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, .S NUCLEAR REQUt.ATORY COMMISSION' REolON 1 g*****,/ ._478 ALLENoALE MOAD KING OF PMUSSIA. PENNSYLVANIA 19408 L JAN3o199; Docket No. 50-336 File No. RI 90-A-225 Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka Senior Vice President Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Mroczka:
The U.S. Nuclear Regulatory Commission recently received information concerning activities at Millstone Unit 2. Details of this concern are enclosed for your review and followup. We request that the results of your review and ~ disposition of this matter be submitted to Region I within 30 days of the date of receipt of this letter. We request that your response . ' contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure. The affidavit required by 10 CFR 2.790(b) must accompany your response if proprietary , information is included. The enclosure to this letter should be controlled and distribution limited to personnel with a -
"need to know" until your investigation of the concern has been completed and reviewed by -
NRC Region I. The enclosure to this letter is considered Exempt from Public Disclosure in accordance with Title 10, Code of Federal Regulations, Part 2.790(a). However, a copy of this letter excluding the enclosure will be placed in the NRC Public Document room.- The response requested by this letter and the accompanying enclosure are not subject to the clearance procedures of the Office of _ Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.- mGygF 47cC roj@RTION,-gmp m$L%&Eswamt, m , 9212220071 920608 ' g aV f[/[i( PDR- FOIA . CUILD91-162 PDR
Northeast Nuclear Energy Company 2 Your cooperation with us is appreciated. We will gladly discuss any questions you have , concerning this infor nation. Sincerely, M Charles W. Hehl, Director Division of Reactor Projects
Enclosure:
10 CFR 2.790(a) Information cc w/o encl: Public Document Room (PDR) Local Public Document Room (LPDR) State of Connecticut
Arrenu u o.9 ., SMPLE RECORD OF ALLEGATION PANEL DEC1510NS ' $1TE: MS~L PANEL ATTENDEES: ALLEGATION NO.: k/ - 9b- g-tlS Chairman - (/),'ac/oS DATE: IfSfgr (Mtg.h234$) Branch Chief - . kmb-n PRIORITY:. High Medium Q Section Chief ( AOC) - bb;[ SAFETY $1GNIFICANCE: Yes No Others - CONCURRENCE TO CLOSE0VT: 00hSC bif(-OE CONF 10ENTIALITY GRANTED: Yes No g\g bhb,j 'hk$ (See Allegation Receipt Report) 15 THEIR A 00L FINDING: Yes 15 CHILLING EFFECT LETTER WA T Yes No HAS CNILLING EFFECT LETTER BEEN ENT: Yes No RAS LICENSEE RESPONDE0 TO CHI ECT LETTER: Yes No ACTION: M c e k h a .M d_ qa dms ICCAML K [ e L 'bRs C Grod tw d uk b wo LLs . U n keok M es/CCn w r
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M n-M . NOTES: A4-1 d .ifI
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' ALLEGATION RECEIPT REPORT Oate/ Time // Allegation No.. AZ"f*-A-0115 Received: /1//7/70 /33b
' /
(leave blant.) Name: /761gM04d Address:
/
Phone: City / State / Zip: Confidentiality: / Was it requested? Yes No ,/ Was it initially granted? Yes No' Was it finally granted by the allegation panel Yes jo N Does a confidentiality agreement need to be sent-to alleger? Yes No Has a confidentiality agt.2ement been signed? Yes' No Memo documenting why it was granted is attached? Yis No Alleger's fl / /
/
Employer: U/16hou>n Position /
Title:
/ Facility: /N, b f 1 Docket No.: 60-3 % (Allegation Summary (brief description of concern (s): AlenAS/n/[4wer ()Seth Q! k?i/bbt b , kNG nnfawh&,fh -Ob? {%Ya tb, Number of Concerns: f Employee Receiving Allegation: b- ac, mmb . . 2da7 (first two/initi,als and last name) Type of Regulated Activity (a) f Reactor (d) _ Other: Safeguards (b) _ Vendor (e) (c) _ Materials (Specify) Materials License No. (if applicable): Functional Area (s): __(a) Operations _ (e) Emergency Preparedness b Construction[(f)OnsiteHealthandSafety Z((c))Safeguards (g) Offsite Health and Safety _ _(d) Transportation _ (h) Other: (NRC Region I Form 207 ' Revised 10/89) a
4 ALLEGATION RECEIPT Mon,Dec 17.1990 ALLEGATION NO.: RI- A-90 __ 1:36 PH Name: Anonymous Address: Unknown Phone: Unknown City / St: Unknown Confidentiality: Was it requested ? Yes _NA No Was it initially granted? Yes No Was it finally granted by the allegation panel? Yes No - Does a confidentiality agreement need to te sent to the alleger? Yes No Has a confidentiality agreenwnt been signeo ? Yes No tiemo documenting why it was granted is attac hec? Yes No Employer. Unknown Position /
Title:
Unknown Facility tilLLSTONE 2 DOCKET NO.: 50-336
SUMMARY
- One concern regerding the acceptability of using non-ASt1E certified flanges in ASME systems at MP2, as addressed by NNECO in Nonconformance Report 290-063. The issues raised include:
appropriateness of NNECO's disposition of the subject NCR; an inadequate ASt1E material procurement, inspection and issue program; an apparent violation of 10CFR 50 (App B), Criterion Vill; and, f ailure of the audit ( program t; identify the above. See the attached materials supplicd with the - anonymous allegation. nut 1BER OF CONCERNS: 1 EMPLOYEE RECEIVING ALLEGATION: WILLI At1 J. RAYMOND ACTIVITY; Y REACTOR FUNCTIONAL AREA: (a) _.X_ Operations (f) ___Onsite H&S (c) _ Safeguards (g) ___ Offsite H&S (h) __.0ther
4 A CLEAR EX At1FLE OF NUCLEAR SAFETY CONCER IS PRESENTED BY NCR tt290-063. THE DISPOSITIO!4 STATES: UEE-AS-IS ASTri MEETS THE REOUIREMENTS OF ASME. THIS t1AY BE TRUE FOR SECTION II, BUT NOT FOR SECTION III. WHAT ABOUT THE REQUIREMENTS OF NCA 38007
-THESE PURCHASE ORDERS DO NOT LIST A YEAR OR ADDENDA OF THE CODE, tiOR A CLASS OF MATERIAL
-THE CERTIFIED MATERIAL TEST REPORTS DO NOT LIST A YEAR OR ADDENDA OF THE CODE, tJCR A CLASS OF MATERI AL
-THERE IS NO CERTIFICATION TO ASl1E SECTION III AT ALL AtJD NO NCA 3800 PROGRAM STATEt1Et1TS. _
-NO CERTIFICATION WAS RECEIVED FROM ThE SUPPLIER THE JUSTIFICATION FOR FDCR E-125-79 (FA 79-155) STATES:
"IT IS NOT FOSSIBLE TO DETERt11NE EXACTLY WHICH LINE THE FLANGES COULD BE ItJSTALLED IN..."
HOU CAN THE SUPERVISOR OF ENGINEERIING AND THE UNIT SUPERIl4.ENDANT COtCURE WITF.1HIS BLANTANT FAILURE TO FOLLOW 10CFR50 EECTION VIII? WHAT LIND OF OUESTIONS WOULD THIS RAISE ON A 10CFRE1 NOTICE? Trils EXAMFLE CLEARLY SHOWS THE COMPLETE LAC}-; OF ANY MATERIAL TRACEABILITY NORTHEAST UTILITIES HAS, TO SAY NOTHING OF I N ADEOLtATE ASt1E t1ATERI AL PROCUREt1ErlT , I t1SFECT I ON OR ISSUE PROGRAMS. THE t1 DST CURSORY OF AUDITS SHOULD EASILY DETECT THIS GLARING SAFETY CONCERN. A COPY OF THIS HAS BEEN SENT TO THE SITE f1RC INSPECTOR. I pM s>h nk Aahl2 % } ,dk .
- . - ..n , , - . , , - . - . - - . ~ ~ . .- _, _ ... ~- ,. .,_ ..-,- _ .a ..a OPS $as REV.1140 NONCONFORMANCE BEPORT-Stt MO 3 05 POR NSTR'J0TioNS LNT
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, / PACKACING LEVEL "C"
/ ALL HATERI AL IS TO BE PROVIDED IN ,
2: ACCCRDANCE KITH THE ATTACHED NUSCO CUALITY
,f., . ' , ' ASSURANCE SPECIFICATION NO. 4.01-Je REVIS:0N 2 C 04 R ESP OM OE N C C SHi > Pt NO
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- t. 508481 INVOICE T03 NORTHEAST NUCLEAR ENERGY CO. . '-
ma ut coeerscoworsce ne ewoices to rvacauno Duf r. o un vo, HAmono,co,* ostoi D^tt
. JANUARY 3, 1978 l, ,, ;3:Y'j,{g pCUNNINGHAM SUPPLY 1NC ] FURNISH AND SHIP TO:
NORTHEAST NUCLEAR ENERGY CC. We MILLSTONE POINT GENERATING STA. R. A.GR1SWOLO SUPERVISCR STORES END OF HILLSTGNE ROAD { - PAGE 05 j WATERFORD, CONN. 06385
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OEST F77867 P. J. GILLEh
, , CCNNECTICUT S ALES T AX OF 72 APPLICABLE - MATERI AL ONLY IF YOU ARE AUTHORIZED TO COLLECT , .L; r . U A L & r.U , U. huMUUUlble Mdib 4-174-77. GA LAI. 1 P/ b l O LK OW ANTITY 0ESCAIPTION S'OCA CODE PRICil 6 Ot& COUNT 00CUPENT ENTITLEC "REQUIRENENTS FOR A MATERIAL DISTRIBUTER e 5 (SUPPLIER) CUALIT Y ASSURANCE g PROGRAM".
00CONENTATION IS TO BE PROV10E0 IN ACCORDANCE WITH THE REQUIREMENTS SPECIFIED ON THE ATT ACHED OUALITY VERIF lC ATION 00CUNENT ATION LIST. - FOR ALL ITEMS ABOVE 2%10 5 CAYS
. NET 30 DAYS DELIVERY AT DESTIKATION REQUIRED BY 01/13/78 IN ACCORDANCE WITH YOUR CUOT ATION CATED 12/27/77 ,
CONFIRMING PHONE ORDER 12/23/77 TO 0. GIRICUX. 00 NOT DUPLIC ATE. t r 9 COR R E SPON DENC E SNIPPING P.O.A CK. R E C' O DATE peeveset sawf atPkT mEC's seespesase scaetoubt sawt stame
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- saa Au coeer morwer no Hvoet s to rueoutNo 001, P. o. tox no. HArtrono, coeet o4ici D^tt ' "
- JAhUAPY 3, 197a. .
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p ] FURNISH AND SHIPTO: CUNNINGHAM SUPPLY INC NORTHEAST NUCLEAR ENERGY CO.
. . */* NILLSTONE PolNT GENERATING STA.
R. A.GRISWOLD, SUPERVISCR STORES END OF HILLSTONE ROAD L PAGE 04 j WATERFORD CONN. 06385 1.=w.;.x .. DEST FTT869 P. J. GILLEN CONNECT 1CbT SALES tax 0F T2 APPLICABLE - N ATERI AL ONLY IF YOU ARE AUTHORIZED TO COLLECT LLE U. huhDWUthi, t. U Ab i fiu , f M A H 4-llb-ll. WA~~ CAI. 1 P/dILLK OV ANT ffY D E 5 C R 1 P,TI O N STOCR COO 4 PRICE & Ol$ COUNT ( 6 EACH 2" 30Cp FLANGES, SOCKET WELD CIRO14 $45.24 EACH
- 6 EACH 21/2" 30C i FLANGES. SOCKET WELD OIRO15 $56.84 EACH s
6 EACH 3a 300# FLANGES, SGCKET NELD CIR016 $74.24 EACH CERTIFICAT E OF CCNFCRNANCE TO A182, TP316 V CHEM. PHYS. AND NDE TEST REPORTS. - 7,.g.h' .,. . O.A. REQUIRED P ACK A(E FCR ANSI N45.2.2 PACKAGING LEVEL *Ca.
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ALL HATERI AL 15 TO BE PROVIDED IN ^ 9DJ'
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ACCCRD ANCE WITH THE ATT ACHED NUSCO CU ALITY ASSURANCE SPECIFICATION W. 4.01-J REVISION 2 S.
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- ' I I ,,.; NUMBER . ' . . i , , INVOICE'TO: NORTHEAST NUCLEAR ENERGY CO. -
m Att coeersecnetHct AND DWO*Cf 5 TC PVtCHA5NO Derf, P. O. 6OE 270, HARTPOtD. CONH. 04101 *
*'. DAtt JAhuARY 3e 1978 ' #
, , [I[jh';
'. p, CUNNINGHAM SUPPLY INC ] FURNtSH AND SHIPTO:
NORTHEAST NUCL EAR ENERCY CU.
</ * ! . HILLSTONE POINT GENERATING STA. ;
R.A.GRISWOLD, SUPERVISOR' STORES ' END OF NILLST0hE RO AD
," { . P AG E 0 5 j . WATERFORD, CCNN. 06385 i
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OEST F77869 P. J. GILLEN
! CONNECTICUT SALES TAX OF 73 APP!'C'.RLC - MATERIAL -
ChtY IF YOU ARE AUTHORIZED TO COLLECT i L.3 U. (sunu bu l b I., t. AJ A b i t'.U . F1M A M J-l l :r- I I. WA L A 1. 1 t- / s I G CK OW ANTITY . OCSCA8PT4ON evoca coog PRICE 6 De6 COUNT l DOCUMENT ENTITLE 0 " REQUIREMENTS FOR A HATERI AL DISTRIBUTER'S (SUPPLIER) DUALITY ASSURANCE
.I. P R O G R A M .
DCCUMENT AT ION IS TO BE PROVIDED IN ACCOP. DANCE WITH THE REQUIREMENTS SPECIFIED ON THE
- ATT ACHED OUAL ITY '
VERIFIC
- ATION DOCUMENTATION LIST.
FOR ALL ITEMS ABOVE 2310 DAY:
. NET 30 DAY' DELIVERY AT DESTINATICN REQUIRED BY 01/12/78 IN ACCORDANCE WITH YOUR 000TATICN DATED 12/17/77 '
CCNFIRMING PHONE ORDER 12/23/77 70 D. GIRIOUX
, 00 NCT DUPLIC AT E.
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I. COM R ESP ON DC NC E SHIP #t MO P.O. ACK. REC O DATE seeoutav sgat mEPLT REC D s.uppiese scessouLg SENT$7RPR ,
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, s Al.LECATION RECEIPT REPORT ece v [ l th!7C Allegation No. b7" Yd"4 " O D3 b , (leave blank) Name- Address: Phone: City / State / Zip: Confidentiality: Was it requested? Yes No d Was it initially granted? Yes No Was it finally granted by the allegation panel Yes No Does a confidentiality agreement need to be sent to alleger? Yes No Has a confidentiality agreement been signed? Yes No Memo documenting why it was granted is attached? Yes No A11eger's i Employer: _ M dhii<3 _ Position /
Title:
L Facility: MS 4 Docket No.: SO4M (Allegation Summary (brief description of concern (s): f i b re e d k t vitc4 Talau M MGiAbut cw MF9(tcble'ri, d d b w h a k e.e d - v i
' /
Number of Concerns: k Employee Receiving Allegation: bdwd ! b bbb (first two initials and last name) Type of Regulated Activity (a) 1 Reactor (d) _ Other: Safeguards (b) _ Vendor (e) _ (Specify) (c) _ Materials Materials License No. (if applicable): Functional Area (s): A.(a) Operations e) Emergency Preparedness (b) Construction f)OnsiteHealthandSafety (c) Safeguards ~ g) Offsite Health and Safety (d) Transportation h)Other: (NRC Region I Form 207 Information in this record was deleted Revised 10/89) , in a:cordance withj thg &qdom of Information I Act, exemptions U F b F0IA 4 HL A - s o h.)1lL~M
203 443 5893 o DEC 18 90 12:06 NRC MILLSTONE OFFICE P02 4
.- ~r-~.
hCETING WITH TIMCs ,, , PATE: DECEMPER 13. 1990 TOFIC: CONCERNS 04 B1EF#J!AL PROCEDURAL REVIEW
- 1. ( , fbiennialreviewot procedure IC 241?C. The procedure application was to replace ercore nuclear instrumentation.
lDuringtheprocessofthebiennialreview questions surrounding technical specification appitcability. ( , J using a recently assued IC department instruction 3.02 initiated on 11/26/90. The department instruction provides fur ** er details than the associated administrative control pro 67 Jure, l~.' '~~~'~ ' ppproached the autner of the department an s t ruc tion'\ , with the questions on the review process. , j to the IC department head vsa a memo on questions surroundsog the biennual review process. ,
! s actions were
- s. ppropriate. and ' us not a free open CD&sunication...
enutronment.. - i i i l , noted no departmental training was efforded in the biennual review process, and questionec I 11 technicians were the appropriate personnel for the biennual review.
- 2. stated during restoration from the refuel outage, that maintenance work on the main
~
feedwe r pump coupling was accomplished without a tag-ou.. The shif t super, visors involved were[ .____ The alleger stated a pre-thought decision was made not to take out the feedpump during coupling replacement. InitrJMr f ecependa tion
- 1. t6 further action.
l . - - . . - . . - , - , - - . --. -. ,- . . . .- --
_ yF e ..o 231 SMPLE RECORD OF Al.LCMT10N PANEL DEC1510N$ f
$1TE: A,l-id *d "" 013 D PANEL ATTENDEE 5:
AltEGATION NO.: 475 cl Ch.tr.en - dh/e/h DATE: /2doko (Mtg.h2345) Bra u h Chief - kC b hu 5ection Chief (ADC) - N PRIORITY: High Ktdtus @ Others - M4/*de's-O/r 5AFETY $1GNIFICANCE: Yes @ Unknow CONCURRENCE 10 CLO5E00T: 00 @ $C CONF 10ENTIALITY GRANTED: Yes o ($ee Allegation Receipt Report Y No
!$ THEIR A 00L FINDING:
WARRANTED: Yes No 15 CHILLING EFFECT LET TER BEEN SENT: Yes No HA$ CHILLING EFFECT Yes No RAS LICENSEE RE$p0E0 TO CHILLING EiFECT LETTER: ACTION:
- 1) Inka 2 I M M ',,s Arns ir e*5cC [QT M[eun AS ke.tj[e.Q kn6 -ochvad
- 2) o ' #
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- 3) ' /
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- 5) _
NOTES: ( A4-1
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~
r APPEN01X 4.0 ,,
$ AMPLE RECORD OF ALLEGATION PAREL DEC1510N$
$1TE: iOilI5!~t 3 PANEL ATTENDEES:
ALLEMTION NO : Al-9/-/ - C83 S Chafrman - P$ ( l DATE: 1 / 70 / (Mtg.(92345) Branch Chief - kd,.S l AthatU PRIORITY: Nigh Medium ( L ) Section Chief ( AOC) - kuham'D Ci Cl SAFETY $1GNIFICANCE: Yes No(Unknowrti Othars - Vft00AT : CONCURRENCE TO CLOSEOU1: DO BC SC astIkfbd+mbbs M ah , - i CONFIDEN',lALITY GRANTED: Yes Fo ($ee Allegation Receipt Report)
!$ ThEIR A 00L F]N0!NC: Yes / No
!$ CHILLING EFFECT LETTER WAR NTED: Yes No
/
HAS CHILLING EFFECT LETTE,R BEEN SENT: Yes No HA$LICENSEERESPONDEDTdCHILLINGEFFECTLETTER: Yes No ACTION:
- 1) M Ty r nN,[GYld'It lnc bic I f M /% o$ / \ t n $))(c cv
/
l
/
- 2) !C3eck 3) 4)
5) NOTE $: ( A4-1 l l
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- prmT-- NPCnRLb U.m2;, UP P U. t. V'W 4 e At LEGAT!DN5 AND COMPLAINTS - GLNERAt. RI 1210.1/2 3
; AFPENDlx 3.1 ; L ALLEGATION AECElPi REPOR1 c.
