering position on this subject. During the week of June 10,1994 CRDM Ventilation on Salem Unit I was exhibiting high ambient temperature on the outlet of the motor / fan units. Although bezel alann ( temperature setpoint >l50F) was lit on 14 CRDM fan only, P-250 computer readings were as follows: FAN AMPS FAN EXIT TEMP 11 45 FAILED T/C 12 61 157.7F 13 59 162.2F 14 64 161.5F 4
In addition on Unit 2,23 CRDhi fan is clear & tagged for maintenance. All 3 CRDhi vent, fan bezel alanns (setpoint= 150F) are lit although P-250 (setpoint = 160F) temperatures are not in alarm yet. This Engineering hiemo presents short term and long tenn actions which need to be taken to address OPS concem on CRDhi cooling. SIIORT TERhi REC 05151ENDATIONS
- 1. There is no operability concern at the present time for the Salem Unit 1/2 with the recorded temperature of approx.162F.
Three motors on Unitl and all motors in Unit 2 are from Reliance Electric with insulation class IL The 13 CRDht fan motor is a Westinghouse motor with insulation class F. The motor insulation or both types of motors can withstand temperatures greater than 200F. Ilowever, the motor bearings are limited to 220 to 240F. The bearing temperature is expected to reach these temperatures if ambient air (ctoH wtr) temperature is 170F. This temperature limit is because of thennal expansion of the bearing races and the loss of oil viscosity ( the ability to lubricate the bearing surfaces)per memo from Reliance Electrie. The CRDhi magneticjack assemblics would start experiencing reduced life at temperature in excess of 392F. This temperature relates to the fan exit temperature of 170F, The CRDhi coil stack is limite ' to 175F, Based on the above temperature restrictions, the maximum fan exit allowable temperature can be increased from 160F to 170F. i
- 2. Attached is data taken with AD-46, WO. 940613104 which shows that the system operates cooler with 3 CRDM vent. fans instead of 4. OPS Procedure St.OP-SO.CBV-0001(Q) requires operating with 2 CRDM vent. fans. 3rd fan could be started if temperatures exceed 150F. It is not recommended to put the 4th fan into service.
During the AD-46 and before,11 CRDM fan was found to be drawing 45 AMPS compared to the other 3 fans drawing 55-60 AMPS. When 3 fans were operating including 11 fan, all 3 fans including 11 fan were found to be drawing approximately equal current in the range of 55-60 AMPS. A procedure revision request for changing OPS Procedure Sl/2.OP-SO.CBV-0001(Q) will be submitted by the System Engineer by July 1 1994
%g o%MM J0%& OF TV4 TD M YMuo N 61EU'6., TD CAR
- 3. System Engineering would like control room to record once every shift the CRDM v.mt im outlet temperatures for fans 1/S (indicate which fans are O/S) containment average temperatu e, and containment 136 fl el. temperature
( P 2<0 Pts 1044A or 1064A) for the months of July & August. System Engr wJ' b: ion to provide Station Management trending of the CRDM fans dunn s e.n 2 months. LONG TERM RECOMMENDATIONS
- 1. An inspection of RX head insulation was made on 6/14/94. The insulation was not inissing any pieces. It may have larger gaps than on Unit 2 or the insulation on Unit 1 is degraded compared to Unit 2. This may result in somewhat higher temperatures on Unit I compared to Unit 2. The insuhtion on Unit 1 is planned to be removed ca Unit! during IR12 for CRDM alloy 600 inspection. Any required corrective action will be taken at that time.
- 2. The bezel alarms should be repaired with the existing wom orders during next forced outage on Unit 1.
- 3. The bezel alarm setpoint at 150F should be changed to 170F in order to avoid unnecessary CR alarms. The CR operators can start the standby fan based on P-250/ Doric alann readings. The procedures will be revised to reflect this position. The P-250 and Doric alarm set points should be changed.fc.m l@*f to 170F based on Westinghouse letter on the coil stack and Reliance Electric letter on CRDM fan motor.
~. ) i
- 4. E&PB Revitalization Dept is requested to evaluate the capacity of 4+e each CRDM fan.
- 5. The installation of vibration probes for online monitoring of vibrations on the fan / motor at power will be requested.
APPROVALS Preparer & Date I4M h-/ 7-94 Verifier & Date [. 971uAtr 4//7/9/ Group Head & Date4 f 6-/7 W
CRDMFANS.XLS [ , AD 46 .WO 940613104 -DATE 615 94 6 CRDM VENTILATION FAN EXIT TEMPERATURES I t i t i I TIME 111CROM 112CRDM 13CRDM !14CRDM ' REMARKS
; FAN ! FAN FAN IFAN l CONT AVG T = 103 t I i 1332 116.9 O/S 154.8I'161.1 158.11 1341- 158 122.5 O/S I'161.1 ! 157.31
-1349 157.9 157.2:141.2 O/S I 159.3 1357 158.91 157.2i 159.9i132.3 O/S ALL 4 CRDM VENTILATION FANS 1/S_ i 1410 *161.5 i 156.9I'160.9 159.3 ;
1500 '161.5 t 156.71'160.7 159.31 i 1530 '162.1 ! 157.3 l ' 161.1 159.5j j 1600 '161.7 1 156.9!'161.3 159.51 1630 *161.7 ! 156.91'160.7 159.11 : 1700:'161.5 e 157.2 t
- 160.9 159.51 1730l'161.7 i 156.91'160.9 159.31 11 CRDM iVENT FAN ITAKEN O/S I 1800l119 O/S , 155.4 *161.3 158.51 18301118.2 O/S I 155.4 *161.9 159.11 r
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a..... " ACTION REGUEST DESCRIPTICM ------ -- ------.--TCM2722 C18 STAND fuPUT *w
- DISPL AY pe@E -
ACTION RE2 =0.x _ 950618200 AR STATUS-DATE: IslTED 08/18/95 PACE 1 or 3 DATE 1:ITIATED: C4/T5/95 ImITIAfcR: KEN 0*GARA R 137D_ LAST UPD: 08/18/95 Ret IZIT DUT : WLE, _, sPRIORITY : 3 OUTAGE REeMT: u _ PtaD START: RESP DEPT : mPL_ _ SIGulFICANCE LEVEL: 2 OPERAalLITY7: a REPORTAELE7: E 'AR TYPE : CR _ C.R. INST / led: ,_ Wott Arouse: _ Muc : _ SYSTEM: _ UNIT-CMPWT ID : $2 FEG =0: F/ CMP DESCR LOCATION : EMIS HUNG 7 : _ A/S DATE-flME : PIR mo: Cl3 TAG: REF. WO/ACT: ~ CAC REVlfWED7: _ ASSES $s4ERT/AWIIT f ulTD7: , A/A muusBER:
' StaseARY : INCORRECT (ComSERVATIVE) PEAK PRESSUDE LIMIT LCD met: -
DESC: hEVERAL CALCULATIONS WERE REQuTLY COMPLETED BY as'E 70 (FITRE 3.4-3) SASED 04 A COOLDOWu RATE OF 20 DEGREE F-mR. LICEaSE AEuD. 86 NOWEVE3, THE P/T CURVES FOR U2 WERE REVISED IullGNSE AMEND 129 IMPLEMENTED _ ON APRIL 20, 1994 THE LATEST P/T CURVE IS IuTER*sETED TO MAvE LIMITING PE AK PRESSURE OF AROUe 495 PSIG BASED 0u THE 20 DEGREE-WR COOLDOWu RATE. TME CURVE PEAK PRESSURE L'IMIT IS GREATER TMAN inE 475 PSIG LIMll mailF1 13 int 50.59 AND CDDE CASE N-5 4 $UeMITTAL 70 mRC ADOITIonALLT,WCAP1136( ISSUED 04 05/26/92 IDEuflFIES A LIMif tmG PEAC PRESSURE OF 495 PSIG AT 85
*)EGREES F. TNIS WCAP WAS USED TO SUPPORT LIQWSE CnANGE (AMEmo 129).
MESSAGE: CsetS: I N I T, APP. RE JECT, ACCEPT.CLOSE,CRWD,PF T,Pf 8,htXT,W,PF4,COM,CA4 3 MT JOB LU 866
.......... ....... ~ ... - ACTION REGUEST DESCRIPTIO4 - - - ------ -------TCM2122 C0pe4AND INPUT ===>
DISPLAY 940DE ACTION REG wo : _950818200 At STATUS-DATE: INITED 08/18/ M PAGE 1 0F 3 DAIE 1 ITIAIED: 08/18/95 INITIATOR: REN O'CARA R 1370_ LAST UPD: 08/18/95 RM1 I2If DEPT : mLR PRIORIIT 3 (R.tAGE REGMT: N PtND START: RESP DEPT : RPL_ _ SIGalFICANCE LEVEL: 2 OPERASILITY7: 4 REPORTA8LE7: m AR TYPE : CR _ C.R. thST/IND: WOP". AROUND: MUC : SYSTEM: UNIT-CMP 4T 10 : $2 FEG 40: F/ CMP DESCR: LOCail04 : E*IS MUNG7 : _ A/S DATE-flME : Pit NO: E".lS TAG: REF. WO/ACT: - CAC PEVIEWED7: , ASSESSMENT /ALDIT INITD7: , A/A NupeBER: P#eqARY : thCORRECT (CONSERVATIVE) PEAC PRESSURE LIMIT _ LCD 48R: _- DESC: SEVERAL CALCULATlows WERE RECEuTLY COMPLETED 8T wee 10
- T%IS ISSSUE IS WOT AN OPERASILITY OR SAFETY ComuRM FOR Unli 2. THE CURRENT POPS UNIT 2 TECM SPEC SASES Conflaut$ TO SE MET BASED ou TPE 475 PSIG LIMIT._
THE ACTUAL PEAK PRESSURE LIMIT, PER WCAP 11366, PROVIDES ADDITIONAL OPERAT _
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y , . h I {, gh/ ' , MID-LOOP INSTRUMENTATION TEAM TEAM CHARTER I. DEVELOP PLAN AND SCHEDULE TO ENSURE EQUIPMENT AVAILABILITY AND PELIABILITY DURING REQUIRED PERIODS. II. REVIEW CURRENT INSTRUMENTATION REQUIREMENTS TO DRAIN TO MID-LOOP. CONSIDER THE FOLLOWING: CORE LOADED AND UNLOADED INSTRUMENTATION REQUIRED BY TECH. SPECS. COMMITMENTS ASSOCIATED WITH INSTRUMENTATION REQUIRED DURING REDUCED INVENTORY CONDITIONS (IDENTIFY SPECIFIC COMMITMENTS) INDUSTRY PRACTICES FOR SIMILAR EVOLUTIONS III. REVIEW PAST OUTAGE DELAYS AS A RESULT OF INSTRUMENTATION ASSOCIATED PROBLEMS. IDENTIFY CONTINGENCY ACTIONS NECESSARY TO PREULUDE FUTURE DELAYS. IV. IDENTIFY AND TRACK TO COMPL TION ASSOCIATED PROCEDURE CHANGES. V. INTERRACE WITH OTHER " HIT" TEAMS OR OUTAGE MGMT AS NECESSARY TO DETERMINE SCHEDULE REQUIREMENTS. PROPOSED TEAM COMPOSITICN TEAM LEADER - OPERATIONS DEPARTMENT - JOE SERWAN PROCEDURE WRITER - MIKE DARRAUGH TECHNICAL DEPARTMENT - CHARLIE LASHKARI DAN LAUGHMAN CONTROLS SUPERVISOR - WILL HAMMOND
- - - ~.- -.-- . . - _. . _. . . . - -. - -- .