ce d
$ TN[ f Allegation No. ~ U3 E (leave Diana) hee- _
Address: ,_ t.doud 467 $iO4 T'ho ne : V ed M ii
- City / State / Zip: WowW u c7 W4 Ccnfidentiality:
Was it requested? Yes No j was 't initially granted? Was it finally grarted by the allegation panel Yes Yes No No [ N es a ccnfidentiality .!greer. era r.eed to be sent to alleger? Yes No
'a c a confidentiality agreeecr.t been signed? Yes No _
%v documenting why it was granted is attached? Yes No _
Alleger's Enployer: WcM nl $ 8' Position /1.tle: WoudNd 'y W facility: D'hH s'[ms 3 Dockr. No.: >L F2,3 - ( A11egatir,n Sur ary (brief description of concern (s): h h4ML 5 M d , % .k ( 1 M tfr(A dvvJ.d AOt>M 4 M ML __ M ld c t k r t cow 6 y pec$r tw ca u) h .' 7 f W6 5 b' Nu nber of Concerns: 7b Eeployee Receiving Allegation: bLN (first two Initials Tno last name) Type of Regulated Activity (a) _Y Reactor (d) ,_ St 'eguards (b) vendor (e) ._ Otner: (c) _ Materials _ (Specify) Materials License No. (if applicable): Functional Area (s): __(a) Operations _ (e) Emergency Preparedness __(b) Construction f _ (c) Safeguarcs 1((g))OnsiteHealthandSafety Offsite Health and Safety _ (d) Transportation _(h) Other: (NRC Region 1 Form 207 Revisec 10/89) i A3.1-1 g
FEB 21 '91 12:20 t4RC 111LLSTC04E OFFICE FU3
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n ., m , w, n , ; r z .n :s n ; April 17, 1991 ' FREED 0t,'i 0F INFORMATION ACT REQUEST Mr. Donnie Grimsley i Director [0 Z~A -7/-/4 2. Division of Rules and Records i' Office of Administration W Y Y 7 -- 7/ U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i
Dear Mr. Grimsley:
Pursuant to the provisions of the Freedom of Information -!' Act (5 U.S.C. 5 552) and to the Nuclear Regulatory Commission's (NRC) policies and regulations (10 C.F.R. Part 9, Subpart A), we ! request all correspondence to and from the NRC regarding ! allegations with respect to Northeast Utilities' Millstone and , Connecticut Yankee (Haddam Neck) nuclear power stations, to the extent that such information is not confidential or otherwise exempt from public disclosure. ' It is important that we obtain any available information as soon as possible. We encourage the NRC to respond promptly, within ten working days as provided by 10 C.F.R. Part 9 and the NRC's policies. We agree to pay such fees as required under 10 - C.F.R. S 9.33 pt sfdg. for the search, review, and provision of such records. If you have any questions regarding this request, I can be reached at (202) 371-5876. Sincerely, Cl l 0kd Claudia C. Guild Legal Assistant l
i SEP 151389 ' RI-87-A-0113 , 50-336 This refers to our telephone conversation today and to your previous conversation with Douglas Dempsey of my staff on September 8, 1989. During those conversations we discussed the settlement agreement which you presented to Mr. William Raymond in an allegation dated April 20, 1989 and a proposed NRC rule pertaining to Section 210 settlement agreements, which appeared in the Federal Register 54 FR 30049. You questioned whether it is permissible for a licensee or its contractors to make an offer of settlement to an employee which restricts the communication of information to the NRC. The matter was referred to the NRC legal staff fur its review and will be addressed in separate correspondence. The NRC is considering amending its regulations to clarify whether an offer to enter onto an agreement which includes such prohibitive clauses violates NRC regulations. Public comments on the proposed rule have been invited in 54 FR 30049, dated July 18, 1989 and I am enclosing a copy of the Federal Register Notice for your information. Please note that the expiration date for public comment is September 18, 1989, although comments received after this , date will be considered if it is practical to do so. If you have any additional information or queb ions, please contact me directly. Sincerely, Original Signed By: Donald R. Haverkamp, Chief Reactor Projects Section 4A Division of Reactor Projects Tel: (215) 337-5120
Enclosure:
As stated Information in this rccord was de'eted m of Information in accordance ptions - f with(the frefI Act, F0IA-exegIH W
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l g* RI-87-A-0113 0001.0.0 ff//Sh 0FFICIAL RECORD COPY 09/15/89
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%I . 68s. N. t(K M .ept n m:st April 17, 1991 FREEDOM 0F INFORMATION ACT REQUEST Mr. Donnie Crimsley Director [0.Z~A -7/-/ /, il Division of Rules and Records office of Administration U.S. Nuclear Regulatory Commission b Y~/f-7/
Washington, D.C. 20555
Dear Mr. Grimsley:
1 Pursuant to the provisions of the Freedom of Information-Act (5 U.S.C. S 552) and to the Nuclear Regulatory Commission's (NRC) policies and regulations (10 C.F.R. Part 9, Subpart A), we-request all correspondence to and from the NRC regarding allegations with respect to Northeast Utilities' Millstone and Connecticut Ya.7kee (Haddam Neck) nuclear power stations, to the extent that such information is not confidential or otherwise exempt from public disclosure. It is important that we obtain any available information as soon as possible. We encourage the NRC-to respond promptly, within ten working days as provided by 10 C.F.R. Part 9 and the NRC's policies. We agree to pay such fees as required under 10 C. F.R. S 9. 33 2e_t seg, for the search, review, and provision of such records. If you have any questions regarding this request, I can be reached at (202) 371-5876. Sincerely, aaa!.,Nat Claudia C.Uiuild Legal Assistant f/ hM
/ pe assg$8 g UMTED STATES
[ n NUCLEAR REGULATORY COMMISSION - REOlON I k, /[ 478 ALLENDALE e40AD
***** 10NO OF PRUSSIA. PENNSYLVANIA 194ot SEP 2 6 1989 (
Allegation No. RI-87-A-0113 50-336 This responds to your undated letter to Mr. E. C. Wenzinger, which was received in the Region I office on August 4, 1989. In it you contended that NRC follow-up regarding your personnel safety and procedure compliance concerns, reported in Inspection Report No. 50-336/88-13, was cursory, inadequate, and improper. Regnding the " operator in attendance" tagging issue, inspector inquiries were unable to reconcile the conflicting claims involved. We acknowledge the Stephen Scace memorandum of January 29, 1988 which you enclosed in your letter. We note that the event to which it refers occurred in a site outbuilding not subject to nuclear tagging controls at the time. No connection relevant to nuclear safety is' discernible. On going review of Millstone nuclear power plant performance continues to confirm proper nuclear safety attitudes, and controls and procedures adequate to protect the radiological health and safety of the public. As is evident in this case, persons of good will may continue to differ. At this point, we believe that further NRC review of these allegations would divert limited inspection resources from other issues of potentially greater , nuclear safety significance, and therefore be counter productive. Since your - letter presents no new facts or examples of licensee failure to control nuclear safety activities, we intend to close your allegations as unsubstantiated. The NRC will continue to monitcr licensee performance in the areas of work control and tagouts of plant systems important to nuclear safety through routine resident inspector coverage. Thank you for your concern. Sincerely, Deputy Director of Reactor Projects Division of Reactor Projects informatica in this record was deleted in accordance wit {h{Fgm of InformaUon Act,exe 'tions __ h d MMCf8- , F0lA -
L r. t CERTIFIED MAlb RETURN RECEIPT REQUESTED
~
4 E. C. Wenzinger, Chief Projects Branch 4 Division of Reactor Project c' ' ' United States Nuclear Regulatory Commission Region 1 ' 475 Allendale Road 4 ' King of Prussia, Pennsylvania )%406 ret Docket No. 50-336 * - ' L RI-87-A-Oll3 Reply to your letter of June 12, 1989
Dear Mr. Wenzinger:
The in ident in qcention, involved isoproper use of the site Administrative Control '>cedure (ACL) for electrical tagging. Additionally it involved the improper use of the site ( ACP) ovsr the control of work. , Regional Report #50-336/DPR 65(88-13) indicates that no tags existed during the removal of wiring to support the carpenters work. This report further states that work was accomplished by NNECO personnel without an approved automated work order (AWO) and an unapproved modified version of the "onerator in attendance" procedure. To date, NNECO has been unable to support it's claims of such an operator in attendance version, dnd has even attempted to Solicit and Coerce itS oWn employees to come forward and validate their contention of this unapproved modified operator in attendance procedure. Attempts to date, such as my request to Mr. William Raymond of your staff to obtain names of the personnel, has failed. The request was made by telephone to Mr. Ray d on Januar 1988 My suggnst on is that your aff contac and Mr. af Unit #2 e ctrical maintenance t verify this soliciting and coercion. I
._m_ _ _ . . . _ . _ . _ ___ . . _ . _ _ . _
t Further, let me bring to your attention a letter dated , January 29, 1988, from Stephen E. Scace to Distribution that the-administrative control procedura for tagging was clearly violated. To further support my claims of violations of electrical tagging procedure deficiencies, an ongoing investigation at Millstone 2 has before it allegations of improper use of the same administrative control procedures. Additionally, based on testimony of NNECO employees presented to your investigative team in a recent three day hearing, it appears that a substantial problem of procedural non compliance is evident at Millstone 2. It is my contention that had your resident inspectors not accepted the company (NNECO) explanation at face value, and conducted more than just a cursory look they would have found these clear violations of the final safety analysis report. I further submit that a new and proper investigation will show that NNECO's motivation for such procedural non compliance was based on cost saving measurers ra ther than nuclear or personnel safety. I hope that these specific details will shed new light, and therefore, lead to a more balanced and effective new
- investigation. -
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Enclosures / pc: Victor L .2lo Kenneth Carr i i
,_ . . _ - . _ . . , _ . . . . . . . . . . _ . _- ~ _ _ _ . _ . . . . . . , _ _ _ . . _ . . _ . _ _ . _ . _ . ... - . _ _ . . . _ _ . _ . - . . . . . ~ . . . .
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; NUCLEAR REGULATORY COMMISSION
- j "# RE GION 1 ,
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u, * ,e Doc Le t /Lir.ense : 50- 336/0PR-65 g9 g Northeast Nuclear Energy Company ATIN: Mr. Edward J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06101-0270 g Gentlemen.
Subject:
Routine Resident inspection 50-336/88-13 ($/3/88 - 6/13/88) This transmits the report of the above subject inspection of Millstone 2. The inspection findings have been discussed with Messrs. H. Haynes and J. Keenan of your staff. No violations were cited, and no reply to this letter is required. Your cooperation with us is appreciated. Sincerely, L I" Lee H. Bettenhausen, Chief Prcjects Branch No. 1 Division of Reactor Projects
Enclosures:
- l. NRC Region 1 Inspection Report 50-336/88-13
- 2. Appendis A, List cf facilities Potentially Affected by Gamma-Metrics 10 CFR 21 Report
- 3. Appendix B, Followup on Allegations Not Specific to Millstone Unit 2 cc w/ enc 1:
V. D. Romberg, Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licens.ing D. O. Nordquist, Director of Quality Services
- 5. E. Scace, Station Superintendent Public Document Room (POR)
Local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) NRC Senior Resident inspector State of Connecticut
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'l U.S. fNCLEAR Fi3'JLATCRY CCMi'.1551CN REGION 1 ( ,
Report No. 50-336/93-13 Docket No. 50-336 License No. OPR-65 Licensee: Northeast _ Nuclear Energy C: rany .t' E0. Box 270 ' hrifore, CT 0610'. 0273 Facility f;ar.e: fiillstone Nuclear P:wer Station. Waterford, Connecticut 8 Irs:2: tion At: Milisteet Urit 2 ,
. Ca.es: May 3 - Jure 13, 19'5 Ins c.: tors: Peter J. Habignorst, Resident inspector David Jaffe, Licerstr; Freject Manager, NRR Wtilian J. Rayrond. Senior Resident inspector James Trapp, Recctor Engineer, Division of Reactor Saf(ty Repe-tiq Inspe:t:r: Peter J. Habighorst, Resident Ir spector 4 pre.ed by: TM Mg Caoe% vgf rejects Se:tton IB
$ , Lnief, sea:t:r Cate les e:t'en Su- a y; 5/3 - 6/13/2B (Racort No. 50-336/88-13) -
A-eas Irsre ted: Routine NRC resident, region-based, and specialist-inspe: tion of
~
plant operations, surveillance, raintenan:e, radiation protection, physical secur-ity, cutage activities, allegations, Licensee Event Reports (LERs), Safety Issue Management System (SIMS) items, anc :or.mittee ' activities. Resul ts: No unsafe conditions were identified. Additional follow-up-is warranted on a 10 CFR 21 Report concerning wide range nuclear instrumentation susceptibility to moisture intrusion (Detail 4.6) and allegations (Detail 8.4, and Appendix B).
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t T23LE OF CO?iiENTS i PAGE ' ' l.0 Persons Contacted.................................................... 1 2.0 Su tary of Facility Activities....................................... 1 3.0 Previously Identified Items............... *
.......................... 1 3.1 (Closed) Violation 87-16-01: Fire Protection f Feed-a t e r I so la t ion Va l ve s . . . . . . . . . . . . . . . . . . o r Aux ili a ry
.................. 1 3.2 (Closed) LHresolved Item !!-06-02: Control Of Overtime Ouring Outage A:tivities....... ............ ....................... 2 3.3 (Closed) Insee: tor Follen up Item: Location Repair of-Boric i l.:id Evildup ins'ce Containeert
.. ....... .......... ......... 2
'0.
P . a n t T ours a r.d Ope ra t iona l Sta tus Reviews. . . . . . . . . . . . . . . . . . . . . . . . . . 3 41 Safety Syste.m Operabil'ty.. . . 4.2 .................................. 3 4: Feactor Vessel "0"-ring failures and Corrective Actions......... Piant Shutdown Due To An Inoperable Control Element Drive 3 4.4 ;e.Pew r : r. ofaplant n i s In: m ( 1cen: C E 0 M ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5. . . . . . . . . . . Reports...................... 6 4.5 kre s sua izer Level Control Channel fa ilure. . . . . . . . . . .. .. .. .. ..... .. .. . . .6. . 4.6 4.7 Wice Ran e Nv: lear Instrutentation, 10 CFR 21 Report... 7 L i c e n s e e s D i s r o s i t i o n o f C ECM a l 4 . . . . . . . . . . .. .. .. ... .. .. .. .. . . .9. . . . . . . . . 5.0 Pry s i:a l Se c u-i ty . . . . . . . , , . . . . . . .
... .............................. 10 <-
f.0 Su veillan:e.... . . ..... . .... ....
.............................. 10 7.0 Maintenan:e. ....... .......
........................................ 11 7.' Eoric A:id Bulldup en Reactor Cuolant 7.2 Nozzles................... 11 #
Boric Studs.Acid Corrosion and Licensee Evaluation of Reactor Vessel
....................................................... 12 3.0 Allegations.......................................................... 12 l 8.. 88-A-0015, " Health Physics Ccncerns Inside Containment"..
l 8.2 12 Tel ephone Call From a Concerned Ci tizen. . . . . . . . . . . . . . . . . . . ..... .. .. .. 13 i 8.3 8.4 88-A-0029, " Update on Procedural Compliance In Metrology Lab". . . 13 88-A-0040, " Update on -Improper Radiation Monitor Calibration and Other Concerns"...............................
-8.5 . ...... 14 87-A-0113, " Con t ra c to r _ Work- Ac ti vi t i e s" . . . . . . . . . . . . . . . . . . . .... .. .. . ~18 S.0 Licensee Reevaluation of Nuclear Complaints and Employee Concerns.... 20 4
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_._ =_ _._ __. _ ._ _. _ ._. _. _ _ _ ._ _ _ ___ __ _ _ .. _ _ _ _ _ _.__ _ .. _ _. _ _ Table cf Center.ts . L').f t
. . . . . . . . 20 Items...............
. 10.0 Safety Issses Managemen: System (SIMS)
" Reactor Coolant Pump Operation".. .. 20 ,
10.1 License Amendn.ent No.116 " Spent Fuel Poo) Tenperature". . . . . . . 21 10.2 License Amendrent No.114 < 22 11.0 Ce=,ittee Activities................................................. 22
- 12. 0 L i : e n s e e Ev e n t R e po r t s ( LE D s ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
22 13.0 Periccic and Special Reports...............-.......................... 23 14.0 Pana'gament Meeting............... t A 0;:e":i x A. Li st of Fa:ilitie s Pctantially A'f ected by Gar--a-t'etric s 10 ~~R 21 Report A:rer:tr B, ~ollow up cn Allegations Kot Spe:ific to Millstene Unit 2 i I b l T. i l' 1 l
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1 t torf':01Y B , FO'.tCt 'JT' 0F OiFER ALLESt*10NS HDT SPECIFIC TO ' PitL510'.E UNil 2
}. EE-A-0035._ Erpired Fire Watch Ovalifications :
On January 13, }cEE a contractor employee approached the inspector concerning the fire watch program. The alieger stated two specific concerns: *
-- Individuals who serve as fire watches have expired quall'ications. .
-- Jn one case a fire watch did not have security access to all arets to me in te was ass 4gned. .
i i ++ ' r s te::: r teviesec the t sitir{ fire watch progran on February 1, 2, and 10 ir ren; se t t%e allege r's c;rcerns. The prograr. adeovately add *tssed - tra ; ss at: resstesibiitties cf fire wat:h personr.el (roving, and continuous v.t t:ne s ) . :n rev+ex, rar.:cr. checi.5 of fire watch qualifica*. ion (dates and
+
es p .-a tit r.) inte.sh:ut the Millstone Station were conducted, No cases of e *1rs: # ire watch cualifications were found. The alleger cid not recentact ine irsre:t: as was p1. net, so no a:ditional information er feedi:ck was rec,1ce:. This allegation was unsubstantiated by inspe: tor review.
~'t ir!;c:.t- rarec-ly sele:tec assig ed firewatches to review security acesss a,t*t .:st ;n ar,c co cared it to arets covered by the fire watch. No irace-0.a:ies were noted, In cen: 1usser, the inspe:.:r's review resulted in an unsubstantiated alleg6- ,
tien; note er, inis r.atter was referred to the licensee' f or appropriate con-siceration. This ratter is considered closed without more specific informa-t tien et cualification inade:uacies being needed.
- 2. i *ut .
- a d'er Pa*6e as F.adiet:tive-At 11:15 a rm , on Ju*.e 10, a citi:en f rom Niantic, Connecticut called the insoe:ter concerning radioactive rarkings on a truck trailer located .in down-town New Lonoon.
At 11:30 a.m. , the inspector re;uested assistance f rom .the licensee's health p"ysics ergar12atien to investicate the alleger's concern, : The . licensee agree; to support the inspector with a-cualified radiological technician and appropriate instrueentation, The inspector, with assistance from the licensee, evaluated the internal surf aces of the truck trailer for contamination and I adiation levels and found no measur. ole indication. The _ inspector returne: t he call to the alleger on Juna 13, to .nform him of the results. The alleger j stated that he was satisfied as I had-no further questions. This item is l CICsew. I I i l l-p
-- - - , . . _ _ _ . _ . _ . . _ _ _ _ . - . , _ . . _ _ . . . . _ _ . . . . _ ~ . . _ _ _ _ _ _ . . _ _ . . _ _ . _ . _ ._. _ _ .-
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- 3. F' *-D: 3, !: tritter Ve el Activities iris se:tien add-esses tne issues raised by c forcer contractor empi:yee (alleger) ree;rding the control of work activities by a contractor at the 5.i'.e (Contractor A). Tne allecer's employment at the site ended in the Fall of '
19E7. The alleger subsequently contacted the NRC by phone in the Fall of 1987 to dis:vss concerns about electrical tagging (see item B.3.6) and security lignting (see item B.3.3). The alieger also corresponded with the licensee ir, the Fall of 19E7 to discuss concerns about electrical tagging. That cen-cern invcived two instar.ces during work in an onsite warehouse that was tting converie: to Unit 2 naintenance offices. The alleger rersested a r.eeting with the li:ensee. ine alleger contacted the NR: by phone in December 15E7 to set up a reeting. ine is sues c~15:vssed teiow were receiveo during a January 13,19Ei rece g with ite alleger. inese issues were referrec to licensee ranagerect fee a r.- cessss)- i*e l':ensee h:c initiate: action on the the alleger's cet:t ens pri;r t: LC," involveeent and, tesec on av:it findings, had initiatto :orre:- ti.t a : t '. :m s . The status if a:tiens anc licenste findings on all 1:.aes were t erie:1cally dis:ussed with the inspecto" by the licensee, and were sur. ar'.U.d in an Mtj 19, 1955 n,ererandue to the Station Services Superintentent. SU;- t it stta' inf errition on tne original issues was cb'.ained during various fol-ica vo te' son:re ;onversations anc during a follow-up inspector retting wit 5 t*' allt;t- ct June E, 25EE. Tne licensee conta:tec the alleger to set uo a r:s t ;. w:a:n was seteculto af ter the end of the inspe: tion peric:. 3.1. lisse- Installatten of 16 Pole Ligris in the Simulator Evilding Tartirg L:t in July 1957 ,
~ 's issue concerned the acccLa:y of itehting installed without ar.: hor bait teit l a,es, in the sirulator buildin; parting lot.
L etste review found that installation of the mounting bolts without a terpiate was a:ceptable. - The Contr.ctor A Superintendent, a :trtified
; o'essional engineer - civil discipline, used engineering ]v:;ement to c' rett irstallation of the mounting bolts. A template was not recaired tased on discussions of the installation with the light manuf acturer, who agreec that installation of 4 bolts on a 10 inch bolting circle was acceptable. ine licensee concluded that no further actions were required and that no corrective actions were warranted. The inspector coserved ite :ariing ict lights and noted no obvious ir. adequacies witn tr.e in-staliation.
The licensee stated that the alleger was terminated because cf the dis-pute the Contractor A Superintendent over the use of the template and after removing the mounting bolts before the cement was poured. The alleger was reinstated af ter discussions with Contractor A upcer retage-rent. The licensee considered this to be an internal Contracter r.atter ; at: cet:1ucec nc further acticn was recuired. l
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1r.e inste:t:r rcted that this issue ard its resolutien had rs '- act on
- nuclear saft.y. Tre inst e: tor idtntified no it.a::ecuacits in t*t t , r , ;-
tion cf this issue.