- 1. Adaptability 7. Professional Development ,
- 2. Initletive/ Accountability 8. Professional Competence
- 3. Customer / Client Satisf action. S. Safety
- 4. Judgement / Decision Making 10. Managing Human Resources
- 5. Interpersonal Communications 11. Innovation S. Project Management Part 11. Professional / Technical Dimensions For the dirnension defined below. read the definitions and enclosed guidehnes. Then document specific examples of- behavior for each dimension, and circle the most appropriate rating.
Significantly Meett Needs Below Mating Exceeds Standard gevelopment Standard
- 1. Adaptability Modify behaviors in response to change: ef factiveness in perferming work under varying conditions. flexibility and an open-minded approach te situations; coping with unforeseen emergencies.
ECustomer/ Client Driven Elean Cost. Effective W Adaptability EProactive/ Adaptive ETeamwork & Collaboration
, Documentation:
i Rating l lSl l lMl l lNl l lBl
- 2. Initative/Accountal lity l Seeking and accepting responsibility for work and work consequences; developing creative j
solutions to problems doing what is necessary to get the job done. Einitiative/ Accountability B Adaptability ECustomer/ Client Driven Elean. Cost-Ef fective ECorporate Citizenship Documentatiotr l Rating l lSl l lM l l lNl l lBl
~
l l l 3 I
Significantly _- Meets Needs Below
-. Exceeds Standard Development Standard (M)
Rating (S) (N) (B) 4
- 3. Customer / Client Satisfaction Identifying and meeting customer / client (internal and externa 0 no is while meeting time, quality and cost requirements.
ECustomer/ Client Driven 5 Adaptability Elnitiative/ Accountability ETeamwork & Collaboration Documentatiort Rating l lSj l lMl l lNl l lBl
- 4. Judgment / Decision Making Making timely decisions and analyzing f acts and data, utilizing business rationale and expertise.
E Initiative / Accountability E Customer / Client Driven E Adaptability 5 Teamwork & Collaboration Documentatiort Rating l lSl l lMl l lNl l l8l . 5, Interpersonal Communications Expressing ideas both orally and in writing by preparing reports, making formal presentations, keeping others informed ETeamwork & Collaboration E Customer / Client Driven E Corporate Citizenship E initiative / Accountability Documentatiort Rating l lSl l lMl l lNl l lBl 4
.' h Significantly - Meets Needs Below Exceeds Standard Development Standard Rating (S) (M) N (s)
- 8. Project Management Organizing tasks and people in order to achieve specific objectives.
ECustomer/ Client Driven ELean. Cost Effactive E Adaptability Einitiative/ Accountability ETeamwork & Collaboration ECorporate Citizenship e Documentatiort Rating l ISI LiMl l INl l }Bl
- 7. Professional Development Continuing developmental ef forts to remain current with state of their profession (e g certification).
E Adaptability Elnitiative/ Accountability Documsntatiort Rating l lSl l -!M I l lNl l lBl
- 8. Professional Competence Possessing and applying state-of the-art knowledge, skills and disciplines essential to perform orofessional/ technical re;ponsibilities; performing necessary research to arrive at reasonable solutions.
E Adaptability Elnitiative/ Accountability Documentatiort Rating l lSj l lMl l lNl l lBl l 5 i , I
,I
l .' Significantly Meets Needs Below Exceeds Standard Development Standard Rating (S) (M) (N) (B)
- 9. Safety (check if not applicable O)
Minimizing public and employee exposure to hazardous conditions. . ECorporate Citizenship ECustomer/ Client Driven Einitiative/ Accountability ETeamwork & Collaboration e Documentatiort Rating l lSl l l Ml l l Nl l l Bl
- 50. Managing Human Resources (check if not applicableO)
Setting performance expectations: providing feedback; developing employees; promoting cooperation and teamwork; identifying empicyees' needs and working with them to improve; encourages adherence to core values. E Customer / Client Driven ECorporate Citizenship Elnitiative/ Accountability R Adaptability Documentatiort Rating l lSl l l Ml l l Nl l l Bl
- 11. Innovatior.
Exploring and implementing new and diffarent ways of solving organizational problems. Ef factive innovation requires challenging the " status quo" with supportive, strategically directed skepticisim and taking risks both generating and implementing alternative approaches to organizational problems. E Adaptability E Initiative / Accountability 5 Customer / Client Driven E Lean, Cost Effactive Documentatio6t nating l l Sj l l Ml l-lNl l l Bl 6
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' SYSTEMLENGINEER PRESENTATION TO MANAGEMENT
- t REACTOR' COOLANT SYSTEM-
- 1. - STATUS'OF UNITS 1&21 UNIT 1 UNIT l' TRIPPED MANUALLY AT 0952 DUE TO HI-HI
- LEVEL IN 13'SG
- 1PS1 WAS ISOLATED DUE TO LEAKAGE, REPAIR PLANS WERE IN- T PROGRESS RX-HEAD VENT' FLANGE WAS FOUND LEAKING ON'1/14/93 THIS LEAKAGE.WAS FOUND TO BE'AFFECTING CONTROL ROD-2D4 FOR LAST WEEKS. REPAIR PLANS WERE BEING DEVELOPED FOR LEAKING FLANGE.