3.2 1ssue
Coordination of Work Activities to Remove Temporary Security Lighting This issue concerned alleged poor control of work activities as evidenced by a job where two contractor groups were assigned to do the sane work. Licensee revien deterrined that tu: onsite contractor groups were brth assigned to work on te.porary security lighting adjacent to a warehouse onsite. The job was first assigned to Contractor B, who usually works lighting Jcbs. The start cf wo L by this group was delayed. The alle-ger's ertleyer, Contractor A, was then assigned to do the work by the 5 s tion Services Enginetring Department at the recuest of Security, k nik rrtoaring to co tr e work, Certractor A perscr.nel ncted that Con-t a::t- E was coing the jet; the work croer to Contractor A was then can:ellec.
'. r+ p r c b b r o f w:* l cc: :1r.ation origir.ated within the Security Deoart-rv i Licensee actiors wire a::ressed in a memorancut from the $tation 5t- 'ces iuperintendent to the Security Supervisor dated February 1,19EB, c"; f r0- M e Ovality Se vices 5sttrvisor to the Station Service $vper4n-it
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.v:>r A - tssm;m tm-i n -fu.1 e yn
' The inspe: tor noted this issue and its resolution had no impact on nuc-lear safety, The adecuacy of lighting in the protected area has been previtssly revi?we by P.? LE an:: noted discrecancies were rt;,;ived , (fM Ee; ion 1 Ir.spectier Repo t 50-336/67-20). No inadequacies were identifiec in the resolution of this ite n. 2.3 Is ssej korker Whole tery Counts and Termir.ation Exposure Repcrts 1515 w:.s a new ailegatic* raised by the alleger on April 25, HEO. This ite" involved two indivicuals wno worked at the Millstone Station in the 39E5- SS4 time period. Bcth workers repnrtedly stated to the alleger tnat they terninated erployment at the site without obtaining an exit whole body count, and without getting termination exposure reports from the licensee. The inspteter asked the alleger to provide the r.6me and acdress of b:th incividsels to allow f urther followup. Af ter ccnferrt) with both workers, the alleger identified Individual A, who wore a respirator while onsite and who also had a known cesium uptake. No information was providec on the second individual. Individual A re-portedly did not want to talk directly with the NRC and no address was provided. This item was turned over to the licensee. for review and ; evaluation.
ipperdix E 4 The licensee provided, for MC revie , exposure bistory files er in:1- l sic.;al A. These sho.,ed that te wor 6ed at the site periodically f ro r 1974 L t.ntii :'E3, with the last wort date onsite being 8/15/53. The licensee ' stated that Individual A wor 6ed at the station during outages on Units 1 and 2 and produced termination exposure reports corresponding to each s work period. The licensee noted that the 7 termination reports were . addressed to 5 dif ferent addresses used by the worker and noted it was i possible the last report was not received at the last address on file ' or forwarded in the mail. The licensee stated that another copy of the 0:teber 1983 termination egosure re; ort would be provided if rec,uested l by Indivieval A.
- The inspe: tor reviewed the last exposure report dated 10/}0/E3 and cover-ing the peric:t f rom 5/13/83 to B/15/83. The inspector noted that the re:orced cuarterly espesures for all work periods were low and well aithin reg.latory limits. The inspector noted further that a comritted
', GC Forr 4 cated 5/33/53 was on file and properly reflected the er:osure history record. The inspe: tor also noted that the licenste's bealth thy;tcs recores show that ]ndividual A properly completed the prereQui-site t ea tring, r.ecital screenings, and respirator fit te sting rieetsd to wee respirators f or w:rk in radiolcgical areas at Millstone.
Lice see records shcw the resuhs c' whole body counts (WBCs) trformed for 2*. civic 41 A. The last WEC result on file was performed on 5/12/53, S M e t. was t>e entrance covet usually done by the liceasee as a ratter . cf relity (a ed r:t be:ause of an MC requirement). Tne licensee's re-tores cc* fir ec that ]ncivicLel A lef t the last work assignment at the stition without a termination VBC. The licensee concluded, f ro'r his res te. of the exposure history and work activities, that a terrination WE: was not neece: to meet regulatory requirerents based on records 15at I'N no airborne exposure time (MEC-hours) were accumulated by the it.ci-; vic.al. 'nis item is discussed further below. he inspector noted that radioactive potassium and Cesiuv137 were re- : pcried in the WBC results for 7/23/E0 and 3/7/81. R6dioactive rotessium I is natu-ally occurring and is fcund in all people. Cesium-137 is not i l~ " raturally occurring and is produced by nuclear. fission. The inspector noted that the cesium levels recorded in 1980 and 29S1 were 0.315 nano- , curies (NCi) (+) or (-) 3.092 NCi (at two standard deviations) in 3950; -; anc 1.792 (*) or (-) 2.437 NCi in 1981. At these levels, the isotore was present in trace amounts, just at the limits of detectability and f ar below the action level f or follow-up investigation. The inspe: tor l noted that subsequent WBC results recorded on 11/6/81, 3/26/82 and i 5/23/23 did not show any cesium. The inspector did not determine the - source of the uptaka. The inspector noted that the licensee has no re-cord of an incident report involving Individual A. Thus, the cesium , uptate does not appear to.be relattu to work at Millstone,
'he licensee provided for inspector review all radiation work permits .
( U s) use: by Individual A for work in the Unit I and Unit 2 radie-Icptally contr:lled areas during the period from May 23 - August 25, i lL L
i - 1 ;;,e n c i r E 5 t Tr,e ir:,;e: tor revie ed the rc;trts alo+g with the healin ;*ysics lM:
,=r ty rest'*:, sro.eing the radiciogical conditions ir. the tort artas cf irterest. ine survey results for airberne actisity were revie.e: in t.artituiar, as reccrded in Air Activity Logs for dates corresponding to.
the RlPs. The surveys showed that airborne radiological conditions did not reovire the use of respirators because of pre-existing concentratior.s 'l a t t he wo r k s i a.e . Air activity results were at or below the 3.0E-9 v '/cc MPC lirit specified in 10 CFR 20 for unidentified isotopes in
- restricted areas.
ite inspe:ter r.cted f rc,m the RWP records that trost work activity by In-c'vicual A in the Millstone 2 containu nt involved walldowns, inspectio 5, a*,o installaticn of cables and conduits for neutron detectors and re-L i s t a n:e- tempe ra turt -de tt c tors . Sofre of the Rv!Ps did require the use cf itsoiratcrs arc Indivicual A indicated respirators were used. The ue :1 res;',rators ir, th+ se tr'.tances appears to have been a precauti:n 1: t rcie:t a;ainst Ecssitie irgstion of radicacthe material ecde air-t:rre csring tra c urse cf wori su h &$ during core drilling activities c- ::s t a i n'r e r t w c ' l s w i t h general area contamination levels ir. the range c: !. 0 6 . C; r cisirte;raticns pe minuts (dpm) per 100 sq. cm. Eased on t e : ciegical cen:;itiens at tre job sites werted by Indivic.al A, and c:asicer;*; that liter as re:cres show that no MPC-brs were ret: 'ded for
- 5;;.al A, the instetter concluot: that a whole body ccunt u s not n re: by MC regulati:ns. $;ecifically, no whole bo0y cou t was reaist y as ;a cf the ticassa) assessments required by 10 CFR 2: ; O.( a )( 2 ) .
%e ins;e: tor revie ec 5ta: on Frecedure SHP 4970, Interr.al Exposure
;:' *.re' (Eica ssays), ;evision a cai,ed 4/22/85, which establist es the ,
I' :t ra te 's t ,n ssay t r:yar and st-ts the criteria under w>iich UE05 will De ;trf:eced. SHP 49;' s reovirts a WE; if it is cetermined that a lirit cf 40 (ffe:tive MPC-hou-s is exceeded for a worker. This is consisttnt witn 30 CFE 20. /saditier,:.lly, 54P 4907 requires " routine" W30s for all ot-scerel issued d:sirt:try at the site at the sttrt and end of errployee*t, an: at least ontt per year. i*ese whole body counts are (in part) a s;reening r:asuretrent used t:. validate the adequacy of other controls establishe: for work in radicicgical areas. The perf ormance of such VE:s is a licensee ad .inistrative practice that exceeds NRC requirenents. Essed on the above, even though SHP 4907 was not tr.et in this instance, there is no safety significance and there is no violation of NC require-r vnts. The li:ensee stated he would perform a WBC on Individual A if re recue:tec ene. The alleger was reauested to relay this inf ort..ation to ]ndivicual A. Tne licensee stated that, since the reported cesium contamination was below their investigation threshold of 20 nanocuries, no further follow-uo would have been taken when it was first noted. The licensee stated that all contractors are made aware at the start of employmer.t that it is espected w:rkers will get a whole bo0y count upon termination. There is no rechanisr. in place to enforce inis requirement in all cases. The
Appendia B 6 licensee estiftated that about $ or or fewer workers leave the site witt..ut and exit VEC. The licensee feels this is acceptable sirce, ts an assesst.ent tool, the 95', of the workers who co get the termination whole body counts confirtr the success of the respiratory protection pro-gram. Ite licensee stated that all persons are evaluated per 10 CFR 20.103(a)(3) hen required. The inspector reviewed the inspection secord for all three Millstone units and noted that recent NRC Region I reviews 4 of the internal erposure controls have found the program to be acceptsble. Minor deficiencies have been noted, but no inadequacies have been iden-11fied in the respiratory protection and bicassay assessment function. Du*.ng a June 8 followup rneeting with the inspector, the alleger proviced addltional inferr.ation about Individual A. The alleger stated tnts, curing Individual A's last work assignrnent at the site, he was wore.ing in contairrent without a respirator when plant operators turned on a f an wnict , a
'; caused centart.inatier. to be blown on the workers. The alleger had feat: fvther inf ortration f rcr In::ividual A as to whether Indivicual A was
;ositivetocounts hase ccr.;&nir.ation or, a rasal seter. or, his clothing or skin or whether he snowed s W
en 5 retter was again re'/iewed with station HP ptrsonnel, Sho stated that incicett riccri woulc have been written for any instance involving 56 in contaninttion or a pessible intake of radioactive material, No
'r:ictnt reports involvir; Irdivicual A are on file. The inspector could
- t. :
c' rec;1y ;.ursue f re-this Ir civio.;&1rnatter fviner4 without accitional specific informaticn The inspector sent messages to ]ndividu.1 A via toe alieger. In ividual A had not cottacted the inspecto as of Jure } 3,195E. Essed on the above, the inspector concluded no further
- f. .: actions are w&rrantcd on this ratter.
- 33. Is:ve: Category 1 Weldir; by an Unqualified Welder ; [
This issue involved a concern that Contractor A, who ersployed the Elleger, was using an electrician to perform Category I welding insice the con-tainment. No et .er specific information $ 3s provided as to the name o' t int affected. wcrter, the plant insolved, or the name er type of plant systens ~ Unit 2 was assured-to be the affected unit since that unit was in an outage at the time the information was provided. The inspector reviewed welding activities in A. by-Contractor the Unit 2 containtrent on January 23, ISEB for None was identified. to the licensee for action. Subsequent routineThe issue was referred inspections of Unit 3
-thd Unit 2 outage activities did not identify welding by Contractor A.
Licersee follow-up identified infortnation which partially corroborated the alleged concerns, but did not show safety significance. Contractor A does not htve a Quality- Assurance program and is not used by the licensee to perform ruclear safety-related worL The licensee devtloped a list of work assigned to Contractor A at all three units i since Octcber 1956. Until recently, most wort by Contractor A was per-' , forced for the Station Services Engineering (SSE) Departnent, whose re-t
.._..,__,s,___ ,.i,,,, , . _ _ _ , , . , _ _ . .,.,._,,,,,,,_,,_,,,,..,___.,,.,,____..,.._.._--y ,,
_ . ~ - - _ _ - . - - . . - - . - - _ - _ -
/;pendix E 7 spoesibilities intit.Se non-ruclear projects such as grounds et i r.t e n a n c e , I.
ard naintenance, installatier, and modification of outlying buildings. C(ntractor A was also used in the May - August 1957 period to install and modify gates f or high radiation areas (HRAs) in all three Units, as
- discussed further below. During the 1987 - 1988 outages, Contractor A' was used as a general laborer force under plant personnel supervision to (i) move shield blocks around reactor components - Unit 2 contain-ment; and (ii) perform naterial accountability control in reactor work areas - Unit 3 containnent. No Category I work has been assigned to Ccntractor A and no saf ety class weiding has been performed by its ei-
;1oyees.
Contractor A carpenters and electricians have welded in non-Category I work activities. Qualifications f or electricians were based on welding training obtained during union apprenticeship. Also, Contractor A car-penters welded stif feners on HEA gates to provide added reinforcement, wtet reeced. That wor 6 was corpleted within the plant buildings anc was controlled by work c ree rs. One Centra: tor A electrician performed tro ron-Cate;ory I weids outside plant buildings: one job involved relocating a secuity ir4rusion cetection system; the second involved mounting a ritronave anterna on the s .oe of the condensate polishing f acility. -The li:ensee_concluced that completion of these welding activities in ac-corcance with general construction work practices was acceptable. I,o fu *her NRC actions are cee eo warranted or are plant.ed on this matter.
-re inspector revieaec station ac-inistrative recuirements to determine
.. mt. t h e r a ny c e nt r t '. 5 snov'c have been applied to the work. Th's review inclucea specific consiceration cf ACP-0A-2.03A, Non-Category I welding.
Tr e 1 certified welcing activities did r.ot involve ASME or AN51 code work; tr e w:n coulc rot af f ect Category 1 systems; and the work did not in- ,
.cive an activity for which a completed weld documentation package would .!
De recuired o* desired. Based on the above, the requirements of ACP-QA-2.0% we re ceerec not at:11 cable to welding Contractor A performed. No in ocecuacies were identifiec. The inspector reviewee the licensee's actions on this issue and identi-fied no inadecuacies. Inspector review of HRA gates during routine in-spections in all three units has found the gates to be sufficiently strong to provide an ace:;uate barrier. No inadequate conditions were 1 certified.
3.5. Issue
High Radiation Area Gate Modifications This issue concerned the controls applied to work done to modify and install alarms on about 70 high radiation area gates in all three units. The alleger wa: responsible for completing the work in the March-August 1987 time period. The alleger stated that an inexperienced engineering te:hnician, newly hirec by
- he contractor, was assigned responsibilit, for the job. That individual's cesigns were reportecly aborted af ter atout two weeks work anc af ter wcrking on two gates. The job was al-
~ _ _ _ .__ _ _ _ _ _ _ _ _ _ _ _ _ .. _ .._. m ._ _ ..__. _ ___ _ _ .
l Appen!i> B 8 legecly perfer ed based on "blactboard" cesigns withcut writic, guicance \, and criteria. Probler5 reportedly incluced lights that were r:t needed and a 50,000 restock charge when another type of light was selected, and the use of alarm bells that ran on 24 vac and required the use of step ' down transforrters. This item was referred to the licensee for revfew. The inspector noted that the High Radiation Area (HRA) gates and associ- , ated alarms are not nuclear safety-related systems or components. The - inspector asked the licensee to specifically address whether installaticti k work associated with the design not used resulted in radiation'ev.posures which could have ' een avoided if better guidance had been proviced. The inspector reviewed the results of the licensee's review of this issue, independer.tly reviewed the detaileo design change package under which ; the work was accomplished, and interviewed the individual responsible
, fcr cortpletion of the modifications as the designated plant project engineer (PE).
The wert involved int redificaticn of Esisting gctes and the R:tallatio* L' - E ' ;a te !, that provided barriers to the entrante to desigra.tc high ra::tation areas in the three Millstone units. The werk was controlled witn cttailed guicance provided in Anomated Work Orders and in tne fol-locir.g plant cesign change requests (PD Rs): F;;E 3-105-86, HEA Gate Alart9s & Warning lights (MP1) f 2:E 2-0*hE7, HRA Gate Alares & Warnirg Lights (MD2) _ F';;4 V03-E7-002, HRA Gate Alarms & Warning Lights FDOR "03-86-372, HRA Wire Mesh Gates I' : licensee is recuired to control access to high radiation areas 'as cef tred by 10 CFR 20 are to lock the access ways if sne cose r:tes in- < volve raciaticn levels in excess of 1 Rem.'hr. The main purpose of the P:G v.c 5 :: acd warning lights and audible ala*ms to existing ar.d new gates te notify pers:nnel that the gates were not properly secured (IccLec closed) after passing through the gates. This action was in resoonse to licensee and NR:-identified concerns that the HRA. gates were being left open following access to the rooms. The PD R was also used to add gates to new areas based on surveys by health physics personnel, and to stiffen existing gates. In addition to the guidance provided by the PD Rs, the following references also provided written guidants on gate f abrication, modification and installation: l Stone & Webster specification 2199.241-932, Specification for Wire Mesh Doors tN5:0 Electrical Installation Specification SP-EE-076 Field sketches LPKA.120286 and LPU-B 120286 Various /.W0s for installation activities during Pay-August 3957
-V --P--W e-rvu--igv'- e.d 4y mdC----e-w=-P-re_-**5'd '*fM "- **
**et* f'
5
/p;endix o 9
( lhe design chan'ges described in the FZRs were prepared by an ergineering teChr.ician in tte Station Services Engineering (SSE) Departreent and were revientd and approved by the licensee's engineering staf f for each unit , as recuired by station administrative procedures. The review by the unit staf f an:. Plant Operation Review Corr.mittees determined that the change, while intended to improve compliance with the requirements of 10 CFR 20,203(c)(2) and Technical Specification 6.12.1, did not involve an un-reviewed safety question per 10 CFR 50.59 or adversely affect the opera-tion of safety systems or structures. Power for the ci cuits would 15e provided from non-class lE supplies and the gates would not impact seis- I mic walls. The f unction of the alarm circuit was to provide audible and I visual indication that the HRA gate was open longer than 10 seconds (the tire delay allowed for normal transit without alarms). An override switch was provices to bycass the alarrn f unction for periods when the c $te wu id be lef t open for estended periods to allow r.cVerent of mate-r:415. The ( arm would also activate, however, if the f unction switch was r:t retu-rdc tt ttt "at,to" positior. when the gate was closed. The lictnsee dete mined that the original PDCR design was develeped in late 1956 wie tne intention of using explosion proof alarm lir, hts simi-lar to those already installed on existing gates. The esplosien proof lights were orderec from funds available in the 1956 operating bucget. The inititi circuit design was developed to_use, to the extent possible, conoonerts alreacy available in station stores, including transforcers, T.', VA: ala-tr telis, and alu-inum stield ( ALS) cable. The ALS was chosen to riniri:e w:rner radiatier exposure since it would allow installation c' tte t rcus t without conduits. Additional bells and relays were or-cered as necessary. As the cesign change proceeded, it was concluded it,at the est.losion proof lights were not need:d and strobe lighu were . ~ cretrec instead. Ttis acticn was taken even though there was a $6294 restock charge on the explosion pro 0f lights, because there wts still a net savings in e ces s of $10,000. Since the explosion proof lights were never instelled, inere was n; additional exposure required for the Jcb as a result of the enange. Tre w:.rt m originally assigned to Contractor B. The licensee deter-I mined that the work was proceeding too slowly and the Station Services l Crgineering Department reassigned the job to Contractor A. i The licensee determined that the original circuit design as described l t in the approvec PDORs would work. Proper functioning was demonstrated i by censtruction and bench testing of a test circuit in the shop. However, l contract electricians (including the alleger) recognized that irprove-ments in the circuit design would reduce the number of cable terminations needec and would result in less ALS cable being installed (estimated at 3 to 12 feet per gate) and thus reduce the work time required in radi-ation areas. Even though the gates controlled access to HRAs, the typi-cal wori area for the gates was the_ immediate area of the cate arc the neards electrical panel, which were not in "high raciation areat." (Personrel exposure required to do the job is dist;ussed further below.) i The n.; 1fied circuit in its final form included the use of an additientl I
- _ - . _ . - _ = ~ - .- - - -_- -- - -___.- ._ _ _ .
j A; pers'.x B 10 instantaneevi relay to repla:e one contact f ro'r the gate closurt limit k switch proposed ir, the criginal cesign. Inspector intervieves with the project engineer determined that he proposed several interim configura-tions that were found unacceptable during follow-up revievs with the alleger. Ultimately, the final circuit design chosen was the one pro-posed by the alleger. The licensee stated that the selection of the ' ! f.nal design was left to the discretion of the Station Services Engi- , neering Department, since unit engineering had determined that circuit l changes were conside-ed minor in scope and would not impact the conclu-sion of the 10 CFR 59.59 safety evaluation. , The licensee reviewed the radiation exposures for the job and concluded they were not excessive. No esposure was incurred on Unit 3. The work cn Unit 1 experted D.Eib r.an-rem for 264 man hours, for an average dose rate of 3.1 m*sen/hr. The work on Unit 2 expended 0.710 man-rem for 236
' man hours, for an ave-age dose rate of 3.0 mrem /hr. A tabul u icn of incividual espesure s f or all contractor personnel v.ho worked on the ,4c3
(.c.ich incluced tiposures for all work during the peried and net just it.e hF.A gate job) 'W(d perst f il esposures were not e .cessive. In-snectt- -e iew of the tabulation noted that inoividual quarterly txpo-su es v.e-e less nan IR c'.e- % all cases, except f or one incividual weith a maxime, cuarterly erposure of H0 mrem. The licensee concluded, based on radiation wor) permit (RWP) records for the installation of the first few gates, that espesures were not e>cessive. Further, the licen-see's ac-i*istrative tiposure limits we e not increased for any worker cu-in; the jtt. ins;ec ur revie, founo no inadequacy in the licensee's C C n " '. U s i o n . The '.1censee further corcluded that, in spite of the additional duse ttvings reaii:ec in going f rom the original to the final circuit cesigr, reascnable reasures were taken to minimize radiatien exposures for the ' design change. Trase reasures included use of ALS cable instead of con ' duit 5 to reduce installation time, selection of power supplies to mini-ti2e time spent in radiation areas, testing the preliminary cesign in l the snop and prefabricating and testing materials as much as possible j in the shop to minimize installation. time in radiation areas, and licen- ! see supervisory monitoring of work progress and reassigning the job to another contractor when the first contractor was deemed unacceptably slow The inspector identified no inadequacies in the licensee's finoings. The inspector noter tnere was a difference between the licensee's i.nd the alleger's version of the job. The alleger stated that the critical circuit design did not work and modifications were required on the first two gates modified in the plant, k' nile the inspector did not resolve - the different versionst he did note, that based on the low dose expended for the entire job, rework of two gates in the field would not change the conclusion that exposures were not excessive, n - . - - - - . - . - - - - . - - - . _ . - _ - . _ . - - - . - _ - -. - -_ - -.-
A;;endia B !! Licensee review concluded that the engineering technician assigned as project engineer to the job had adequate experience to perfort. the work, s Tne technician is a contractor personnel employed by Contractor A and ' assigned to a staff position in the SSE Department. The engineering ' technician worked for 3 years during Unit 3 construction and startup'for the architect engineer. The technician gained experience in the elec-trical discipline while working with major electrical components and through involvement in switchgear testing. The technician also worked for 1.5 years in the SSE Department working on projects involving t,he electrical discipline. The licensee did note that the technician had some dif ficulty administratively coordinating the setups necessary for the first HRA gate job in Unit 1, which involved allowing suf ficient lead tire to process tagouts and R'r.'Ps so t'. tat the work could start on time. This difficulty stenced from his lack of erperience in processing the acministrative cont rols. The licensee concluded this inexperience did rot cause unnecessary radiatic' esposure. The inspector identified no inactauacy with the licensee's conclusien since delays in starting work , er in obtaininr tne prerequisitt 16;out would not result in rac.ation ex:osure. H o..e v e r , ir. addition to the above described dif ficulties with the licen-sae's a:tirti strative controls, tht inspector noteti that other prcblems cccurred in fcilowing the recuirements of the tagging procedure, ACP 7.000, as ciscussed furtner in Issue 6 below. While the tagging for the W gate jcos was found to be cone safely, it was not cone in full com-Li san:e w :th the ACP, as f olic+s: (i) operator-in-attencance tag;ing wcs pe ferred by cor. tractor electricians on some occasions, which did not- ' r4 st the recuirements of tagging procedure ACp-QA-2.06A; and, (ii) single tegs used for N1tiple gates (7 primary breakers covered 5 gates each) were ort:essed without using the SF 210 continuation form required by : the ACp. Further, during ir.terviews with the contractor technician, the ' inspector oc',ec thct the technician had been responsible for the initial pet;aration of the 30 CFR 50.59 safety evaluation for the PDCRs, 5ut had not uc trainto; in the 10 CFR 50.59 process. The inspector found that the cont-actor had be: cme f a ,iliar with the administrative requirements by reading the associated aaministrative procedures. In response to inspector inouiry, the licensee stated that contractor personnel are proviced on-the-job training in the station administrative recuirement as needed, and that this training involved reading of the associated administrative procedures. That was not formally documented. Although the inspector identified no -safety issue relative to the HRA job, and no inadecuacies were identified in the-completion of the-FDCR-per the requirements'of NEO 3.03, the inspector identified this area as meriting further NRC review to determine the general adequacy of training provided to contractor personnel on licensee administrative requirements. Even though no. safety-related (Category I) work is assigned to the SSE Departr.ent, the inspettor expressed the concern that the licensee needs to f ormally cocurer,t required training to assure contractor person *si - - ..--~ - .....-. - - _ . -.-- - - . _ _ . - . - . _ . _ . ._ . , - - . . ~ - ,
l l ip;endia E }2 L are fully 'ariliar v.ith all admir.istrative pro:edures and require ents ttey are e,rected ic fcilos. This area will be reviewed f urther on a subsecuent routine inspection as a potential element of licensee per-formance. ' In sum .ary, while followup of this issue did confirrr the alleger's statement in part, the licensee concluded tht.t guidance was provided anc controls were applied suitable to the activity. Further, rea sonatrie ef forts were talen to maintain radiation exposure hours less than the ALARA goal of 1 Rem per unit. The inspector agreed with the licens,ee's - conclusions. The instector noted that it is neither unacceptab".e nor unusual for approve 0 cesigns to change to reflect imoreverents icentifie:
', in the intert: tion between design and implementing groups during the dt s gn charge pro:ess. ta unacceptable conditions were identified during the ir f f r: tor's review.
3.0. 1:t o: I:*tren:e 1: Cc trols for Electrical Switchirg Part A: E,ercising 4E0 volt breciers without tags or AW3. This issue cot: erne: alleged actio'is, on one occa sion in 1987, by ttt C:rtra:t:e A Su;eri*ttteent to ranipulate 450 velt cir:vit breakers on a vi*enoese canel in orcer to trNble-shoot a prcbles with the air con-citioring. inis a:ticn ..as taken weitnout a
- eet order or tags to contro at 6:tivity.
- r _? : Movirg 5 wirs teus e (O c :t ri:a l ci rc ui t . ,
C a secc c c:cesior in :9E7, the alleger pro:e: sed an AWD a-d taeging - crcer to retose . n eie:trical wire so as to cllow installation o' a wit-c: in 6 arehouse all. Tne wire ><as cove:: instead by station mainten-ante personrel with:st tags or an A'.!3. The alleger was fired daring an ensuing argveent cn the control of Contractor A work activities and, he f eels, f or f ollow".ng ac .inistrative reauirements for wnich he vtas held resronsible. Durin; this incident, the Contractor A Superintendent al-lepecly statec that he did not core under N'J or NRC jurisdiction fer following actinistratise requirements. These items were referred to the licensee for follow-up and dispositionir and to address the following: (a) assuring work activities are C0nducted per established pro:ecure controls; and (b) assuring the control of con-tractors is appropriate and that station policies are followed, part A: Licensee review of the first issue concluded that the Contractor A Superintendent operated 480 volt breakers in the Unit 2 warehouse without a tagout, but a tag was not required for the specific activity. The licensee determined that, on August 17, 1987, the Contractor A Suprintendert we rke:: with a representative of a lo:al air conditioning ( AC) comoany to ins estigate a report that air conditioning in the main-
A;pentir 5 23 t-ter.Ance sher, recent'y converted f rom a warehoust , was not opers.ttrg. The tw air conditioning unit had been recently connected, in en unre-lated action to a 4E0 volt distribution panel using a plug-in circu t i breaker. In this application, the breaker is routinely used as-a swit: 1. Other circuits fed from the same distribution panel included welding machine supplies, which were hard-wired to fixed receptacles. With the A: representative a. the cooling unit, the Contractor A Superir.tendent u.anipulated one circuit breaker to turn the AC on. -The designated AC circuit had a temperary label. The unit was left on for the eemairder of the day anc then shut off using the same breaker. The 1 -a eee con-cludes that the actions by the Superintendent were proper b no "wcrk" was done 09 live circuits and no tagout was reouired or nee The
, ins;ector reviewed the state ent of applicability for ACP-0.. c.06A, the i l i t e r. see ' t staticn tagging precedure, and identi'ied no inadequaties.
Tr; n.spe: tar noted that there is a difference between the allnger's and the licens+e's ve-sien of the activity. The alleger stated tr.at the status of the air-conditin unit was not known by the the Contractor AC stricter. dent when he c.ted 6 or 7 oreakcrs in an attempt to start. the u i t. The inspector could ret resolve the cifferent versions. Tne ins:o:t ar conwacted the sileger to obtain any aeditional specific in-fcrmat. that shoss activity cn August 17th-involved "wo-L" on live c ". r :;i t s . No a:diticnal infernation was available to resolve tne cif-ferences. Inspect r review cf this matter did not substantiate tne al-le;er's ar;... , PA-t B- Liceesee r niew of this issue was documented in memoranda cated
'C/ d/57, 11/5/57 et: 5/19/EE. The licensee initiated reviews of this is sue ir. -esper.se to cencerns raised by the alleger in a 10/16/67 letter it the Statier Services Superintendent. The licensee's review cenfirraeJ eisent41 f a:ts ir the alle;er's statcment, as follows. On September 5, 2927 ca penters recuired an electrical wire to be moved in orcer to instail a wine s 3r,in erterior wall of the maintenance shop. The al-leger, acting as Contractor A electrical foreman, obtained a work order (M2-s7-10330) and "blut" safety tags (clearance 1569-87) to pe-forn the work - acting acpsrer.tly independent of direction from the superintendent.
Tte tags we e hung on Tnurscay, September 10 toldo the work on Friday, I. but the work was done on Wednesday by NNECO maintenance personnel. The
- werk was cone-by NNECO cfter the Centractor A Superintendent suggested
- the utility could take the required actions at a cost less than what wculd be charged bv Contractor A. The_ wire / conduit was moved by_-NNE00
, personnel without a.1 authorized work order and by using a modified vor-L sion of the " operator-in-attendance" controls specified in tagging pro-L cedure A:P-QA-2.06A. 'This item is discussed further below.) The action to move the wire was completed before the alleger's tags were hung on l September 30. Vnen the alleger learned on-Friday, September 10 that-the work was already done, an argument occurred with the Contractor A Super-ir,t+ cent en the centrol of work. During i' i t meeting, the Contractor A 54 trinter. dent terminated the alleger fo- it -cbordination and for
- . _ = _
IJpendi B 14 t cha girt time fcr electricians whc did nct participite in the j;d since the only work activity C0mpleted wts by the alleger to process the AW3 and tags. ' The licensee determined that maintenance personnel completed the job on , September 9 in about 30 minutes using 3 workers to control the bre ker on Warehouse 4 Lighting Panel 28. This power source is fed from the non-scf ety related Flanders line. Workers were posted at the breaker and within line-of-sight of the work area and the breaker to ensure the'cir-cuit remained de-energized while the conduit was removed and the wire was moved down to allow installation of a window. The actions taken ret the inttet of "cperator-in-attendance" tagging permitted by Section -- 6.1.9 of ACP-QA-2.00A, but Tag Log Sheet SF-210 was rc' used as reovired
.Jr ty the precedure. The inspector noted that information recorded on SF-i$g 210 w uld have incluced identification of the equipment coverec in the c-der, its location, and appli. cable work order. No other specific in-f c r : t i c r. e n the re;uirec positi cr of the breatc- would base beer re-c.irio. The inspec.cr notec f urther that hac t .e job been coverec by lag;: g Craer IEE9-E7, as initiated ty the alleger, Blue Tags woui: have beer. used (,chich essentially releases the equipment to the person re-stoniitle f or the work) and would have allowed the breaker to be posi-11oned "cs recuitec" by the work party leader. Thus, in the cor trols c ' f ectec y either the Eiue Tag or operator-in-attendance r.ethods, the cesleed rositien of the breater is left to the oiscretion of the verk
; 4 -;3 in res;o-se to the alleger's concern, the licensee nad the onsite Indus-tr'61 Safety Departrent (15D) revies the actions taken in this particular situation. That review concluded that safety was not compromised and '
inat the assu ed ri!* 'n the operation was reasonable and controlled. Sinct ACP-QA-2,06A was written primarily for the control of inplant - ecutoment, er since less stringent controls than are required by the ACP may oe cesirable for work in outlying plant areas, the 150 recem-renced consideration be given to m difying acministrative procedures for rcm plant syster raintenance work to recognize the method used. To this etc, sne Sittien Superintencent issued controlled routing 6926 to address tagging in outlying buildings. Actions to draf t a new procedure were in progress at the end of the inspection period. Notwithstanding the above conclusion on the safety of activities on Sep-tember 9, the licensee concluded the actions were not completed as re-quired by station procedures. Corrective actions were 13 ken by the Station Superintencent in a memorandum (MP-11440) issued to the station on January 29, 1985 which reemphasized the need to follow the require-ments of ACP-QA-2.06A as the nnly currently acceptable process fe- c.afety tagging. The inspector identified no inacequacies in the licensee's conclusions or terrective actions. I
remmarrununes 1 " ~%~5* ',~~
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( J Z^_""'"_"'*."'". L L January 29, 1988 M ' TO: Distribution W / // y FROM:
- Ste hen F.. Scace /
Station Superintendent - Millstone p/ (./ (Millstone extensio- 4300) ~ /
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SUBJECT:
STATION TAGGING / , ..- , -
. - f, A E~cPntly electrica} vork was done on site in an outlying building that required f s a.f e ty ta _gingsk'hile this work was done safely, it was not done in full b - CE vith Safety Tagging ACP-QA-2.06A.
This memo is to remind you that ACP-QA-2.06A represents the only acceptable process for safety tagging and must be followed. A review is underway to determine if a procedure change is warranted to provide a modified approach to safety tagging for equipment not related to plant operation. Until this review and associated changes, if any, are complete, no ( deviation from station procedure is approved. SES:bjs Distribution: List S -
- List DH E. R. Foster
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4; sa..', . - _,,s. ./ . ' . b l v 1 . 12.: TFA15150 PROGRA.". t Training programs have been established to provide all personnel associated v2th the operation of Millstone-2 :ntens:ve training in the various disciplines to ensure that each individual can safely and ef fectively perform his various job functions. These programs ' are performance based and have been developed to satisfy the criteria promulgated by the Institute of Nuclear Power Operations in INPO - 85-002, "The Accreditation of Training in the Nuclear Power Industry." The following training programs have satisfied the criteria of INPd 85-002 by either obtaining accreditation or having submitted the
, self-evaluation reports (SER):
o Shif t Technical Advisor Training (accredited) c Reactor Cperator Training (accredited) e Senior Feacter Operater/Shif t Superviser raining (accreditedl e Noriteer. sed Operator Training (accredited) o Chemistry Technician Training (SER submitted)
.o Electrical Maintenance Personnel Training (SER submitted)
( c Radiolog: cal Prctecticn Technician Train:ng (SER subm:tted) o Instrument and Centrcl Technician Training (SER submitted) i o Mechanical Mainte::ance iersonnel Training (SER submitted) , t o Technical Traintng for Technical Staff and "anager (SER submitted) I l Since the Self Evaluaticn Report is an evaluation against the !NPO criteria and basically a statement that the training programs are ready for accreditation, no change will be submitted to this sect:en of the FSAR solely for the purpose of reporting completten of the accreditation process. 12.2.1 Orran::stien and "anaeerent Management of the Nuclear Training Department is the direct responsibility of the Director, Nuclear Training. The Director reporta to the Senior Vice President, Nuclear Engineering and Opan. ions through the NUSCO Vice President of Nuclear and Enn ' .nental Engineering. The Nuclea r Trainir.; b -pa rtment is organized into four 4) branches and one (1) section to provide focused management '( of specific service areas. The br'anches and sect 2on a re Operator Training, Technical Training, General Nuclear Training, S:mulation Computer Systems Engineering, and Administrative Records Services. Each branch manager a.J 12.2-1
l JUN 10 EG2 ( 12.5 PLANT PROCEDURES t In accordance with the Appendix A and Appendix B of DPR-65, as well as 10CFR50, Appendix B written procedures are required. To meet these requirements, many different levels of procedures and instructions have been established. Administrative Control Procedures (ACP) exist which have estab?ished a system for control of the performance of all auditable activities conducted at Millstone Nuclear Power Station. These procedures are issued by the Station Superintendent. Administrative Control Procedure 1.01 describes this program. Station Procedures are those procedures written on a Unit and Service Group level 9hich control the specifics of station operations including maintenance and modifications, test, inspection, calibration, special processes and plant operating procedures. Station procedures consist of Unit procedures applicable to individual units as well as common site procedures applicable to the entire station. ACP #QA-3.02 establishes the requirements for controlling the Station Procedures. k 9 12.5-1
- ______-______. __ __ - - _ - - __ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ---~
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' e[go 9 UNITED STATES l
.' [ ,a i NUCLEAR REGULATORY COMMISSION j'
t
, a WASHWG TON. D. C. 20555 NOV 0 2 W t s
\ ,
1 _ - I am responding to your letter to Chairman Kenneth Carr of October 4,1989, in which you provided comments concerning the proposed rule pertaining to Section 210 settlement agreements and comments concerning the U.S. Nuclear Regulatory Coemission's (NRC's) statutory mandate. Although the commeat period for the proposed rule expired on September 18, 1989, it is our practice to consider comments received after that date if it is practical to do so. Your comments were received before we completed the proposed final rule and will be considered along with all other comments received by September 18, 1989. On April 27, 1989, NRC sent a letter to all utilities, NSSS vendors, major architect-engineers, fuel cycle facilities and major materials licensees reemphasizing 'their responsibilities to assure that they, and their contractors and subcontractors permit their employees to contact, without restrictions, the NRC with concerns about potential safety issues. The letter notified these organizations that it is not acceptable to have restrictions on communications with NRC in any agreement. Organizations identifying potentially restrictive language in an agreement were requested to (1) notify the af fected party or parties to ignore the potential restriction on communications with the NRC, and (2) notify the NRC that , agreements with restrictive language have been identified. Seventeen agreements were identified with potentially restrictive language. NRC contacted each party involved in these settlement agreemer.ts either directly or through their attorney. This process identified four individuals who - indicated that they had safety issues not previously identified to NRC and NRC is pursuing these safety issues. With regard to your comments concerning the Nuclear Regulatory Commission's actions in regulating the nuclear power industry, the Commission expects licensees to comply with its regulations and, when licensees are found to be deficient, as in the case of operators at the Millstone Unit 3 failing equalification examinations, the Commission's record indicates that it takes appropriate steps. Information in this record was deleled in at:ordance with the it Act, exe tiens fe d{0!Information F0lA- 'I O M// j
. 2-Thank you for your comments. The action thet tne O dnission takes with regard to the proposed rule, along with a summary of the comments received g on the proposed rule, will be published in a subsequ;5t Federal Register
- notice.
Sincerely, f Thomas E. Murley, Director Office of Nuclear Reactor Regulation e 9 9
%- y s , ' * . . /psECCg UNITED STATES 0
NI) CLEAR REGULATORY COMMISSION 4 REcloN I I 475 ALLENOALE h0A0
' KING OF PRUSSIA, PENNSYLVANI A 19400
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RI-87-A-0113 50-336 s, letter --- .f s to your statement to the lac at East lysm, Cu.-Mime, on and to your June 27, 1989 tal call to the imC senior res at Millstcne. A copy of transcript of statement and the Oct r 16, 1987 NRC letter to you (signed by my ~ te, Ebe C. McCabe, Jr.) are enclosed for refennae.
'Ibe enclosed transcript has been reviewed by me and other NRC managers on a regional NBC Allegation Panel. Our review noted that your general ocmcern r about Irm-ruelear activities rot being wnulled by the nuclear tagout I procedure remairs.
We also noted that the trarscript of your - statament allegen that the October 16, 1987 letter frcan the NRC : ... they (tagging procedures) were rxat beirg followed in a non-nuclear cepecity, that would carry over into the nuclear capacity." We differ cm this point. Our Octcibar 16, 1987 letter (Enclosure 2) clid not make such a statement. Our regulatory position is that nco-nuclear activities must tot adversely impact nuclear activities, but nuclear grade pro::edures need not be trwxt to achieve that. If nomwy, we can require ruclear grade controls over related ncn-nuclear areas, in the case ycu identified, NRC follow-up fourri ro unsafe ccniition and to carryover into the nuclear area. , We do agree with ycu that performing similar activities urder differient proccdures can potentially have an adverse inpact on folicwing prerMmes, includirg safety-related procedures. We do not agree that reeily unless nuclear grade controls are generally applied to rcrenuclear activi es. 12sser controls are acceptable in mot,t ncn-nuclear cases. If a carryover into nuclear safety is specificall irdicated for an activity which is not covered by the licensee's nuclear ity assurance program, an evaltaticn is made, appropriate cxantrols are established, and related procedures are upgraded. As our Octcber 16, 1987 letter to ycu stated, we have ccotirued to mcnitor activities involving nuclear safety, incitriirg electrical tagging. Overall, we have fourd the licensee's control of these activities to be acceptable. Our dimicns with the l?tensee have provided informaticn that, after the non-nuclear electrical tagcut ocx:urrerce you described, the licensee deidad to impose adiiticm1 control over non-nuclear tagouts. After considering a information in this record was deleted in accordance wit thefrydom of Information Act,ex tions t IL FolA- .4L NM ] N
.g 3
2 JUL 11 M l' ren-nx: lear tag:ut procedure, the licensee cpted to use the ruclear tagout i thrurfrut the site. This was their choice ard not a recpirment l l by the NRC. ! We acknowleckye ycur June 27, 1989 tal request for fitrther folicv-up of ycur prwicus allegations. We also leipe receipt of yazr talephone statement that NRC follow-up of your irput in 1988 was ineffective in trres not bai rq provantirg Wiat - stated to be the present problem of - ina r llegatisms by lowed y at M111stme Staticm Many different lac irspectors have fcund and to f;ns sourd am and compliance with ruclear safety pwcedIres at Millstene. )SC rwiswa of Millstone ruclear power plant performance have fcarx1 a pet:per saf perspective and apprrpriate controls cmtr activities affecting sat . He see 4 no tracable linkage between your allegaticos and those of the other m11==vs you mentioned. Nrther, ro significant safety iraia?ww'y has been identified in NRC folicw-up of your allegaticras. Our evaluaticn of the transcript of your statamt and of our tal record of your June 27, 1989 call concitded that they contain insuffici identificaticn of safety specifics to warrant additicnal NRC foll . If you have specific details of any tr==,ta$ safety ww-f a= or of uros to 4 cultrol ruclear safety in accordance with the establ prognen, we would apprtclate your sutnittirg thm to this office in writirg, in detail, within 30 days, for cur censideration. Otherwise, we plan to close ycur allegaticns as unsubstantiated, with ro further NRC follow-up. l I bcre that the above inforration resolves your ccncerrs to your satisfacticn. Thank ytt for this cpportunity to address your ecocems. j
- g. C. kmzirger,yx< , n Pro]EL'tS s .