NEW RX HEAD O-RINGS (HELICOFLEX) HAD BEEN SATISFACTORY 13 RCP SHAFT VIBRATION CHNL 1- FAILED INDICATION 11 & 12 RCP FLANGE VIBRATIONS READING 0 & ARE SUSPECT NEED IMPROVED INDICATOR UNIT 2: t
-UNIT 2 AT 100 % POWER ALMOST'ALL~RCS COMPONENTS ARE OPERATING SATISFACTORILY 21 RCP SHAFT VIBRATIONS CHNL 2 HAS FAILED INDICATION 3-
. SIMULTANEOUS FAILURE OF 2R15 AND 2R19A THRJ D HAS
. SIGNIFICANCE FOR SG TUBE RUPTURE DETECTION <
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- 2. DCPs and MODIFICATIONS 2EC-3190,1EC-3219 PORV /PSV DISCHARGE PIPING STRESS &
HANGER ANALYSIS, LOOP SEAL DRAIN PIPING,PZR SAFETY VALVE FLEXI-DISC MODIFICATION BY CROSBY AT WYLE LABS, REMOVE LOOP SEAL INSULATION BOX FROM PSVs, REPLACE ALL EXISTING DRAIN VALVES, REPLACE REMOTE MECHANICAL OPERATORS (REACH RODS) TO ENSURE THEY WORK FROM OUTSIDE EllCLOSURE.NEW SUPPORTS AllD MODIFICATION TO EXISTING SUPPORTS, EXTENSIVE WORK INSIDE PZR E!1 CLOSURE SIGNIFICANT RISK TO OPERATIONS AND MAINTENANCE DEPTs FROM PORV LOOP SEAL DRAINI!1G AND INTERCONNECTION OF DRAIN LIllE WITH PZR. 2EC-3180 REPLACE EXISTING 4 CE CLAMPS WITH ASME CLAMPS WORK ON TOP OF REACTOR WITH CORE UllLOADED WILL ALSO ELIMI!1 ATE CHECKING TORQUE ON STUDS EVERY OUTAGE. 2EC-3161 ADD O-RI!!GS TO AIR ACTUATORS O!! 2PR1 & 2 2EC-3162 ADD O-RINGS TO AIR ACTUATORS ON 2PS1 & 3 SIMILAR DCPs ON UNIT 1 -NUMBERS NOT TAKEN T-MODS EXIST ON BOTH UNITS,THESE T-MODS SHOULD BE REPLACED BY ABOVE DCPs.E&PB TEST PROGRAM MAKIIIG SLOW PROGRESS AT MAPLEWOOD LABS.IF TEST PROGRAM DOES NOT FIND ROOT CAUSE,E&PB WILL NOT PRODUCE DCP. 2EC-3176,1EC-3214 MODIFY PZR ENCL. PLATFORM , OSHA /NRC CONCERN 1EA-1033,2EA-1026,2EE-0034 MINOR DESIGN CHANGE BY E&PB TO CORRECT DOCUMENTS, P-250 COMPUTER ALARM SETPOINT CHANGE FROM 185F TO 175F PER NRC SER 1PS4 5 ACTION STATEMEllT 3. 4.10.1 VALVE ISOLATED NEED MINOR MODIFICATIN TO TUBING LAYOUT. EXISTING DR DISPOSITION ON HOLD. 22 RCP NEEDS BALANCING WEIGHT-WO 920317192 POSTPONED TO 2R7 13 RCP NEEDS BALANCING WEIGHT OF 600GMS -WO 921014138 POSTPONED TO 1R11 2EC-3168 PZR ENCL. REPLACEMENT OF MIRROR INSULATION BY NUCON,THIS PROJECT BY REVITALIZATION IS ON HOLD. FUNDING DURING 93 MAY HAVE BEEN REASSIGNED REVITALIZATION PROJECT PZ-01 WILL PROVIDE PERMANENT TEMP. MONITORING IN PZR ENCLOSURE UPGRADE & RELOCATE E/P CONVERTOR ON UNIT 2 REPLACEMENT OF PS 1&3,PS 2&4 MOTOR OPERATORS ON PS24&28 THIS ITEM HAS BEEN REJECTED BY VP-tlO BASED ON INADEQUATE DESCRIPTION. ONE SI EVENT PREVENTION COULD JUSTIFY IT.IF OPS MGR WANTS,SE COULD PROVIDE JUSTIFICATION FOR AN APPEAL TO VP-NO. E&PB DCP ON DELETING PRESSURE SWITCH & SOLENOID VALVE IN CONTROL-AIR TO PORVs POSSIBLE DCP TO RESTORE AIR CONNECTION ON DOQVe To '"
g-
- 3. TRACKING AND TRENDING ALL IST PUMPS & VALVES ARE TRENDED BY IST PROGRAM PZR EllCL TEMP FROM BITTOLOGGERS IS REVIEWED FROM UNIT 1
& Ut1IT2.THE DATA HAS BEEN ERRATIC , CONFUSING AND OF LITTLE VALUE.