Division of P wr ject ,
Enclosures:
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- UNITED STATES t
') 8 NUCLEAR REGULATORY COMMISSION REGION I 431 PARK AVENUE
% Elho OF PMut&tA. PENNtYtVANIA 1H06 fnfortnation in this record was In accordance with, the Docket no. 50-336 Act,e ptions1 A 0 nfomaUn RI-87-A-0113 F01A- -/ t. L 46 00T 1987 -
Dear This refers to your September 15,
1987 discussion with the Millstone 2 Senior Resident Inspector, Mr. T. Rebelowski . Your concerns, as we understand thee, are about being fired as an electrical fnreman for the W.J. Barney company and that your supervisor stated that electrical tagging procedures need not be followed for construction of the new Unit 2 maintenance building. In regard to your being fired, as Mr. Rebelowski informed you, if you believe there was any discrimination against you, you need to file a complaint with the U.S. Department of Labor (DOL) within 30 days of the incident. The 00L procedure for filing such a complaint was mailed to you from this office on September 16, 1987 with a note from nie emphasizing that such complaints must be filed within 30 days of the alleged discrimination. An additional copy of that procedure is enclosed with this letter. You may also contact a local 00L O'fice on such matters. l l In regard to not following electrical tagging procedures, we understand your j concern to be about carry-over into nuclear construction areas. We do inspect ! nuclear safety-related tagging .t nuclear power plants. At the Millstone l site, we have generally found that the licensee complies with NRC ' l requirements for control over such work. Discrepancies we have noted have been acceptably corrected. Since there are no unaddressed nuclear safety I concerrs in this area as a result of our inspections, and since there was no ! specific nuclear safety inadequacy identified in your input, we plan no specific 'Silew uo of this concern. We will, however, continue to monitor control over set vities involving nuclear safety, including electrical tagging. In addition, we will continue to identify aspects which could be improved to the licensee. Thank you for bringing your concerns to our attention. If you have any additional information or questions, please contact me. Sincerely, M O.kM h Ebe C. McCabe, Jr., Chief Reactor Projects Section 18 Tel: 215-337-5128
Enclosure:
As Stated , k
s
/pe auco,*g, UNITED STATES i ;, NUCLEAR REGULATORY COMMISSION 3e sII.. M 1 j REGION I l .- 5 g . a .- ,e 475 ALLENDALE ROAD t, .g, j:' KING OF PRUSSI A. PENNSYLVANIA 19406
....+
Docket / License: 50-423/NPF-49 M II Northeast Nuclear Energy Cor;.any , ATTN: Mr. Edward J. Mroczka Senior Vice President - NJClear Engineering and Operations Group
- P.O. Box 270 Hartford, Connecticut 06101-0270 Gentlemen:
Subject:
Millstone 3 Routine Inspection 50-423/88-10 (5/24/88 - 7/5/88) - The enclosed report refers to the routine resident safety inspection conducted on , May 24 through July 5,1988 at the Mills' one Nuclet- Power Station, Unit 3. The results of the inspection were discussed wi~ h Mr. C. H. Clement of your staff at the conclusion of the inspection. This inspection noted, as good performance, the conduct of a Special Procedure written to correct a malfunctioning Control Bank. The Integrated Safety Assessment generated prior to performance of the Special Proc . re showed a thorough safety review. No reply to this letter in requi ea. Thank you for your cooperation. Sincerely, h Lee H. Bettenhausen, Chief Projects Branch No. 1 Division of Reaceur Projects
Enclosure:
NRC Rtgion I Inspection Reoort 50-423/88-10 cc w/ enc 1: W. D. Romoerg, Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licensing D. O. Nordq ist, Director of Quality Services S. E. Scace, Station Superintendent Public Document Room (PDR) local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) NRC Senior Resident Inspector State of Connecticut
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'l U.S. NUCLEAR REGULATORY COMMISSION '\.
REGION I Report No. 423/88-10 Docket No. 50-423 License No. NPF-49 Licensee: Northeast Nuclear Ene,rgy Comoany P.O. Box 270 Hartford, CT 06101-0270 Facility Name: Millstone Nuclear Power Station. Unit 3 Inspection At: Waterford. Connecticut Inspection Conducted: May 21 - July 5,1988 Reporting Inspector: G. S. Barber, Resident Inspector Inspector: W. J. Raymond, Senior Resident Inspector Approvec by: (1 C d M , b ~7[/Yh E. C. McCabe, Chief, Reactor Projects Section IB Date Insoection Summary: Inspection on 5/24 - 7/5/88-Areas Inspected: Routine onsite inspection (115 hours) of:-Plant Operations; Status of Prev,ous inspection Findings; Failure of Control Bank "A" to Move During Sur -? veillance Testing; Control Building Isolation Signals Caused by Radiation Monitor Valoperation: Plant Incident Reports; Allegations;- Gamma-Metrics: Post-Accident Moni*aring Instrumentation,10 CFR _21 Report; _ Licensee Event Reports; Maintenance; Surveillance; and Action on'Information Notice (IN) 84-68. Results: No: violations, deviations or unsafe plant conditions were identified. Li:ensee conduct of a Special Procedure written to correct a malfunctioning Control Bank #is a_ notable strength. The Integrated Safety _ Assessment generated prior to performante of the Special Procedure-shoved a thorough review of safety aspects. h
. o TABLE OF CONTENTS \ PAGE 1.0 Persons Contacted.. . .. . .. . ........... ............. .... .... 1
'a 2.0 Summary of Facility Activities. .... ..... .. ... ... .... 1 3.0 Status of Previous Taspection Findings. .. ... .. .... .... ... ..... 1 3.1 (Closed UNR 87-17-01) Follow Actions to Correct Items Identified During Plant Tour... . . . ........... .. .. . ....... 1
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3.2 Followup on Unusual Event Declassification. . . .... . .. ... 4 4.0 Review of Facility Activities. .. . . . .. ... . .. .. . 4 4.1 Failure of Control Bank "A" to Move During Survaillance Testing. 4 4.2 Control Building Isolation Signals Caused by Radiation Monitor Maloperation. . . . . .. .. .. . . 5 - 5.0 Plant Operational Status Reviews. . .. . . 6 5.1 Plant Incident Reports (PIRs). . . .. ... 6
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6.0 Phy s icai Security. . . . . .. .. 7 7.0 Allegations (87"A-011'i. . . . .. . ....... . 7 7.1 SFP Lighting Fixture. .. . . . . .. 7 7.2 Solenoic Valves Environmental Qualification.... . . . . . . 8 S.O Gamma-Metric' Post Accident Monitoring Instrumentation, (10 CFR 21 Report). ... . ............. . .. ..... 9 -> S.0 Licensee Event Reports (LERs). . ... . ... . . .. . 10 10.0 Mr ntenance. . . . ...... ... . . .. 11 11.0 Surveillance... ... . .. . . ........ ......... ........ 11 12.0 Corrections to Previous Inspection Reports.... . ..... ......... . . 12 13.0 Information Notice (IN) 84-68.. .. .. . ...... ......... ..... . . 12 14.0 Management Meetings.. . . . . . ................ . ... 12 i e6
n, .
-l DETAILS *i 1.0 Per;ons Contacted Inspection findings were discussed periodically with the supervisory and man-
~
agement p?r u nel identifled below:
.S. Scace, Station Superintendent C. Clement, Unit Superintendent, Unit 3 M. Gentry, Engineering Supervisor R. Rothgeb, Maintenance Suoervisor K. Burton, Staff Assistant co Unit Superintendent I
!- J. Harris. Operations Supervisor D. McDaniel, Reactor Engireer I R. Satchatello, Health Phjsics Supervisor M. Fearson, Operations Assistant 2.0 Summary of Facility Activities The plant began tne inspection period at full power on 8:30 a.m., May'24 Power was decreased to 90*; due to a seal leak on the "B' turbine driven feed pump. After' repair, full power again was reached at 10:00_p.m. Power was subsacuently lowered to 90!. at 8:03 a.m . May 26, for condenser-backwashing, I and was returned t: 100*; at 3:35 p.m. the same day. Power remained at 100% l , until 12:12 a.m. , June 4 when it was reduced to 90?; for condenser backwash. -I L Full power was achievad at 4:00 p.m. On June 25, another thermal backwash ~' ! recuired a power muction to 90*. at 5:00 a.m. , June 25,- with. a return to full power at 6ila p.m. L ' i In addition, there were 2 short duration decreases (less than 24 hours):of ' , 2*e full power to perform surveillances. The plant continued to operate at-full po,er througn the end of the -inspection period. ! ! H l 3.0 Status of Previous Inspection Findinos 3.1 . M osed UNR 87-17-01) Items Identified Durino Plant Tour- j The Regional Administrator _ toured all three Millstone Units on August 19, 1987. During this tour,.7-generic and 11 unit specific items were L identified. Generic issues applied to-all. units and have been. addressed by the licensee along with the Unit! Specific items. The generic items are summarized below:
- a. Overall material condition of the plant, including housekeeping and-fire prevention controls was very-good. An especially notable finding was- that there were no instances where damaged flexible conduit was cbserved. Damaged-pipe lagging was observe'd.in_some locations.
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- b. The lubrication program for rotating equipment should be reviewed '
to ensure that grease does not enter the motor windings.
- c. The arogram for using calibration stickers and_ trouble report's -;
should be reviewed and revised to be implemented consistently- ; throughout all three units. l
- d. Review the use of nails in scaffuld erection to ensure thav do not enter rotating machinery and sumps,
- e. Control of radiological areas was generally good except that drip ;
pockets in the "C" charging cubicle were not collecting _ leakage. '
- f. Review control of pip caps-on vent and drain valves to ensure.they provide the necessary boundary against leakage.
The licensee has pro /ided tne following resolutions; l
- a. All Millstone 3 Department Heads reiterated to their respective de- 1 pdttment the need to exercise care when working in eQuipmeat spaces !
and discussed the camaged. pipe lagging. The Millstone 3 Maintenance- J Department has 2 individuals assigned full time to lagging. Damaged lagging is repaired as assigned by supervision when these workers are not supporting lagging requirements of ongoing work. Routirie _i resident inspector tours have noted imoroved addressal of damaged lagging,
- b. A history re ort was submitted to the computerized maintenance man -
l agement system (PMMS) and no motor failures were~ attributed to , j grease in the motor windings. tiillstone 3 uses lubrication' tech-
~
nique sheets for motor bearings and no problems were identified ) l ouring the use of the'sneets_ _ The. observed instances of overgreas-ing occurred prior to system turnover and have since been corrected. A program now exists to inspect and clean these motors, as necessary;
- c. Millstone Station now uses a standard set- of Trouble Report tags and stickers which are stocked at the warehouse,. These tags and stickers are fluorescent and include _the applicable Trouble Report-or AWO number. ACP-QA-2.02C'was revised to clarify the use of Trctble Report stickers.
Ret resentatives from all 3 unit's I&C department met to evaluate i the use of calibration stickers. A-review of applicable standards-indicated that stickers should display only the actual calibration date and the identity of_the person performing the calibrati3n; there is no Date Due" requirement for installed plant equip ant. Calibration frecuency, grace periods, and scheduling dates for all Millstone equipment is controlled via the Production-Maintenance Management System (PMMS), applicable PORC approved procedures, and t , EZ, ._ _ . - r
3 the Unit Technical Specifications. Calibration is also subject to , system or unit availability. The use of this block for installed plant equipment is, therefore, not required or desirablu. ( It should be noted that Measuring and Test Equipment (M&TE) does require the "Date Due" b!ock and this practice is consistent within , the station M&TE Lab. The difference in the two types of equipment is rotable. The sticker used for M&TE equipment will remain unchanged and the "Date Due" filled in as required. The sticker for installed plant equipment has been redesigned without the "Date Due" block and will Both be u;ili:ed stationwide on all future equipment calibrations.
; tickers will comply with the following documents: (1) NU Quality Assurance Program Topical Report (WUCAP); (2) ANSI N45.2.4-1972/IEEE 4 Std 33c-1971; (3) Regulatory Guide 1,30; (4) ACP-QA-9.04 These new stickers are on order and will be stocked in tne warehouse.
Consistent implementation has since been observed at all three units.
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- d. Tne station's practice of using nails for constructing scaffolding in equipment spaces was reviewed and determined to require clarifi-cation via a proceaure change. The use of bailing wire as a sub-stitute for nails was evaluated with the conc' ai;n that bailing wire could create more safety and potential problems than nails.
This woulc, therefore, not be an acceptable replacement for nails. Acditionally, as a rule it is not practical to construct scaffolding in cutsice areas and carry it into the work sites. Many plant areas have small access / egress p6ths (radiologically controlled areas, vital areas, etc.) and do not facilitate large pieces of equipment. The use of nails for constructing scaffolding appears to remain the-most feasible fastening material. To better control the use of
~
nails ACF 2.19 " Scaffolding Program" was revised to delineate good construction practices with nails, hazards associated with loose nails and necessary precautions to be taken. Loose nails are now specifically required to be removed after scaffolding erection. Metal scaffolding is now in general use,
- e. The collection bags were repositioned for the items noted and the Maintenance Department retightened valve packing to the extent possible. MP3 has a weekly management inspection by Operations and Health Physics to review the control of radiological areas. This group will continue to look for and correct collection methods which have become ine.ffective. Routine resident inspector tours 'sve identified no recurrence of this problem.
- f. Units 1, 2, and 3 have determined that adequate controls are in place to ensure leakage boundaries are maintained. The coundary valves are aligned by approved station procedures. No credit is taken in the FSARs for Unit 2 cr Unit 3 for leakage protection I
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utilizing pipe caps. The drain or vent valve forms the actua'l:sys- ' tem boundary. Additionally, any system leakage is investigated in-- response to radwaste drain collection increase? or sump alarms. ., The inspector reviewed the licensee's corrective actions and'notedlduring inspections that housekeeping was very good, with one exception being the installatit of conduit covers. This problem was identified to .he licensee and improved eformance has been observed. Various rotating? equipment was spot checked for grease in the motor windings with negative findings. No loose nails have been observed around scaffolding. During the Unit 3 outage, scaffolding inside containment was metal stock fas-tened with clamps, minimizing the use of nails. The. inspector frequently i accompanied the Operations Assistant and HP_ Supervisor on their weekly tours and noted prompt correction of deficient radiological conditions. ' The use of pipe caps was spot-checked with no negative findings. No inaccquacies were noted. There were 11 specific material discrepancies identified. All items have been correctec. Licensee actions were comprehensive and responsive to NRC concerns. No inadecuacies were noted. 3.2 Followuo on Unusual Event Declassification Region 1 Inspection Report 50-423/SS-08 Detail 5.3 identified that the licensee failed to terminateoor declassify an Unusual Event. Implemen-tation of a procedure change to EPIP 4701 has been completed and Step 3.3.1 now recuires termination of Unusual Events by the Shif t Supervisor.
- j. The inspector had no furtner questions on this' item.
a.0 Review of Facility Activities 4.1 Failure of Control Bank "A" to Move During Surveillance Testing During' routine surveillance testing to verify control . rod movement,, con-- trol bank "A" (a non-controlling group) failed to respond to rod motion commands at 5:00 a.m., May 26. Prior to and during attempted; movement, no _ urgent or_ non-urgent -failure alarms were received. The' licensee ente *ed technical specification-(TS) 3.1.3.1.c. af ter verifying that all-rocs aere trippable and aligned within 12 steps of their demanded posi-tion (they were all-full out). TS 3.1.3.1.c. required the inoperable-rods to be returned to an operable status within 72 hours. The' licensee began troubleshooting activities in accordance with Auto-L mated Work Order (AWD) M3-88-08154. The'AWD required instrumentation-L and control (l&C) technicians to determine if thenfailure'to move _was the result of a. bad bank select-switch or movement logic card. The in-spector observed the technicians' activities in the-control rod drive logic instrument racks.
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1 g -1 The technicians (techs) used a calibrated digital volt meter (DVM)'(QA _
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5059 Cal Due 10/22/88) to determine high (14.0-15.0 VDC) and low (0.5-I.0 l VOC) voltages at various logic test points. By using Westinghouse , ;) Manuals for Rod Control, Figure 4-65, Change 2, Sheets 8 and 9, they wre l able to determine that the master cycler select card (A106) was net pro-cessing signals as required. Before replacing the card, the techs had ; to ensure no adverse action would result from pulling the card. If a , logical output that was at a high voltage went to a low voltage, a: block l or inhibit signal could have been removed, and a reactor trip could occur. The tech identified 10 to 12 circuits that would be directly affected. In his review of the first circuit, he discoverec that eight branch cir-cuits were also affected. Because of tne likelihood of many brmh cir-cuits, the technician discussed the drawing review with his supervisor and other managemenc personnel. A decision was made to contact Westing-house (W) to have them check the resul*5 of pulling card A106 on their full scale replica of the control rod drive system. W determined that no adversa action resulted when the card was replaced. The-licensee reviewel this report and, even though the Unit 3 systems were similar (90?. .du;1tcate components) decided to write a Special Procedure to re-place tha cefective card. This Special Procedure (88-3-2) placed all-rod cont-ol power cabinets in argent alarm. This condition energizes both thr stationary and movable grippers for each rod,- holding the rods in place. It allows maintenance on logic cabinet circuitry and portions of the power cabinet circuitry not associated with the stationary- grip-pers. It also requir es a manual reactor trip if rod motian is needed. The ir.spector reviewed the integrated safety evaluation kritten to ensure ' that the procedure met 10 CFR 50.59 requirements-and identified no-in-adequacies. , Tne procedure to replace the defective card was received and approved by the licensee on May 27 and implemented on May 28. The card was e-placed without incident. Some additional troubleshooting was necessary ! before rod group "A" was returned to service. Rod group "A" was returned to an operable status af ter a satisfactory retest and TS 3.1.3.1.c was exitec orier to expiration of the 72-hour action statement. No inade-quacies were noted. l 4.2 Control Building Isolation Sionals Caused by Rad <ation Monitor Malopera-tion The licensee reported that a Control.Beilding Isolation (CBI) was gene-rated at 12:34 a.m., June 11. The signal-was generated when a radiation monitor (3HVC*RE16A) was removed from service to correct a lockup problem-(the display failed to update the. previous reading). All equipment re-sponded as required to the CBI. _ Control room pressurization was defeated I, within the 60 second time delay after verification of_the invalid actu-ation signal. While resetting the monitor's conversion factor to correct
- the earlier problem, a second CBI was received. The plant responded as t .
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6 designed to the second CBI. Proper setpoints were established and the radiation monitor was returned to service. The first CBI was the result of ceenergiziag the monitor to reset the lockup problem. The voltage transient caused a high alarm which generated the CBI. The licensee has ' cautioned operations and maintenance personne regarding this equipment response. The second CBI ..as caused by using an incorrect conversion , factor in establishing an alarm setpoint. The need to be attentive to detail was emphasized to the personnel involved and the procedure was reviewed to ensure it adequately detailed the required actions. No in-adecuacies were noted. The inspector has no further questions on this issue. 5.0 Plant Operational Status Reviews The inspector reviewed plant operations from the control room and reviewed the ope ~ational status of plant safety systems to verify safe operation of the plant in accordance with the technical specifications and plant operating procedures. Actions taken to meet technical specification requirements when epuipmer.t was inoperable were reviewed to veri'.y the limiting conditions for cperations were met. Plunt logs and control room indicators were reviewed te 'dentify changes in plant operational status since the last review and to ver fy that changes in the status of plant equipment was properly communicated in the logs and records. Control room instruments were observed for correla-tion between channels, prcper functioning and conformance with technical specifications. Alarm c:nditions in effect were reviewed with control room operators to verify procer response to off-normal corditions ar.d to verify cperators were knowledgeable of plant status. Operators were found to be cognizant of control room indications and plant s.atus. Control room manning ar,c shift staffing were reviewed anc compared to technical specification re-cuirements. No inadequacies were identified. The following specific activi ' ties were also adcressed. 5.1 Review of Plant Incident Recorts The plant ircident reports (PIRs) listed below were reviewed during the inspection teriod to (i) determine the significance of the events; (ii) reviev the licensee's evaluation of the events: (iii) verify the licensee's response and corrective actions were proper; and (iv) verify that the itcensee reported tne events in accordance with applicable re-quirements. The PIRs reviewed were: number's 1-88 dated 1/5/88, 37-88 dated 2/19/88, 40-8B dated 2/22/88, 41-88 dated 2/23/86, 42-88 dated 2/23/88, 63-88 dated 3/27/88, 64-88 dated 4/4/88, 66-88 dated 3/28/88, 6/-88 dated 4/12/88, 69-88 dated 4/13/88, 70-88 dated 4/14/88, 73-88 dated 4/14/88, 74-88 dated 4/15/88, 76-88 dated 4/14/88, 77-88 dated 4/20/88, 78-88 dated 4/20/88, 80-88 dated 4/22/88, 83-88 dated 4/25/88, 84-88 dated 4/25/88, 85-88 dated 4/25/88, 86-88 dated 4/23/88, 87-88 dated 4/27/88, 88-88 dated 4/26/88, 89-88 dated 4/28/88, 90-88 dated 4/28/88, 91-8B dated 4/28/88, 92-88 dated 4/29/88. The following items warranted inspector followup:
. ~ - .-. _.. - . - - -- . - . . . - - . .. -
y 7 t= 65-88 dated 4/7/88, Spurious OTdT and OPdi alarms. See Inspection Report (IR) 50-423/88-05, De eil 4.2. 6A-88 dated 4/13/88, Turbine / Reactor Trip on Low Condenser Vacuum, :ee IR 50-423/88-05, Detail 5.1. 71-88 dated 4/14/88, Reactor Vessel inner 0-ring failure, See IR 50-423/ 55-05, Detail 5.2. 72-SS dated 4/14/88, RCS U.1 identified Leakage, See IR 50-423/88-05, De-tail 5.3. < 6.0 Phvsical Security Selected aspects of site security were verified to be croper during.in'spection l tours, including site access controls, personnel and Tehicle searches,'per-l sonnel monitoring, placement of physical barriers, compensatory measures,
- _ guard force staffing, and response to alarms and degraded conditions. No inacequacies were identified.
- 7. 0 Allegations (E7-A-0113) l During a meet'.a with the alleger on June 8, 1988 to review to status of NRC findings on reviously icentified concerns (as addressed in NRC Inspection Report 50-33 U88-13), accitional concerns were identified. These issues are discussed below.
7.1 ;FP _ighting Fixtu m I : L This issue involved a concern that an electrician working at the site L for the Millstone 3 outage in November 1987 received exce. ive or need-l less exoosure because he had to replace lamps in the spent fuel pool i underwater lighting fixtures. This matter was referred to the licensee management for followup and evaluation on June 9. l- Based on inspector ciscussions with the licensee and reviews of the j Millstone 3 scent _ fuel pool (SFP) design and radiological conditions,
- this concern was not' substantiated. Underwater lighting for the SFP is L provided by dual lamp fixtures attached to 8long poles that are sus--
I pended from the top of the pool. There are 18 light fixtures.in the SFP. l The- fixtures are shielded from spent fuel stored in racks on the bottom l of the pool by about 20 ft. of-water. Based on reviews-on June 13, the- , inspector _ noted that dose _ rates at the edge of the pool, which contained L the partial core discharge from cycle 1, was less than 0.2 mrem /hr. JCon-tinuous exposure _to_such a field during a 40-hour work week would result l- in a cumulative quarterly dose of about 104 mrem; NRC regulations permit radiation workers to receive from 1250-3000 mrem per quarter, depending I- on previous occupational exposure.' Therefore, replacement of these lamps i- would not result in a substantive radiation exposure or exposure rate. l The inspector noted further that radiochemistry results for the pool b
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water showed low activity. levels _of 10-4 uCi/ml b' eta gamma, which would j require that work on pool fixtures be conducted with some precautions
- to control the potential spread of minor amounts of radioactive contamin- ,
ation. Light bulbs are changed out by removing the lighting fixture _from the . <: pool and working on the deck along the side of the pool. The light fix-tures use high intensity bulbs that rely on water submersion for proper cooling. The licensee stated that, prior to the Millstone 3 refueling outage, all lights in the pool had to be replaced after the lighting
- fixtures were mistakenly energized in a dry condition, which burned out ,
the bulbs. Based en the above, the inspector concluded that no excessive exposure could have occurred to reiamp the SFP lighting fixtures. 7.2 Solenoid Valve Environmental Oualification This item involved a question on the design adequacy of 300 solenoid valves used in Millstone 3 that were found to be undersized and were l scheduled for replacement during the first refueling outage. The item , was referred to licensee management on June 9.for evaluation and response. Based on the information contaired in the allegation, the licensee con-cleded the item most probably referred to actions in-progress to address concerns raisec by the NRC in IE Information Notice (IN) 84-68, Potential Deficiency in Improperly Rated Field Wiring to Solenoid Valves (50Vs). L Experience at an,ther nuclear facility had shown that field run cabling
.o certain solencid valves had an insulation' temperature rating of.144
- cegrees F. In certain SOVs wnere the field cable terminates at coil-lugs mounted inside the valve ~ housing, the insulation could be subjected to temperatures in the 250-280 degree F range, which couid degrade the in-
- sulation prematurely and cause ins'ulation failure. . Licensee reviews in l- response to IN_84-68 concluded the problem potentially applied to Mill-l stone 5, although no failures of the type described had occurred. The present status of licensee actions and evaluations were described in a NUSCO Qualification Engineering memorandum GSP-88-022 dated February 3,.
1988. The licensee concluded that about 300 SOVs used in various safety-related and non-safety-related applications possibly had the problem described in the-information notice. . Subsequent review concluded that the Target-Rock valves installed in the plant _were.not of concern, and that_certain ASCO valves were. ' Of 209 ASCOs in use,117 ASCO Type NPK valves were . deemed a potential concern since they were normally energized. Data was obtained from the valve vendor to determine the temperature rise of the coil internals for various ambient temperatures. Initial calculations by-the architect-engineer ( AE) showed the valves had a qualified life - of about 6.3 years at the elevated temperature. -The AE recommended that modifications be made at tne first refueling outage (starting _in November
9 u , 1987). The initial modification considered was to terminate the field t cable at a junction box outside the SOV and to install cabling with a higher temperature rating from the junction box to-the valve. Plans to modify the valves were initiated at the start of the refueling outage through the issuance of work orders. The plans to modify the valve wiring were deferred to the second refuel-' ing outage based on updated field inspection data obtained during the first refueling outage and a revised engineering evaluation (thermal life calculation PA-78-236-743-GE), which showed that.the cables installed had a longer qualified life than previously caiculated. The field in-spections were conducted to identify the specific type of cable (Okonite, Kerite, etc.) associated with each 50V and to inspect'the condition.of the wires at the termination lugs. No signs of burning, cracking or other evidence of heat stress were observed. The revised thermal life calculations were based on actual ambient temperatures for each valve (which were generally less than those assumed in previous calculations) and showed each valve had a qualified life good through the end of the second refueling outage. Subsequent licensee action on this item is_ tracked by Commitment 3 0030 dated March 5, 1988. The licensee plans to replace the valves on a phased basis according to the calculated end of qualified life. .The licensee plans to upgrade the 50V coils to an NP type, which is provided from the manufacturer with a high temperature pigtail. The inspector icentified no inadecuacies with the licensee's eval.uations or plans to disposition the issues raised by IN 84-68. Actions to replace components baseo on calculated qualified life.as a preventive maintenance activity is an accepted industry practice. Based on the above, concerns regarding the safety of the plant and the adequacy of plant systems to perform intended functions were not sub-stantiated.
8.0 Gamma-Metrics
Post-Accident Monitoring Instrumentation (10 CFR 21-Recort). The licensee notified the inspector on May 23 that-solder connections on. Gamma-Metrics (G-M) cable assemblies may be susceptible to moisture intrusion during a dcsign basis accident'(DBA). On February 19,'1988,: G-M made a pre-- liminary 10 CFR 21 report that identified.recent environmental qualification tests of a solder joint-en a metal hose of a G-M cable assembly.which had-failed to hold pressure-at elevated temperatures. The licensee was notified. of this problem by letter from G-M on February 22. -G-M also provided a. letter
.to the licensee on May 10, to include guidance for' inspection of, neutron flux-monitor cabling and evaluation of.a-retrofit to provide additional sealing of the conduit at specific connections to prevent moisture-intrusion.
The G-M neutron flux monitor and cabling assemblies are used'to provide the operator neutron flux inoication-from the' source range to 150% power in Post Accident Monitoring Environments (Regulatory Guide 1.97). Millstone 3 tech-
10
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nical specification (TS) 3.3.3.6 specifies the operability requirements of -- wide range neutron flux monitors with respect to their post-accident monitor-ing function. TS 3.3.3.6 requires that at least one of two channels of Post- , Accident Neutron Monitoring be operable in Modes 1_through_4. If_both chan-nels are inoperable, one channel must be restored to service or a shutdown must be begun _within 7 days. ' The licensee completed their review of the environmental qualification of ti.e G-M cable assemblies on May 27 after contacting the vendor. A phone conver -- ' sation was held on May 25 between the licensee's Qualification Engineering (QE) vendor department and the vice president (VP) of Gamma-Metrics. The subject of this conversation was the February 22 and May 10 letters from Gamma-Metrics. This problem was identified by the vendor during recent qualification testing of Gamma-Metrics Neutron Monitors for BWR applications. The root cause of the problem was identified as voiding in the solder joints of in-containment cable assemblies, allowing moisture to migrate to the vari-ous cable connectors. The voiding results in voltage discharges across the insulation in these connectors. That results in signal degradation that looks like increased neutron flux. In the licensee's discussion with the vendor VP, it was noted that Gamma-
- Metrics had changed their shop fabrication procedure for performing the pre-tinning of the solder joints in question. Prior to 1984, a solder pot. dip prc
- ess was utilized, and this was the method used to fabricate the cable assemblies tested in the original qualification testing. The latest qualifi-cation testing (the tests that failed) was done on cable assemblies-fabricated using an iron-applied tinning. .The G-M VP also stated that they have compared finished solder joints f abricated using both methods, and have observed voids, similar to the failed cable assemblies only in those samples using-th' iron-applied tinning method.
i L The G-M VP stated that Gamma-Metrics shipped the original cable assemblies l for Millstone Unit 2 and Millstone Unit 3 in December of 1982,_ ano that the tinning procedure definitely changed after December 1982. Therefore, the l licensee's QE concluded that the original cables are fully qualified and l. operaule since they are identical in form, fit, function,' materials and manu-I facture to those which were successfully qualification tested in Gamma-Metrics - P Report.No. 010, Rev. O. The inspector reviewed the issue and had no_.further Questions.
-9.0 Licensee Event Reports (LERs)
Licensee Event Reports (LERs) submitted during the report period were reviewed- . to assess LER accuracy, the adequacy of corrective actions, compliance with 10 CFR.50.73 reporting *equirements and to' determine if there were generic implications _or if further information was required. Sele <:ted corrective actions were reviewed for implementation and-thoroughness. The LERs reviewed were:
-- - . . . - - - - . - - . , , . - . - - - - - - - . --. ~.
4: 11 4-c LER 88-14-00, Reactor / Turbine Trip due to. Low Condenser Vacuum. S e e _: Inspection Report 50-423/88-08, Detail 5.1. LER 88-15-00, RCS Unidentified Leakage Action Statement Improperly Ter-minated. See Inspection Report 50-423/88-08, Detai) 5.3.. LER 88-16-00, Mode Change with Action Statement in Effect due-to Ferson-nel Error (NV4 88-10-01). This licensee-identified item was evaluated as being of low safety significance, appropriately reported and corrected,- and not a result of inadequate corrective action on a prior viola? ton. Thtrefore, no Notice of Violation was issued. No inadequacies were noted. i 10.0 Maintenance The inspector observed and reviewed selected portions of. preventive and cor-rective maintenance to verify compliance with regulations, use of administra-- tive and maintenance procedures, compliance with codes and standards, proper QA/0C involvement, use of bypass jumpers and safety tags,. personnel protection, anc equipment alignment and retest. The followingEactivities'were included: i Main Generator Voltage Regulator, AWO M3-88-09C68 -l Particulate and Gaseous Radiation Monitoring Auto Filter Repairs, d (3HVR-P3), M3-SE-09138 j l Fire Protection Low Pressure CO2. Inlet Isolation Valve Body-to-Sonnet Leak, M3-88-01000 , No inadequacies were noted. 11.0 Surveillance. The inspector observed portions ci surveillance tes;s to assess performance in accordance with approved procedures and-Limiting Conditions of Operation, , removal- and restoration off equipment,-and deficiency review and resolution. The following tests were reviewed: Auxiliary Building Filter- System Operability Test, SP 3614A.1 Emergency Generator Fuel'011. Particulate Sample, SP36468.8-
= Auxiliary- Feed Pump (3FWA*P1A) Operational. Readiness Test, SP3622.1 -
No incdequacies were noted.
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12 7 12.0 Correctionr to Previous _ inspection Reports Region I Inspection Report 50-423/88-02, Detail 6.1 (pg. 9) should read, '
"no inadequacies were noted vice " inadequacies were noted," <
Region I Inspection Report 50-a23/88-02, Detail 7.0 (pg. 10) should read, ',
"which was dfe on January 12, 1988" vice "which was due on January 10, ,
1988." Region-I Inspection Report 50-423/88-08, Detail 7.0, LER 88-12-00, should-read "NV4 88-08-02" vice "NC4 88-08-02." 13.0 Information Notice (IN) 84-68 The inspector reviewed licensee acticr.s taken to address the _ issues raised in IE Infcemation Notice ( d) 84-65, Potential Deficiency in Improperly Rated Field Wiring tn Solenoid Valves. The review was performed to determine . whether the licensee actions taken ,r planned to address the NRC concerns were. acceptable. Licensee actions are discussed in Section 7.2 above and were found satisfactory. No inacequacies were identified. _14.0 Management Meetings
-Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also dis-cussed at the conclusion of the inspection. No proprietary information was covered within the scope of the inspection. No written material was given to the~ licensee during the inspection period.
2 "c e. l 3~ ,q UNITED STATES g, p NUCLEAR REGULATORY COMMISSION
. o ! REGION I
\' "' 475 ALLENDALE ROAD KING OF PRUSSIA., PENNSYLVANIA 19406 t
fl0V 3 4 gg Docket No. 50-423 Northeast Nuclear Energy Company ATTN. Mr. Edward J. Mroczka Senior Vice President - Nuclear Engineering and Operations P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region 1 Inspection No. 50-423/90-19 This letter transmits the report of the routine resident safety inspection conducted by Mr. K. Ko!aczyk and others of this of fice on September 5 through October 15, 1990 at Miilstone Nuclear Power Station, Unit 3, Waterford, Connecticut. The r eport covers activities authorized by NRC License No. NPF-49. The inspection findings were discucsed by Mr. Kolaczyk with Mr. C. Clement and others of your staff at the conclusion of the inspection. Areas examined curing this inspection are described in the NRC Region I inspection report which is enclosed with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, measurements, and document revier Based uocn the results of this inspection, it appears that one of your activities was not conducted in full compliance with NRC requirements, as set forth in the Notice of Violation enclosed herein as Appendix A , and further discussed in section 7.3.1 of the enclosed inspection j repcrt. The violation concerns the failure to post a compensatory - fire watch when a fire barrier was taken out of service in accordance with plant technical specification 3.7.13 " Fire Rated Assemblies". This licensee-identified violation has been categorized by severity level in accordance with the " General Statement of Policy and Procedures for NRC Enforcement Actions," 10 CFR 2 Appendix C (Enforcement Policy 1930). You are required to respond to this letter, and in preparing your response, you should follow the instructions in Appendix A. We have noted that previous lapses in the fire protection program were attributed to inadequate procedures and poor labeling of fire related assemblies. The recurring nature of these problems and your failure to initiate adequate preventative measures resulted in issuarte of a Notice of Violction in December, 1988. Although improvements have been noted through revised procedures and improved labeling of fire rated assemblies, our review has concluded that ineffective communications between fire watch personnel continue to result in poor implementation of the fire protection program. Other enfortm..ent action was Jb//MM39: I ch
4 Northeast Nuclear Enorgy Company 2 3 considered, because of the repetitive nature of these events, but not t taken, based upon our determination of the low safety significance of the violations. Nonetheless, the failure to address a continuing
-program weakness is of concern. Therefore, we request that, in your ,
response to the Notice of Violation, you inform us of the actions that will be taken to ensure the fire protection program is effectively implemented. One non-cited procedure violation is included in the enclosed report, in Section 3.7, involving a procedure violation during this reporting period. This violation occurred in spite of corrective actions in response to a previous procedure violation, discussed in Sections 3.6 and 3.7. I am concerned about the apparently repetitive nature of these procedure violations, even after receipt of the training discussed in your letter of July 31, 1990, to the Region I office, in response to the previous violation (90-08-01). Please respond to this letter, within 30 days of its receipt, describing those actions you consider necessary to minimize the occurrence of additional procedure violations. The response requested by this letter is not subject to the clearance procedures of the Office of Management an Budget, as required by the Paperwork Reduction Act of 1980, PL 96-511. Iour cooperation with us is appreciated. St.ncerely,
- l ^
Edward C. Wenzinger, Chie Projects Branch No. 4 l , Division of 1 Reactor Projects
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. NRC Region I Inspection Report No. 50-423/90-19 cc w/encls:
W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Services R. M. Kacich, Manager, Generation Facilities Licensing S. E. Scace, Station Director, Millstone C. H. Clement, Nuclear Unit Director, Millstone 3 Gerald Garfield, Esquire Public Document Room (POR) Local Public Document Room (LPDR) Nuclear Safety Informction Center (NSIC) NRC' Senior Resident' Inspector State of Connecticut
.y r #N i U.S. NUCLEAR REGULATORY COMMISSION y REGION I .
Report No.: 50-423/90-19 Docket No.: 50-423 License No. NPF-49 Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit No. 3 Inspection at: Waterford, Connecticut Inspection Conducted: September 5 through October 15, 1990 , Reporting Inspector: Kenneth S. Kolaczyk, Resident Inspector, Millstone 3 Inspectors: William J. Raymond, Millstone Senior Resident Inspector Kenneth S. Kolaczyk, Resident Inspector, Millstone 3 David H. Jaffe, Project Manager, Project Directorate , I-4, Office of Nuclear Reactor Regulation
<//k!seu Donald R. Haverkamp, Ch(ef b nlN!%
Date Reactor Projects Section-4A. ! Division of Reactor Projects
.Insoection Summary: Inspection on September 5, through October 15, 1990 (Inspection' Report No. 50-423/90-19).
Areas Inspected: Routine onsite inspection at Millstone 3 during. normal and L backshift work periods of operational safety . including plant operations, radiological and chemistry controls. and security; maintenance and surveillance; engineering and technical support; and safety assessment and quality ver+ Y1.:a ti on . Results: See Executive 4. mary i e 9/p?c2%W
18
\
After completing a review of the Millstone 3 procedures which implement the fire detection program, the inspector determined that they provide sufficient guidance concerning the type of action that should be taken when fire detection and protection assemblies are , disabled. Therefore, NRC violation 88-23-01 which tracked licensee correction of procedural weaknesses in the fire protection program is closed. The inspector determined that inadequate communication be-tween plant personnel and ineffective implementation of station pro-cedures continue to degrade the fire protection program. Inadequate communication and procedure implementation resulted in LERs 90-18 and 90-27, and PIR 90-127. In a February 21, 1990, letter to the NRC in which the licensee responded to the SALP report 88-99, the licensee stated that communications in the fire protection area would be im-proved. The inspector believes that improved communications between the operations department and the fire watch patrols should be imple-mented. The plant director informed the inspector that he has held discussions with the NFM supervisor concerning implementation of the fire protection program. During those meetings, the plant director emphasized that he expects fire watch personnel to ask questions ,1f a condition appears to be abnormal and inform the plant shift supervi-sor. Additionally, he emphasized the importance of performing duties correctly the first time. The director indicated that a task force has been established to examine ways to improve communicatior.s be-tween the operations department and fire watch personnel. The recent failure to establish a fire watch within one hour when the fire related assembly was disabled is a violation of TS 3.7.13. Although previous violations of this technical specification have + been considered licensee identified, and per the policy of 10 CFR 2 ' Appendix C, no violation was issued, the recurrence of these events suggests additional corrective action should be taken by the license to prevent recurrence. Accordingly, enforcement discretion is not being exercised in this instance and a violation (50-423/90-19-02) is being issued as a result of continuing weaknesses in the fire pro-tection program with specific regard to the poor communications that have existed between plant personnel and inadequate implementation of the fire protection procedures. 7.4 Quality Assurance Issues Closed 7.4.1 Adeouacy of Comouter Software The inspector reviewed the following three concerns of a former licensee employee regarding the adequacy of computer software pro-grams at Millstone Unit 3. (1) Nuclear Engineering and Operations Procedure 2.24, " Quality Software Programs" is inadequate in that it does not fully implement recognized industry standards. For example, NEO 2.24 does not fully adhere to ANS 7.4.3.2. " Application Criteria for Programmable Digital Computer Systems in Safety Systems of l
i 19 Nuclear Power Generating Stations." (2) The verification / valida- C tion process for computer software programs at the Millstone Site is deficient since the verification / validation process does not include a functional test with specific predetermined acceptance criteria. (3) Personcel qualifications for review of software are not specified by procedure. Therefore, the possibility exists for unqualified people to conduct reviews of sof tware packages. These concerns were outlined in a June 6,1990, letter which- provided the concerns to the licensee for investigation. In an August 3, 1990, letter which responded to the issues, the licensee concluded . that the individual's concerns were not substantiated. The bases for this determination is provides as follows: In response to the first concern regarding the technical adequacy of NE0 2.24, the licensee ' noted that revision 1 of NEO 2.24 dated October 1989 committed to , regulatory guide 1.152 which' endorses ANS 7.4.3.2. To implement the requirements of 'ANS 7.4.3.2, Q5-12, " Additional Requirements for Category I Sof tware" was written and and effective on June .22,1989. In response to the second concern regarding the adequacy of the verification / validation process of sof tware at Millstone Unit 3, the licensee reviewed sof tware implementation packages (SIPS) for control grade and quality software instituted from 1987 to the present. The results concluded that the verification /valication performed was adequate based upon the complexity of the changes performed. In followup of the issue which concerned the adequacy of qualifica-tions of personnel who conducted review of sof tware. packages, the licensee concluded that software packages were reviewed by-personnel who are qualified by experience. education or both, This 1s in , accordance with station procedure: ACP-QA-2.13A " Computer Softs.:-e i Implementation." Insoector Review of_ Licensee Resuonse Inspector followup of this issue consisted of a review of station procedures, software packages, and interviews with personnel. The inspector reviewed 05-12 and concluded.that the procedure fol-lowed the intent of ANS 7.4.3.2 and regulatory guide 1.152. There-fore, the inspector concluded that procedures written in accordance-with this standard would follow the standards that the NRC has.re-cognized as at:eptable for QA category I software. The inspector reviewed plant design change request (PDCR) 89-15 which modified the-software program for the inadequate core cooling system (ICCS), which is the only software program that falls within the scopelof RG11.152. Through conversations with instrumentation and control technicians who performed the change,_ computer _ programmers and the system engin-- eer who was responsible for implementation of the PDCR, and through. independent review of the PDCR package,.the inspector concluded that the functional test performed was sufficient to test the effect of the sof tware change on the plant. i; , l f d L o a - _ . - - . . _ _ _ ._ ..