PROCEDURE II-1.3.5 DATA SHEET 1(COPY ATTACHED) IS THE ONLY INFO CONSISTENTLY RECEIVED FROM TECH DEPT CONTACT UNDER TRACKItiG AND TRENDING PROGRAM.I CAN NOT SHUT IT OFF. AN ALTERNATE APPROACH IS BEING PRACTICED BY THIS SYSTEM ENGINEER.ALL COMPONENTS ARE MONITORED BY OPERATIONS UNTIL THE VALUE APPROACHES ALARM OR REACHES OPERATIONAL CONCERN.THIS APPROACH MINIMIZES VOLUME OF DATA COLLECTED AliD MAXIMIZES SE EFFICIENCY.
- 4. HOUSEKEEPING / MATERIAL CONDITION / LOGS ALL RCS CGMPollENTS ARE INSIDE CONTAINMENT AND THEREFORE NOT AVAILABLE FOR ROUTINE WALKDOWN.ALL RCS COMPONENTS ARE I!iSPECTED IN MODE 3 PER PROCEDURE.
- 5. CUTAGE ITEMS SEE SECTION 2 FOR OUTAGE DCPs CONTINUING WITH REFURBISHMENT AND RESETTING OF PZR SAFETY VALVES EVERY OUTAGE.
CHECK PZR SPRAY VALVES FOR LEAKING BY REMOVE T MODS ON PR1 & 2,PS1 & 3 REPLACE / INSPECT STEM ON PR6&7 MAPLEWOOD LABS TO TAKE VIBRATIONS ON RCPs BEFORE SHUTDOWN INITIATE INSPECTIONS ON RX HEAD CRDM ALLOY 600 CRACKING CONCERN,THIS IS AN EMERGING ISSUE WITH THE NRC/WOG MORE DETAILS WILL BE AVAILABLE 1ST WEEK MARCH 93.
OUTAGE ITEMS (CONT.) RCP MOTOR OIL UPPER / LOWER GAGE GLASS-liOMENCLATURE THIS ITEM HAS BEEN CITED BY IllPO IN MOST RECENT ASSIST VISIT. IF HIGH RISK LOCATIOli VALVES FOR BORIC ACID LEAKS HAVE THEIR BOLTI!!G CHANGED OUT BY GENERIC WORK BOOK ! 3,SOME FORCED OUTAGES CAN BE PREVENTED
- 6. KAINTENANCE ISSUES / OBSOLESCENCE ALMOST ALL RCP VIBRATION SYSTEM IS OBSOLETE AND DIFFICULT TO MAINTAIN BECAUSE OF SPARE PARTS.THIS SYSTEM NEEDS TO BE UPGRADED IN NEAR TERM PORV AIR ACTUATOR AND VALVE ITSELF PZR SPRAY VALVES BEING REPLACED BY PZ 01, REVITALIZATION PROJECT REACH RODS FOR PZR VALVES NEED UPGRADE TG ENSURE THEY WORK RCP SEAL INJECTIOli LINES NEED MINOR MODIFICATION TO ALLOW RCP SEAL WORK WITHOUT DRAINING TO MID-LOOP.THIS CHANGE WILL ALLOW CONTINUOUS DRAINING OF BACKSEAT LEAKAGE.
PZR ENCL. CABLE DEGRADATION -INADEQUATE INSPECTION NEEDS CABLE REPLACEMENT AND INSULATION UPGRADE
- 7. NPRDS ITEMS NPRDS FAILURE RATE COMPARISONS WERE MADE FOR PORVs THE PROCESS IS CUMBERSOME AND IS NOT COST EFFECTIVE UNLESS PREPARING FOR AN AUDIT.
- 8. SEEIN ITEMS NO OVERDUE ITEM IN ATS SYSTEM, INDICATING RCS SE IS 7 CURRENT WITH ALL INPO/NRC/ VENDOR OPERATING EXPERIENCE LDOCUMENTS .
AN OE CAN BE GENERATED ON CURRENT LEAKAGE PROBLEM ON RX HEAD VENT PIPING FLANGE JOINT,IF APPROVED BY MANAGEMENT.ALSO IT IS PROBABLY TOO LATE TO GENERATE OE ON PROBLEMS ENCOUNTERED IN DRAINING TO MIDLOOP LAST OUTAGE.
_ __ _ _ . - . ~ . . _ _ _ . _ _ _ _ . _ _ - . ~ . _ . _ _ _ _ . _ _ . _ -
.-- -p-
- 9.- OEF; ITEMS-ALL OEF ITEMS ASSIGNED IN ATS HAVE-BEEN IMPLEMENTED'AND SE IS CURRENT WITH ATS BACKLOG THERE ARE SOME OTHER OEF ITEMS-RECEIVED FOR-INFO ONLY WHICH MAY RESULT IN-PROCEDURE CHANCES IT IS SUGGESTED THAT:OEF REVIEW BY STATION MA1AGEMENT IS CONDUCTED-IN THE PRESENCE OF SYSTEM ENGINEER TO ENSURE MANAGEMENT DECISIONS ARE IMPLEMENTED.
- 10. PROCEDURES ALL PUP PROCEDURES HAVE BEEN REVIEWED AS REQUIRED OTHE
P. PROCEDURE
REVIEWS ARE CURRENT
.NEED TO REVISE CONT WALKDOWN PROCEDURE TO VISUALLY INSPECT FLANGE JOINT ON RX HEAD VENT PIPING AND SMD-REFUELING PROC. TO SPECIFY TORQUE VALUE FOR FLANGE]
JOINT. NEED SALEM-SPECIFIC PROCEDURES FOR MODE 4 LOCA
-NEED SALEM-SPECIFIC PROCEDURES AND STRATEGY FOR MINIMIZING PZR-OUTSURGE PER WOG. RECOMMENDATIONS.
NEED A MAJOR REVISION OF MID-LOOP PROCEDURES INCORPORATING INDUSTRY AND SALEM-1 EXPERIENCE. THERE IS AN IMPASSE ON PUP SUR
V. PROCEDURE
FOR MEASURING CONTROLLED LEAKAGE. REVISION OF PROC. FOR STROKING PRi&2 IS IN FINAL REVIEW.THIS WOULD INCORPORATE NRC RESIDENT'S CONCERNS NRC DID.NOT GRANT RELIEF REQUEST'ON RX HEAD VENT IST TESTING.NEED PROCEDURE & HARDWARE FROM HOPECREEK
- 11. PMs ALL EXISTING PMs WERE-REVIEWED LAST YEAR PRIOR TO INPO VISIT.
NEED TO REVIEW ALL'RCM RECOMMENDATIONS TO ENSURE THAT PMs COMMITTED TO NRC ARE NOT DROPPED. NEED:TO GET A RCM PRINTOUT FOR ALL RCS RECOMMENDATIONS I l l
s 4' Westinghouse identified in letter PSE 93-204 dated March 15, 1993 (NSAL-93-005B) a potential non conservatism in the calculation of the pressurizer overpressure protection system (POPS) setpoint (375 Psig) that effects Salem Units _1 and 2. The pressure difference from tne wide range pressure transmitters (PT403 and PT405) which sense hot leg pressure to the reactor vessel midplane (where the Tech. Spec. heatup and cooldown pressure / temperature (P/T) 'imits are defined) was not considered in the Westinghouse analysis. The Tech. Spec. heatup and cooldown curves are determined in accordance with the requirements of 10CFR50, Appendix G and ensure reactor vessel integrity. The POPS protects the RCS from exceeding the Tech. Spec. limits by opening the PORVs during cold overpressure transients (RCS Temperature below 312
- F) . The c.rrent heatup and cooldown curves (Tech. Spec. Figures 3.4-2 and 3.1-3) POPS limits are 450 and 475 psig for Salem Units 1 and 2, respectively. The Salem POPS analysis calculated a maximum peak pressure during an overpressure transient of 446 psig with the PCRV set at a pressure of 375 psig. Further analysis was required to show that the pressure difference _between the RCS hot leg and the midplane of the vessel plus the maximum calculated pressure (446 psig) did not exceed the Tech. Spec. P/T curves.
The results of this evaluation by engineering are provided ir. letter MEC-93-917 dated December 30, 1993 (Attached). In summary, the calculated maximum pressures ssuming 1 or 2 RCPs in operation compared to the Tech. Spec P'; curves are as folicws: Enti RCPs ir Service Max. Press. Tech. Spec. Limi! 1 2 485 450 1 1 477 450 2 2 485 475 2 1 477 475 Based on the above, when the non-conservatism is removed by adding the pressure difference calculated, the limits of both
;-0060 dated 4/19/94 Salem Unit 's P/T curves are exceeded. DEF#
was issued to evaluate this issue. As identified in MEC-93-917 additional margin on the Tech. Spec. curves can be gained when operating with POPS (RCS less than 312
- F) by taking credit for ASME Code Case N514. This Code Case states that the LTOP systems shall limit the maximum pressure in the vessel to 110% of the pressure determined to satisfy Appendix
,Y l
a
)i G of Section XI, Article G-2215. Crediting the Code Case will allow the maximum allowable plessure (Tech. Spec. P/T limits) for POPS to be increased to 495 psig and 522 psig for Salem Units 1 and 2, espectively. However, utilization of the Code Case will require NRC approval prior to implementation. This Code Case has been relied upon by another utility under these circumstances.
Also, Procedure revisions that have been implemented to limit the number of RCPs in operation to 2 while in Mode 5 will ensure that the maximum pressure will not be exceeded when credit is taken for Code Case N514. As a prudent measure, Salem Operations Dept. is revising the relevant procedures to limit 1 RCP operating during Mode 5. Engineering is in the process of completing plant specific analysis of the POPS utilizing the RHR relief valves (RH3). Westinghouse WCAP 11640 allows plants to credit the RH3 valves for LTOP applications if the Autoclosure Interlocks for valves RH1 and RH2 tied to PT403 and 405 (> 375 psig) have been removed to crevent the inadvertent isolation of RH3. The These interlccks RH3 setpoint along nav'e been deleted from both Salem Units. with the valve capacity were generically evaluated by Westirgrouse to provide the Appendix G protection during low temrerature overpressurization events without relying solely on the PCRVs. Although the plant specific analysas for Salem 1 and
~
2 have not been completed at the present time, the results are expected to produce acceptable results that the present Tech. Spec. P/T limits would be satisfied assuming either i or 2 RCP(s) in Operat:On. These analyses do not credit Code Case N514. Eased on the above, reasonsble assurance exists that the current Tech. Spec. P/T limits would be met when considering the pressure difference between the midplane of the Reactor Vessel and the location of PT403 and 405. Therefore, it is judged that this issue :s not an immediate operability or safety concern. r However, Salem 1 and 2 are considered outside of their design basis per 10CFR50.72 (b) (1) (ii) for reporting purposes. i . I 4* b