20 The inspector revieved ACP-QA-2.13C " Control of Systems Containing , Category I Software" and noted that the procedure which provides instructions for the control, implementation, and testing of category I sof tware, requires personnel to conduct independent reviews in ' accordance with ACP-QA-3.13 " Preparation, Review, and Approval of Design Analyses and Calculations." Add' lons11y, ACP-QA-2.13C re- , quires personnel who conduct independent reviews to be qualified in accordance with ACP-QA-8.16 " Training, Certification and Identifi-cation of Qualified Inspection, Examination and Testing Personnel." Based upon inspector review of the above li .ed t procedures, the in-spector concluded that Millstone Station s N eaures adequately identi-fy the qualifications that an individual v. . have prior to perform-ance of an independent review. Based upon inspector review of the licensee investigation and independent inspector followup, the in- , spector concluded that the individual's concerns regarding computer software were unsubstantiated. This matter is closed. 7.5 Manacement Meetinos
- Periodic meatings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also discussed at the conclusion of the inspection. No proprietary information was covered within the scope of the inspec-tion. Written material available in the public document room, con-cerning the completion of non-code repairs to ASPE piping systems, was given to the licensee during the inspection period, f
se
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[ In j NUCLEAR REOULATORY COMMISSION REGION I -- 476 ALLENDALE 80AD
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AUG 0 31900 Docket No. 50-336 Fil No. RI-90-A-0033 Information in this record was deleted
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m of Information < ct e t on F0IA. -I b1 Dear y
Subject:
Allegation Concerning M111 stone Nuclear Power Station Unit 2 The NRC Region I office has completed its follow up in response to the concern you brought to our attention on March 15, 1990, alleging that improper sur-veillances were being performed on the main station batteries by inadequately qualified personnel. We found your allegation to be substantiated and have documented our findings in detai) 8.2 of Inspection Report 50-336/90-09, dated 28 June 1990. Extracts of that report are enclosed. Based on the results of our inspecaton of your concerns, a violation was identified. Northeast Nuclear Erwrgy Company is required to inform us of the corrective actions they have taken or plan to take. Our inspectors will con-tinue to monitor the licensee's activities to ensure proper resolution of this matter. We appreciate your informing us of your concerns and feel that our actions in . this matter have been responsive to those concerns. Should you have any additional questions, or if I can be of further assistance in this matter, please call me collect at (215) 337-5120. Sincerely,
*h$/ 44<
Donald R. Haverkamp, Chief Reactor Projects Section 4A Division of Reactor Projects
Enclosure:
As Stated bec: A. Vegel W. Raymond Allegation File No. RI-90-A-0033 ,
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[ g l ***** KING OF PRUSilA. PENNSYLVANIA 15404 - . JW 2 8 1990 Docket No. 50-336 k Northeast Nuclear Energy Company ' ATTN: Mr. Edward J. Mroczka Senior Vice President - Nuclear Engineering and Operations P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region ! Inspection No. 50-336/90-09 This refers to the routine resident safety inspection conducted by Mr. P. Habighorst of this of fice on April 17 - May 29,1990 at Millstone Nuclear Power Station, Unit 2. Waterford, Connecticut of activities authorized by NRC License No. DPR-65 and to the discussions of our findings by Mr. Habighorst with Mr. J. Smith of your staff at the conclusion of the inspection. Areas examined during this inspection are described in t.he NRC Region I inspection report which is enclosed with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, measurements, and document reviews. Based on the results of this inspection, it appears that one of your activities was not conducted in full compliance with NRC requirements, as set forth in the Notice of Violation enclosed herein as Appendix A. The violation involved the improper performance of surveillance required by technical specifications on . the main station batteries. This violation has been categorized by severity level in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (Enforcement Policy, 1990). You are required to respond to this letter and in preparing your response, you should follow the instructions'in Appendix A. While the safety significance of
-the violation is low, we are concerned that procedures were not followed on numerous occasions and that a supervisor accepted out of specification results without an adequate documented evaluation. Please give particular attention to these matters in your response.
Your cooperation with us is appreciated. n rely, Edward C. Wenzinger, Chie Projects Branch No. 4 Division of Reactor Projects
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. NRC Region I Inspection Report No. 50-336/90-09 {
h67/863N~~
y 3 APPENDIX A ( NOTICE OF V10:,ATION
- Northeast Nuclear Energy Campany Docket No. 50-336 ,_
Millstone Nuclear Power Station Unit 2 License No. DPR-65 As a result of the inspection conducteo on April 17 - May 29,1990, and in C accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (Enforcement Policy, 1990),the following violation was identified: Millstone Unit 2 Technical Specification 6.8.1 requires, ir. part, that written procedures be established, implemented and maintained as recommended 1.. Appendix A of Regulatory Guide 1,33, February 1978. The Millstone Nuclear Power Station Unit 2 Final Safety Analysis Report, section 8.5.4.2, states that the electrolyte level of each maia station battery cell shall be checked and all water additions recorded to ensure functional capability and detection of batte'ry degradation. Millstone Unit 2 battery surveillance procedure SP-2736A, a procedure recommended.by Regulatory Guide 1.33, requires that battery cell levels which do not meet the specified acceptance criterion be recorded, and that the amount of water added to any cell be documented. Millstone Nuclear Power Stati;n Administrative Control Procedure ACP-QA-8.16, a procedure recommended by Regulatory Guide 1.33, specifies the minimum qualifications required for certifying personnel who perform quality-related testing activities. f Contrary to the above, on March 7,1990, uncertified contractor personnel performed a main station battery surveillance while not under direct-observation of certified test personnel; levels of battery cells not meeting the procedure acceptance criterion were not documented; and water additions required by the procedure were not performed. On March 14, 1990, the amount of water added to individual battery cells was not documented. This is a Severity Level IV Violation (Supplement I). , Pursuant to the provisions of 10 CFR 2.201, Northeast Nuclear Energy Company is hereby required to submit to this office, within thirty days of the date_ of the letter which transmits this Notice, a written-statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3)'the date when full compliance will be achieved. Where good cause is shown, consideration will be given to extending this response time.
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U.S. NUCLEAR REGULATORY COMMISSION REGION 1 ( Report No.: 50-336/90-09 Docket No.. 50-336 , License No. OPR-65 Licensee: Northeast Nuclear Eneroy Company P.O. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 Inspection at: Waterford, Connecticut Dates: April 17 - May 29, 1990 Reporting Inspector: Peter J. Habighorst, Resident Inspector Inspectors: William J. Raymond, Senior Resident Inspector Peter J. Habi9 horst, Resident Inspector Thomas Moslak, Resident Inspector, Three Mile Island
- Douglas Dempse~, Resident Inspector, Millstone 1 Approved by
- /WM h 9 90
! ObnaldR. Haver %mg ief Fate' < Reactor ProjectLM on 4A Division of Reactor Projects Inspection Summary: Inspection on April 17, 1990'- May 29, 1990 Inspection Reoort No. 50-336/90-09 Areas Inspected: Routine NRC resident inspection of plant operations, surveillance, maintenance', previously. identified l items, engineering / technical support, committee activities, periodic reports, licensee event reports, and allegations. Results: See Executive Summary
-2
+
20 The licensee corrective &ctions in response to the violation for the control of loop circuit diagrams were considered adequate. The c actions were implemented, in part, under departmental instruction
- 1.09, which was effective December 5, 1989. Instruction 1.09 provides a method for technicians and other site personnel to verify ,
that the most current revision of loop diagrams is available by use of the generation records informational tracking system (GRITS). The inspector noted a minor implementation discrepancy in instruction : '- 1.09 in that technicians did not initial the drawing after verifica-tion of the GRITS program. The inspector presented this to the licensee. The licensee acknowledged the discrepancy and provided corrective actions. On December 14, the licensee completed an internal review of two cases of the out-af-date loop drawings. The review concluded, in both cases, that the drawings were not distributed to the instrument and controls department from the nuclear records department. The correct drawing revisions were distributed to other Unit 2 depart-rents. The licensee took corrective actions to address the routing error in the distribution of controlled drawings- . The inspector reviewed the results of the program to upgrade instru-ment locp folders. Out of approximately 2900 loop folders for safety and non-safety instrumentation, less than 1.5 percent (41 loop dia-grams) contained drawings with the inappropriate revision. All dis-crepancies were resolved by the licensee. The inspector independently reviewed twenty randomly selected ! instrument folders to assess the edequacy cf licensee loop diagram ! upgrade program. The review was completed by comparing the drawings l in the loop folders with the GRITS program output. No inadequacies , w?re noted. l In conclusion, the loop diagram revisions were out-dated in the associated loop folders and the engineering master file. Inspector review identified that tne particular safety-related work activity was accomplished with current drawings, and no compromise to safe-work occurred. The inspector currently plans ro further action and considers this item closed, i 8.2 Concern: Main Station Battery Surveillance Test performance j . (RI-90-A-033) On March 15, 1990, the inspector reviewed several concerns regarding the performance of technical specification surveillance on the Unit 2 main' station batteries. The concerns included the following items:
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--On March 7, 1990, two non-certified contractor employees performed i technical specification surveillance on main station batteries 201A and 201B with no direct observation or supervision by qualified unit c) 1 maintenance personnel as required by Millstone station quality control procedures. ,
--The electrolyte levels of a large number of battery cells were found to be at or below the minimum required by technical specifica-tions and the surveillance procedure, but had been logged as " normal" i on the surveillance data sheet. 1
--The contractors had reported the unsatisfactory levels to an assistant maintenance supervisor, who responded that the low levels were due to low temperature in the battery rooms and were not a problem.
The safety-related batteries at Unit 2 are C&D type LCV-33, 60-cell, calcium grid lead acid batteries rated at 125 volts de and 2320 ampere-hours (8 hour rating). They are designed to provide reliable power to systems required for safe unit shutdown including, through inverters, the reactor protection and engineered safeguards systems and vital instrumentation. Section 8.5.4.2 of the Unit 2 final safety analysis report (FSAR) states that the electrolyte level of each battery cell is checked and all water additions recorded to ensure functional capability and detection of battery degradation. Technical specification 4.8.2.3.2 requires that battery operability be demonstrated at least once every seven days by verifying that the electrolyte level of each pilot cell is between the minimum and maximum level indication marks. Surveillance of the station batteries is performed weekly in accor- . dance with SP-2736A, Battery Pilot Cell Surveillance, revision 2 ' dated Oc. ember 23, 1988, and quarterly in accordance with SP-27368, Complete Battery Cell Measurement, revision 3, change 1, dated July 23, 1986. Regarding cell level, both procedures require that the electrolyte be between the minimum and maxinum level indication marks. The amount of water added to any cell is to be documented on procedure data sheets. Step 6.6.1 of SP-2736A states that for battery cells not within acceptance criteria, level shall be recorded. Finally, any measurements that do not meet acceptance criteria must be reported to the assistance maintenance supervisor. l These procedures are consistent with vendor technical manual VTM2-127-001A, Station Batteries (C&D Power Systems), and Institute of Electrical and Electronics Engineers (IEEE) Standard 450-1980., Recommended Practice for Maintenance, Testing, and Replacer.ent of Large Lead Storage Battaries for Generating Stations and Substations. l G l
~ ' 22 On March 15, 1990,- the inspector toured the unit 2 battery rooms and observed that the electrolyte levels of most of the 120 cells were t above the minimum level mark and six cells were at the minimum mark.
- This condition was considered by the inspector to be adequate. The inspector reviewed weekly surveillance data sheets for the period ,
February 14 through March 28, 1990 and noted the following:
--In all cases, electrolyte levels were recorded as " normal". A note a prior to step 6.7 of procedure SP-2736A defines normal as half the distance between the high and low level marks.
--Water additions to the batteries were logged on February 14, and March 14, 21, and 28,1990. Tne method of logging additions was not uniform. Amounts added were logged either in milliliters or inches of cell height. On March 21, 25 cells of battery 201A and 43 cells of battery 201B were noted to be at or below minimum level. From the data recorded, the inspector was unable to determine either the as-found levels of individual cells or the amount of water added thereto.
--On March 7,1990 no water additions were recorded and all cell levels were logged as normal. The inspector also reviewed the automated work orders (AWO) for battery surveillances performed on March 7 and 14, 1990. The results logged in these AW0s were consistent with those recorded on the surveillance data sheets.
Licensee administrative control procedure ACP-QA-8.16, Training, Certification and Identification of Qualified Inspection, Examination and Testing Personnel, revision 18, dated March 22, 1988, establishes the method of and minimum qualifications required for certifying test personnel at Millstone. Personnel who are assigned responsibilities , or authority to perform testing activities shall be certified in writing as qualified. Through interviews with the unit 2 maintenance manager and also with the assistant maintenance supervisor involved in the allegation, the inspector determined that the contractor personnel who performed battery surveillance on March 7, 1990 were not certified per ACP-QA-8.16, and had performed the surveillance activity while not under direct observation. The inspector interviewed the unit 2 maintenance manager on April 5, 1990 regarding the allegations and received a written formal response dated April 18, 1990. This response stated that the contractors involved had reported to a maintenance supervisor on March 7 that 8 to 10 cells were below the minimum level line. The cognizant supervisor, who signed off the surveillance procedure and the AW0s, had not been present during the performance of the activity and was not privy to this information. He signed the AWO based on review of the data presented, verification of test instrument calibration, and discussion of the procedure and results with the contractors. As a
- . ._ ~ - _. . _
e - 23 result, the battery cell level data recorded on'the maintenance forms associated with the surveillance performed on March 7, 1990 are g incorrect for the 8 to 10 unidentified cells involved, s The inspector concluded that at no time were the cell olates of bat-teries 20]A or 201B likely to have become uncovered. IEEE Standard ' 450-1980, Appendix D1, Urgency of Corrective Actions, states that the addition of water is not urgent unless the tops of the plates are in danger of being exposed. Thus, the ability of the station batteries j to function if challenged was unaffected by the low water level con-dition, and the inspector considered the safety significance of the condition to be minimal. Nevertheless, the inspector identified the following concerns regarding this event:
--Contrary to the requirements of ACP-QA-8.16, unsupervised surveillance was performed on safety-related equipment by uncertified contractor personnel.
--Contrary to the requirements of SP-2736A, and as a normal practice, cell levels which did not meet the acceptance criteria of the procedure were not recorded on the data sheets.
--The method of documenting water additions to individual battery cells was not uniform, resulting in at least one occasion in which l
the amount of water added to over 68 cells could not be determined adequately.
--On at least one occasion, a decision was made by a Itcensee supervisor to accept cell levels not meeting the acceptance criteria ,
of SP-2736A based on technical considerations not adequately documented. The inspector considered these activities to constitute a violation of NRC requirements (50-336/90-09-01). Licensee corrective actions will be considered during performance of the routine resident inspection program. This item is closed. . 9.0 Management Meetings Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of ! findings was also discussed at the conclusion of the inspection. No l proprietary information was covered within the scope of the inspec-i tion. 'No written material was given to the licensee during the l inspection period. l l
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f[-ff-f-sy3 - . = . . . , set, exeRtioQ me-- rom. f The NRC Region I has completed its fouowup to concerns that you brought to our atteation alleging Out the high radiation area gate alarm circuits wem not property depicted is the applicable drawings; (2), October 19,1989 alleging that the dermition of ' job supervisor
- contained in ACP-QA-2.06A conflicted with that in ACP-QA-2.02C and that the licensee did action when notified of the discrepancy; (3), March 15,1990 alleging (a) that a blue tag on the turbine buildmg component cooling water pmp motor brealer was insufficient for the preventative maintenance to be performed, (b) the wrong breaker was tagged in preparatio for preventative maintenance on the steam packing eahauster, and ( ) that valves wem manipulated outside the tagging boundary during maintenance on the instrument air sy2 tem; (4) March 2,1990 stating that a consultant had been contracted to identify electrical stating that a deficiencies in the maintenance shop; and (5(
drawing of the hydrogen analyzer power circuit was deficient. With regard to issue (1), we p~ found that the licensee determined a general deficiency in high radiation area gate a arm c
~
circuit drawings and has commenced action to correct this deficiency. We documented our inspection in inspection report 50-336/91-08,= Section 3.4 (attached). Your concern was substantiated. We intend no further action on this matter. We referred your concern described in issue (2) to the licensee for their resolution. We then - inspected the licensee's handling of the concern and documented our effort in inspection report 50-336/91-08, Section 3.7 (attached). Your allegation was substantiated in part, that part being that the discrepancy esisted, but the licensee has revised ACP-QA-2.02C to correc the inconsistency. We intend no further action in this matter. With regard to issue (3), we again referred your concerns to the licensee and inspected the followup and resolution of these items. Item a was determined by the licensee to be a proper interplay between operations and the work group involved in the preventative maintenance,
'%r concern that the breaker should be rackedet for the maintenance may be true, but may not have been required for the activity to be conducted. The licensee found for item b that the tagout used was appropriate for the activity being conducted and your allegation in th case appears to be unsubstantiated.' We documented our review of items a sad b in in report 50-336/9103, Sections 3.5 and 3.6 (attached). The licensee did not have enou h} Xf9/~,
4 l information to make a determination of the validity of item c. Appli cabl d e ocuments re lated l to the maintenance on the instrument air system were reviewed, but the licensee did not find , l any discrepancies. Please inform us if you have further information on this issue. - Otherwise, we intend no funher actions in these matters. ; l We understand for issue (4), that the licensee has a program to identify and correct electrical I deficiencies throughout the plant. Part of this program wu the load study to which you referred in your memorandum. At the present time, we are satisfied with the licensee's actions in this matter. We inspected the licensee's followup of the concem stat ! found that licensee engineering did not consider the drawing upgrade that i you proposed tote appropriate. We documented our inspection of this issue in inspection report 50-336,91-08, Section 3.3 (attached). We consider this matter closed. According to our records, we have completed our activities on all of your technical concerns provided to us prior to 1991. t I I We apprecipe you informing us of your concems and feel that we have been responsive to those concems. If you have any additional questions or if I can be of further assistance, please call me collect at (215) 337-5225. Sincerely; t s , hie'n/ - 3
- z. ward Wenzinge Reactor Projects Branch Attachments: As stated 1
t im Allegation File RI 94A-0107, update Allegation File Pd A-0003, update only.(B20, B14,B21.1, closecut) Allegation File RI-90-A-0033, closecut W. Raymond
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***** KING OF PRUSSIA. PENNSYLVANIA 19404 APR 81991 Docket No. 50-336; Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
Millstone Unit 2 Inspection 91-08 , This letter transmits the results of the routine inspection conducted by Mr. J. S. Stewart of this office from March 5 through March 7,1991, at Millstone Nuclear Power Station, Unit 2, in Waterford, Connecticut, of activities authorized by NRC License No. DPR 21. Our findings were discussed with Mr. Keenan and others of your staff at the conclusion of the inspection. The inspection consisted of a review of your responses to concerns referred to you by the NRC and is described in the enclosed NRC Region 1 inspection report. Your responses to the ' concerns reviewed were independently determined to be adequate, such that all of the issues inspected are considered closed. Your cooperation with us is appreciated. No reply to this letter is required. Sincerely, Mward C 'enringer, CMg Projects och No. 4 Division of Reactor Projects
Enclosure:
NRC Region I Inspection Report No. 50-336/91-08 h/h /hh/bV'~
e. a Northeast Nuclear Energy Company 2-t cc w/ encl: W. D. Romberg, Vice President, Nudear Operations R. M. Kacich, Manager, Generation Facilities Licensing J S. Keenan, Nuclear Unit Director, Millstone Unit 2 Gerald Garfield, Esquire Public Document Room (PDR) Local Public Document Rcom (LPDR) Nuclear Safety Information Center (NSIC) NRC Senior Resident inspector Staie of Connecticut bcc wiencl: Region I Docket Room (with concurrences) Management Assistant, DRMA (w/o enclosures) DRP Section Chief G. Vissing, PM, NRR J. Wiggins, DRP C. Hehl, DRP-N' r ,,2-, ,, - - -,< 4 a ,w-r- e ,- - -r,,-- ren,,-, a g e g- + , --r -~- e w- n ----re'-gm-
1 s U.S. NUCLEAR REGULATORY COMMISSION REGION I Report Number: 50-336/90-08 Docket Number: 50-336 License Number: DPR-65 Northeast Nuclear Energy Company Licensee: Facility Name: Millstone Unit 2 Waterford, Connecticut Inspection location: March 5 to March 7,1991 inspection Conducted: Inspector: TAV)
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- 3. S. Stewart, Project $r(gineer Dat6 Division of Reactor Projects Approved by:
b 7)' ' E. Kelly, C f D#te Reactor Pr is Section 4B Inspection Summary: Routine Unannounced Inspection on March 5 thru March 7,1991 (Inspection Report Number 50-336/9108) Areas Inspected: Routine inspection of licensee actions related to concerns provided to the licensee for followup and resolution. The inspector reviewed licensee documents and discussed activities vdth plant staff.