a SALEM PREDECisIONAL ENFORCEMENT CONFERENCE _ i % _IR No. Date Discussion of Issues / Apparent Violations 94-32 3/30/94 The original POPS setpoint analysis (SER date 2/21/80) supported a 375 psig setpoint. A 3/15/93 Westinghouse 5 NSAL informed the lices.see of nonconservatisms in the setpoint methodology for POPS for low temperature overpressure transient conditions. After 9 months of analysis, the licensee concluded that the corrected peak Sect. transient pressure would exceed P/T limits (450 psig, 10 2.0 unit 1; and 475 psig, unit 2), i.e., 485 psig. On 12/3/93 the licensee dispositioned the matter by administratively limiting operation to 2 RCPs when less Sig, than 200* and increasing each unit's P/T limit by 10% Issue based on unapproved ASME coC case N-514 (10CFR50.60). I 15 Though the licer.see knew that the design bases was exceeded, the condition was not reported (10CFR5^ 72/73). In early January 1994, the licensee recognized tne inappropriateness of using an unapproved code case and subsequently elected to take credit for the capacity 20 provide by RHR system suction relief valve RH3 to augment POPS relief. The licensee's analysis indicated that with RH3 available, the transient peak pressure would remain below the App. G limits. The licensee took credit for, d but coatinued to analyze the ability of, RH3 until April 25 94, though no 50.59 evaluation or approval from NRC for TS amendment was initiated. in April 94, a DEF was generated which identified that RH3 was not credited in the POPS analysis or in the design bases. [ Note: For over a year, the licensee had failed to effect any 30 corrective action or come to any resolution of the POPS setpoint issue (10CFR50 App. B, Criterion XVI)]. Though the issue was finally entered in the DEF process, the fact that Salem was outside design bases was still not reporteo. To address im ediate safety concerns, 35 procedui: evision was ma ! to assure that only one RCP would be available in Mode 5 (to limit the dynamic head error affecting P/T limits), and efforts were taken to assure the availability of RH3 (though no TS amendment was initiated). [ Note: Throughout this period, the 40 licensee's effort was directed toward developing a rationale to support that the nonconservatism expressed in the NSAL did not apply to Salem.] In a final attempt, the licensee elected to change the bases upon which the original POPS analysis was founded by relying on 45 procedural controls to limit possible injection sources. By limiting magnitude of mass addition, and revising the limiting transient upon which the setpoint was based, the licensee was able to predict a peak transient of 438 psig (i.e., below the P/T limits) and therefore considered 50 that the existing POPS setpoints continued to be valid. However, as of 12/94 the change in the POPS design basis was not reviewed to determined if an USQ existed (10CFR50.59). G: SALEM.PEC 1
/
, / .J 4 l
jIRNo. Date l Discussion of issues / Apparent Violations J 95-02 4/7/95- A. In refueling outage 2R7 (May 93) the licensee
- installed a design change to accommodate the results of their analysis of pressurizer safety valve performance concerns discussed in 0737, Item 5 Sect.. 11.D.1. The DCP involved system modifications to 4.4 remove the loop seals associated with 2PR3, 4, and 5 by establishing a drain system for the loop seals Sig. (with associated drain valves and header; and Issue change of the valve internals to materials designed 10 to operate in a steam only environment. Upon completion of the modification, the drain valves were added to the valve lineup scheme,. including the common isolation valve 2PR66 in the common drain header. The required lineup specified 2PR66 15 to be open, however the lineup was not accomplished. Further, no evidence could be determined that a post modification test had been accomplished relative to the DCP. Consequently, the 2PR66 remained closed causing water due to 20 condensation to remain within the loopseal throughout the operating cycle to 2R8 (October 1994). As a result, a safety related system existed in an unanalyzed configuration (10CFR50, App. B, Criterion V).
25 B. The following are issues involve 10CFR50 App. B, Criterion XVI. 8.1 On June 7, 1994, material management documentation 30 for limit switches associated with head vent valves erroneously identified the material as NSR. A DEF identified that a switch short circuit could render two head vent valves inop since they were powered Sect. from the same common circuit. However, the DEF did 35 4.5 not identify any operability or safety concerns based on the reviewers conclusion that, whether NSR Agg. or SR, the switches were the same and were issue differentiated only by test certification for the SR component. In February 1995, the licensee 40 determined that NSR switches were actually installed in the Unit 1 vent valves. However no safety evaluation or analysis was performed to demonstrate the acceptability of NSR parts installed in a SR application or the bases for
-45 continued operability of the system and unit.
50 G: SALEM.PEC 2
. L ikfio.Jbote Discussion or issues / Apparent Violations l 95-02 4/7/95 B.2 On Feb. 24,1915 Unit operators put the control of i a PORV in manual mode, rendering it inop, but Sect. failed to adhere to the TS 3.4.3 AS to closed the 2.3 block valve within I hour. The condition was 5- identified and corrected 23 hours later. This is a
-Agg. repeat of a similar occurrence involving Unit 2 on issue March 24, 1994 B.3 On July 6,1994 head vent valve 2RC40 failed to 10 operate during testing while Unit 2 was in cold shutdown. The licensee speculated that the low RCS temperature caused boric acid crystallization that Sect. prevented the valve from functioning. Later when 4.6 RCS temperature was increased, the valve stroke 15 test satisfactorily, convincing the licensee of the Agg. accuracy of their speculation. The valve was issue returned to service on July 10, 1994 with no further review, evaluation, or assessment iaw normal work control process procedures.
20 Consequently no actions were initiated to address maintenance, operability, corrective action issues, or generic implications. 25
'G: SALEM.PEC . 3
lIRNo. Date Dis s' ion of I .Jes/ Apparent V olations j 95-07 5/24/95 The following issues involve 10CFR50, App. B, Criterion XVI. Sect. 1. An oil sample lab report dated 8/4/94 recommended 5 -4.3 resampling and changing oil on the 21 high-head SI Agg. pump due to significant increase in wear particle issue concentration. Sect. 2. An oil analysis dated 11/28/94 identified high wear 10 4.3 particle concentration in the 22 high head SI pump Agg. speed increaser oil. Issue In both these cases, the system engineer, though aware of the findings of the lab reports, did not initiate any 15 followup, evaluation, or corrective measure; or establish a bases for operability or reliability in view of the apparent degraded condition of the equipment. The degraded nature of the equipment was not entered into the Equipment Malfunction Identification System (EMIS) until 20 March 20, 1995. Sect. 3. A lab report, dated 10/6/94, recommended resampling 4.3 the 23 AFW turbine lube oil due to some small Agg. amount of water found and an increase in wear 25 issue particle concentration. The degraded nature of the equipment was not entered into the EMIS by the system engineer until 3/27/95. The system engineer did not initiate any review, evaluation, 30 or establish any basis for equipment operability or reliability. Sett. 4. In May 1994, a systems engineer initiated a work 4.3 request to inspect the 2Al 28 VDC battery charger 35 Agg. due to configuration concerns involving the ground Issue detection circuit. The work order to accomplish the task was not issued until April 1995. Sect. 5. LER 95-05 identified seven instances. between 40 4.3 5/8/90 and 1/14/95, of PSVs being beyond the 1% Agg. tolerance required by TS 4.0.5 for Unit 1. Four Issue instances were identified between 11/14/94 and 1/14/95 involved 2 or the 3 inctalled PSVs. In all instances, the vendor notified the appropriate 45 system engineer by telephone, with written followup reports. Notwithstanding, the responsible system engineer never initiated an Incident Report, and consequently, root cause, operability, and reportability actions were not accomplished. C: SALEM.PEC 4
-, 'IR No. Date Discussion of Issues / Apparent Violations 95-10 7/14/95 The following_ issues involve _10CFR50, App. B, Criterion-XVI.
Sect. 1. _ Though aware of degraded equipment condition -
-5 2-4
. affecting the-22 RHR and the 21.RHR pump minimum
- recirculation flow valves since-1/26/95 and 2/9/95, Sig, respectively, the licensee did not initiate any- i Issue action to-determine and correct the cause of the' condition or establish a basis for operability of 10 the affected RHR systems until 6/7/95 when Unit.2 '
was shutdown iaw TSAS. , Sect. 2. Though aware of degraded equipment condition
- 2. 5' affecting the 12 control area switchgear supply fan
.15 motor.(CASSF) since 12/11/94, the licensee Sig. initiated no action to correct the condition, Issue evaluate _ operability with regard to design basis, or acquire replacement equipment (which was known to be obsolete. The design basis expressed in 20 UFSAR 9.4.6 describes a design consisting of three-50% capacity supply- fans-two operating; one in-standby. On 5/12/95 the 13 CASSF motor trip on overloadi In response the licensee attempted to develop a justification for continued operation i 25 with two of the three fan motors inoperable based on a dubious raticrele. Finally, after being unable to justify operability, the licensee shut the unit down iaw TSAS on 5/17/95. >
30 Sect. 3. 0n 3/6/95 the Unit 1 personnel airlock failed a 3.6- leak test. No root cause assessment was accomplished. PSE&G wiped the seal with a masolin Agg. cloth (oil impregnated) to remove any dirt which Issue was presumed to be the cause. Retest was 35 satisfactory. On 5/3/95, the airlock again failed the leak test. The same solution was applied. Retest was satisfactory. On 5/8/95, the leak recurred. The licensee was prepared to apply the same corrective action until root cause was 40 challenged by the SRI. _ Subsequently, a thorough root cause assessment was performed which identified that the airlock seal was deformed and
- was the cause of the recurrent leak. The previous wipedowns with masolin cloth.merely applied an oil film which' temporarily masked the leakage.
t I % 4 l l- G: S AL EM. PEC-- 5 E l'
., _ _ - , , ~ . .m_. , _ , - . . . _ _ __ .- ,_ _ . _. - _ . . . _ _
, [IRNo. Date [Discussionof issues / Apparent Violations ! 95-10 7/14/95 4. Since 2/29/92, sevsral instances occurred involving i Cont. failure (cracks) of the threaded portion of the a pressure switch instrument pipe nipples associated Sect. with emergency diesel generator jack water cooling 5 6.1.B system. Previously, the failed component was merely rethreaded and/or replaced without any root Agg. cause ef fort being applied. On 6/7/95, the issue licensee finally determined that the cause was due to resonance frequency that could be treated nj 10 modification of the size and mounting of the pipe. Sect. 5. On 7/11/92, the 21RH10 (21 RHR pump discharge 6.1.A isolation valve) was observed to be " clanking." on 4/16/93 the valve was opened and inspected to 15 Agg. investigate the cause of the noise. Two deep wear issue marks were discovered on the disc, but engineering concluded that they did not affect operability of the valve since seat damage was not observed. The marks were buffed out; the valve reassembled, and 20 placed in service. On 6/10/95, the loud noise was once again observed. While insrection of the valve was planned, the 21 pump was not considered'to be affected even though the licensee had no basis to support the conclusion in view of the possibility 25 (as described by the SRI) that if the disk separated from the stem (due to whatever mechanism was causing the loud impact noises), RHR flow would be lost or restricted. Subsequently, the licensee performed a more thorough assessment and developed 30 a reasonable rationale (based on input from the valve vendor and a search of industry experience with the valve type) that support that separation
; of the disk from the stem was not a likely failure.
Subsequently, the licensee committed to examine the 35 valve and determine root cause once the system could be removed for service. G: SALEM.PEC 6
.. ORIGINAL DESIGN-
' r-PR3.4'.kS(TYP) .
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