Conclusions:
Licensee docume:ts pertaining to the closecut of 22 concerns were reviewed by the inspector and discussed with plant staff. All of the issues reviewed were closed. Some minor weaknesses in licensee programs were identified that had significance. flh!l[h()/Lf/~~~ 1
I t
- TABLE OF CONTENTS 1
1.0 PERS ON S CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 2.0 IN S PECTION SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 2.1 Report Details ..................,................. 2.1.1 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to , E. C. Wenzinger (NRC),12tter Number A09188, dated January 1, 1991, RI A-0206 . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.1.2 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), letter Number A09066, dated December 3,1990, RI 90-A-0136 . . . . . . . . . . . . . . . . . . . . 3 2.1.3 (Closed) Licensee Correspondente E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09076, dated December 7,1990, RI A-0144. . . . . . . . . . . . . . . . . . . . . 4 2.1.4 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09163, dated December 21, 1990, RI 90- A-018 0. . . . . . . . . . . . . . . . . . . . 5 2.1.5 (Closed) Licensee Correspor.dence E. J. Mroczka (NhTCO) to E. C. Wenzinger (NRC), Letter Number A09166, dated December 21, 1990, RI 90- A-0204. . . . . . . . . . . . . . . . . . . . 5 2.1.6 (Closed) Licensee Corresponde.nce E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Leuer Number A09187, dated January 4,1991, RI 90 A-0205. . ........................ 5 2.1.7 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09166, dated January 21,1991, RI-90 A-0208. ......................... 5,
............................ 6 3.0 OTHER CONCERNS .......
3.1 (Closed) Configuration Control Discrepancy on NUSCO Drawing 25203-28500 ..................................... 6 3.2 (Closed) Configuration Control Discrepancy identified on NUSCO Drawing 25203 28500 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.3 (Closed) Configuration Control Discrepancy Allegal on NUSCO Drawing 25 203-29014 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.4 (Closed) Configuration Control Discrepancy Identified on NUSCO Drawing 25 203-32031 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.5 Tagout Discrepancy Identified During the Preparations for Authorized Work Order AWO-M2-89-02320 . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.6 Tagout Discrepancy Identified During Preparation for Work Order 7 AWO-M2 89- 10757 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . (Closed) Concern Regarding Definition of " Job Supervisor" . . . . . . . . . 7 3.7 7 4.0 M AN AGEMENT MEETING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
l '4 DETAILS t 1.0 PERSONS CONTACTED
- J. Keenan, Unit Director
- J. Riley, Maintenance Manager J. Becker, Instrument and Controls Manager J. Smith, Operations Manager R. Rowe, Electrical Superintendent
- R. Zisk, Nuclear Safety Concerns Engineer
- J. Criscione, Assistant to Unit Director U.S.NRC
- P. Habighorst, Resident inspector
- denotes presence at exit meeting conducted March 7,1991 2.0 INSPECTION SCOPE A number of concerns have been provided to the NRC related to activities at Millstone Unit 2. The dispositioning of these concems has involved providing the concern to the licensee for their review and resolution, with subsequent NRC overview to ensure the adequacy of the licensee's actions.
Seven documents containing the licensee's evaluation of the above concerns were inspected. Eight additional concerns had been previously provided to the licensee by the resident , inspector, as detailed in Inspection Rewrts 50-336/90-06 and 50-336/90-14, and these were inspected. 2.1 Report Details 2.1.1 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09188, dated January 4,1991, RI-90-A 0206 Issue 1 involved troubleshooting of the reserve cooling pump for the main turbine-generator stator following plant startup. Restoration of the system to a normal status required recalculation of the setpoint for the reserve pump and calibration of the pressure switch. A root cause determination for the improper setpoint included identined procedural weaknesses and improper procedural implementation by the technician. The inspector notef that the procedure has not been upgraded in the licensee's on-going procedure upgrade program, inspection Report 50-336/90-19 reviewed this program and documented the licensee's intention to have all procedures upgraded by 1992. In any case, the problem of being unable to secure the reserve stator cooling pump is of minor safety concern because the system was
4 2 t in operation and the generator stator was being cooled. The inability to secure the reserve pump was identified by operations and appropriate corrective action was taken to restore the system to a normal configuration. This issue is considered closed. Issue 2 involved comp 1.iance of the licensee's operational test program with technical specification requirements. The licensee's assessment that technical specifications were complied with is considered adequate and the issue is considered closed. Issue 3 involved the testing of the alarm associated with control element assembly withdrawal prohibit. The licensee's response that the testing is fully completed by existing procedures was considered adequate and this issue is therefore considered closed. 2.1.2 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09066, dated December 3,1990, RI 90-A 0136 issue 1 involved the replacement of the flow control valve (FCV) for radiation monitor RM 8262 and other troubleshooting activities during the June to August 1990 time period. The inspector reviewed a detailed work history associated with the radiation monitor and also reviewed work order M2-9044311 which was used to control the replacenient of the FCV. The replacement valve was purchased non-QA and upgraded in accordance with the licensee's commercial grade dedication process by documenting the upgrade in a non-conformance report, NCR 290-520. When the replacement valve was taken to the job site, the thread engagement was insufficient to place the component in service. The n: placement valve was ' then removed, rethreaded, bench leak tested, and returned to the job site for installation. These activities were not recorded in the work authorization. After the valve was installed, the system was returned to service, on June 28,1990, without the completion of the leak check specified in the work order. The leak check was performed on August 8,1990, and the work order was closed. During the perio'dfrom June 28 to August 8, operators verified that the radiation monitor was operational and recorded the flow once per shift except for maintenance periods. Furthermore, the radiation monitor is isolated from the containmen on a containment isolation signal making the significance of these occurrences minimal with respect to safety. The failure to conduct a complete commercial grade dedication inspection including checking thread engagement, failure to document the work required to install the replamment valve, and the restoration of the system to service without the completion of the work order are, in combination, considered a licensee identified weakness. Corrective actions included completion of the leak test, closcout of the work order, and discussions with licensee personnel by supervisors concerning the issue. Additionally, the licensee is in the process of
___ _ __ _ . ___ _ _ _ _ . . ~ . . _ _ 3 implementing additional controls to better establish coordination of maintenance activities with plant operations. These actions meet the criteria of 10 CFR 2, Appendix C, V.G.1 for the exercise of discretion; that is, the problems were identified and corrected by the licensee and v ere minor in nature. This item is considered closed (91-08-01). Issue 2 involved a concern that a bypass jumper tagout had been improperly controlled during maintenance controlled by AWO M2-90-08033. Review of the AWO reveals that an Instrument and Controls technician initiated a work order to repair a broken pin connector. When it was determined that the pin in question was not connected to anything the work order was canceled. No tagging was required. This issue is considered closed. 2.1.3 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09076, dated December 7,1990, RI-90 A-0144 Issue 1 involved the operation of the steam jet air ejector radiation monitor, RM 5099. The licensee has determined the adequacy of system performance during the periods in question. The licensee has also stated an intention to replace the radiation monitor with an improved design in 1991. This issue is considered closed. Issue 2 involved communications between an instrument and controls staff member and the manager at a staff meeting. In the case provided for licensee review, either (1) the employee failed to properly communicate the question, (2) the supervisor failed to understand the question, or (3) the employee failed to followup to an incomplete response. This matter is labor-management in nature and is beyond regulatory concern. Administrative Control Procedure ACP-QA-1.20 provides a mechanism for communication by way of informal correspondence which could have been used to ensure adequate discussion of the issue. This issue to considered closed. Issue 3 involved the control of work associated with troubleshooting the control circuit for the
- turbine bypass valves. Work Order M2-90-06498 was authorized on June 17,1990, to investigate a TAVE/ TREF alarm. As a result of the investigation pressure transmitter PT-4300 was found to have drifted and work order M2-90-07792 was written to calibrate the pressure transmitter during the upcoming outage. On September 10,1990, work order M2-90-09684 was written to troubleshoot a decrease in TREF which had caused a TAVE/ TREF alarm. A note on the work order identified the work that had previously been authorized to.
calibrate the pressure transmitter. The work orders were complete. 'Ihere was no need to maintain a second, redundant tabulation of activities. This issue is considered closed.
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- 2. i.4 (Closed) Licensee Correspondence E. J. hiroczka (NNECO) to E. C. Wenzinger (NRC), Letter Number A09163, dated December 21,1990, RI 90-A-0180, ,
The issue involved the management of activities involved in troubleshooting and maintenance of the "A" channel of wide range nuclear instruments during refueling activities. The licensee has prepared a spare wide range drawer which will be installed and tested to i determine if the spiking problem is diminished or eliminated. Recent inspection of this issue was documented in Inspection Report 50-336/90-22. This issue is considered closed. , 1 Issue 2 involved the licensee's management of a procedure change. Licensee actions with l regard to the concern were adequate and the issue is considered closed. 2.1.5 (Closed) Licensee Correspondence E. J. hiroczka (NNECO) to E. C. Wenzinger ; (NRC), Letter Number A09166, dated December 21,1990, RI-90-A-0204. The inspector reviewed the licensee correspondence and had no further questions. The issues are considered closed. 2.1.6 (Closed) Licensee Correspondence E. J. hiroczka (NNECO) to E. C. Wenzinger ; (NRC), Letter Number A09187, dated January 4,1991, RI-90-A-0205. The issue involved the control of work associated with the overhaul of the P5C service water pump motor. Because the activity involved both electrical and mechanical work groups, a number of station procedures and v.ork orders were used to control the activity. Work Order - M2-90-13312 was used to disconnect the motor leads, remove the motor, reinstall the motor, and reconnect the motor leads. A weakness in the coordination of activities allowed the pump motor to be removed without having the bearing oil drained. Having the bearings covered with oil presented no problems to the work being performed and when the condition was identified, the oil was drained to a collection drum. A number of enhancements to the procedures used in the overhaul have been implemented to ensure better coordination of future activities. The inspector notes that the procedures governing overhaul of the - condensate and heater drain pump motors were similarly revised to eliminate potential confusion in these related activities. This issue is considered closed. 2.1.7 (Closed) Licensee Correspondence E. J. Mroczka (NNECO) to E. C. Wenzmger (NRC), Letter Number A09166, dated January 21,1991, RI-90 A-0208. - The issue involved a weld made to a moisture separator reheater manway to stop a small steam leak in the gasket. The inspector reviewed the licensee's response and di='nwi the activities with the engineer involved in the work preparation. The licensee's response is considered to be both complete and adequate. The i'. sue is closed.
5 t 3.0 OTHER CONCERNS A number of licensee employee concerns presented to the NRC were given to the licensee Nuclear Concerns Program for review and disposition. Licensee actions in this regard have been reviewed by the inspectors. 3.1 (Closed) Configuration Control Discrepancy on NUSCO Drawing 25203-28500 A configuration control discrepancy was identified on NUSCO drawing 25203 28500, Sheet 198, which is a wiring diagram for the boronometer. The discrepancy was identified to unit supervision during a walkdown of the installation using the system wiring diagram. The completion of AWO M2-90-06407 corrected the drawing. System operation was not affected. This item is considered closed (RI-90-A-0107) 3.2 (Closed) Configuration Control Discrepancy Identified on NUSCO Drawing 25203-28500 A configuration control discrepancy was identified on NUSCO drawing 25203-28500, Sheet 50A-50D, which is a wiring diagram associated with the pressurizer spray valve positioner. The discrepancy was identified to unit supervision during a walkdown of the installation using the system wiring diagram. The completion of Drawing Change Request M2 P-110-90 corrected the drawing. System operation was not affected. This item is considered closed (RI A-0107). f 3.3 (Closed) Configuration Control Discrepancy Alleged on NUSCO Drawing 25203-29014 A configuration control discrepancy was alleged on NUSCO drawing 25203-29014, Sheet 9,- which is wiring diagram associated with the hydrogen analyzer panel C86. Drawing Change M2P-077-90 was submitted to upgrade the drawing but the cha.nge, when reviewed by engineering, was determined to be an inappropriate enhancement to an intemal wiring drawing. The drawing change request was canceled. Drawing 25203-31116 was referenced as the drawing appropriate for the requested change and this drawing was found complete. This item is considered closed (RI-90-A-0107).
- 3.4 (Closed) Configuration Control Discrepancy Identified on NUSCO Drawing 25203-32031 A configuration control discrepancy was identified on NUSCO drawing 25203-32031, which is a winng drawing for the "B" and "C" engineered safety feature area high radiation area gate alarm circuits. The alarms are provided as wanung systems for personnel that the gate is not shut. In their review of this item, the licensee identified that several high radiation
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. 6 area gate alarm circuits were not covered by drawings and work was initiated to prepare and i verify drawings for all of these circuits. The drawing drafts have been prepared and will be i hand over hand verified by June 1991. Final drawing preparations will be completed in i 1991. This item is considered closed (RI-90-A-0107). 3.5 Tagout Discrepancy Identified During the Preparations for Authorized Work Order AWO-M2-89-02320 A tagout discrepancy was identified during the preparations for Authorized Work Order AWO-M2 89-02320, which controlled the annual preventative maintenance on the steam packing exhauster. The item was reviewed by the licensee and it was determined that the i tagout was appropriate for the work being performed, although the job supervisor replaced red tags with blue tags before the work was done. This type of change is consistent with ; station procedure ACP-2.06A, " Station Tagging." The inspector had no further questions on ; l this issue (RI-90-A-0033 B20.2). 1 3.6 Tagout Discrepancy Identified During Preparation for Work Order AWO-M2 10757 A tagout discrepancy was identified during the preparation for work order AWO-M2 10757, which controlled the preventative maintenance on the "B" turbine building component cooling water motor. This item was reviewed by the licensee and it was determined that the job supervisor, in conjunction with operations, requested that the motor breaker be racked out prior to the maintenance activity. The licensee concluded that the interplay between the work group and operations was in accordance with station procedure ACP-2.06A, " Station Tagging" and therefore was appropriate. The inspector has no further questions on this issue (RI-90-A-0033, B20.1). 3.7 (Closed) Concern Regarding Definition of " Job Supervisor" A concern was raised that the dermition of " job supervisor" provided in administrative control procedure, ACP-QA-2.06, " Station Tagging," differed from the definition provided in A'eP-QA 2.02C, " Work Orders." The discrepancy had no apparent effects on plant activities and the licensee has revised the latter procedure so that the defmitions are identical. This issue is considered closed (B.14.01). 4.0 MANAGEMENT MEETING At the close of the inspection, a management meeting was held to summarize the inspection results. No written material was given to the licensee and no proprietary information was identified.
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- NUCLEAR REGULATORY COMMISSION -
i I REQtON I 478 ALLENDALE ROAD
.[ KING OF PRVsstA, PENNSYLVANIA 19404
-e SEP 2 6 1969 Allegation No. RI-87-A-0113 50-336 *
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1 Dear This responds to your undated letter to Mr. E. C. -Wenzinger, which was received in the Region I office on August 4, 1989. In it you contended that NRC ~ follow-up regarding your personnel safety and procedure compliance concerns, reported in Inspection Report No. 50-336/88-13, was cursory, inadequate, and improper. Regarding the " operator in attendance" tagging issue, inspector inquiries were unable to reconctie the conflicting claims involved. We acknowledge the Stephen Scace memorandum of Januar/ 29, 1988 which you enclosed in your letter. We note that the-event to which it refers occurred in a site outbuilding not-subject to nuclear tagging controls at the time. No connection relevant _to nuclear safety is discernible.' On going review of Millstone nuclear- power plant performance continues to confirm proper nuclear safety attitudes, and controls and procedures adequate to protect the radiological health and safety _ of the public. g= As is evident in this case, persons of good will may continue to differ. At this point, we believe that further NRC review of these allegations would divert limited inspection resources from other _ issues of potentially greater nuclear safety significance, and therefore be counter-productive. _ Since your ; t-letter presents no new facts or examples of licensee failure to control nuclear safety' activities, we intend to close your allegations as unsubstantiated. The NRC will continue to monitor licensee performance in the areas of work-I. control and tagout's of plant systems important to nuclear safety through-l routine resident inspector coverage, i I Thank you for your concern. { Sincerely, 1
-Deputy Director _of Reactor Projects l : Division of Reactor Projects l
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UNITED STATES j m NUCLEAR REGULATOMY COMMISSION i ( REGION I l a78 ALLENDALE MOAD
% = KING OF PautalA, PiNNSYLVANIA 16405 Docket / License: 50-336/DPR-65
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.t Northeast Nuciear Energy Company ATTN: Mr. Edward J. Mroczka Senior Vice President - Nuclear '
s Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06101-0270 1 Gentlemen:
Subject:
Routine Resident inspection 50-336/88-13 (5/3/88 - 6/13/88) This transmits the report of the above subject inspection of Millstone 2. The inspection findings have been discussed with Messrs. H. Haynes and J. Keenan of your staff. No violaticns were 'ited, and no reply to this letter is required. Your cooperation with us is appreciated. Sincerely, h i Lee H. Bettenhausen, Chief Projects Branch No. 1 Division of Reactor Projects
Enclosures:
- l. NRC Region I-Inspection Report 50-336/88-13
- 2. Appendix A, List of Facilities Potentially Af fected by Gamma-Metrics 10 CFR 21 Re .rt
- 3. Appendix B, Followup on Allegations Not Specific to Millstone Unit 2 cc w/ enc 1:
W. D. Romberg, Vice President, Nuclear Operations . l R. M. Kacich, Manager, Generat';on Facilities Licens.ing D. O. Nordquist, Director of Quality Services l S. 5 Scace, Station Superintendent F.plic Document Room (PDR) Leal Public Document Room (LPOR) Nuclear Safety Information Center (NSIC) NRC Senior Resident Inspector-State of Connecticut 0 L i e . .- , - - . , .,n ,, ~~ r -w..,m ,-e , ,, -n -,
4 U.S. NUCLEAR REGULATORY COMMIS$10N REC 10N 1 ( Report No. 50-336/95-13 Docket No. 50-336 License No. OPR-65 Licensee: Northeast Nuclear Energy Company .'- P.O. Box 270 Pa r t f o r:!, CT 06101-0270 Facility Name: Millstone _ Nuclear Power Station. Waterford, Connecticut inspaction At: Millstone Unit 2 Dates: May 3 - June 13, 1998 Ins:s: tors: Peter J. Habighorst, Resident Inspector Ocvid Jaffe, Licensing Project Manager NRP William J. Raymond, Senior Resident inspector Jams Trapp, bactor Engineer, Division of Reactor Safety Repo tin; Inspe:t:r: Pete- J. Habighorst, Resident Ir.spector Apprevec by: y MM%76-
.l . f M:Cabe, 3ief, Ceactor Projects Se: tion IB
[j6g{ Date . Its e:t'e- S m n 5/3 - 6/13/EBJ Ranort No. 50-336/88-13) A-ea s I"s:ected: Resu ne t.RC res'. dent , region-based, and specialist inspe: tion of. plant coerstic"s, sur,eillan:e, maintenance, radiation protection, physical secur' ity, cutage a:thities, allegations, Licer.see Event Reports (LERs), Safety Issue Manage ent Systen (SIMS) items, and :ommittee activities. Desults; No ur. safe tenditions were identified, Additional follow-up is warranted on a 10 CFR 21 Report concerning wide range nuclear instrumentation susceptibility to moisture intrusion (Detail 4.6) and allegations (Detail 8.4, and Appendix B). t I 847//p376~
Appendix B 2 6
- 3. 87-A-0113, Coatracter Wcrk Activities t
This section addresses the issues raised by a forn.er contractor empicyeo (alleger) regceding the control of work activities by a contractor at the site ' (Contractor A). The alleger's employment at the site ended in the Fall of . 1987. The alleger subsequently contacted the NRC by phone in the Fall of 1987 to discuss concerns about electrical tagging (see item B.3.6) and security lighting (see item B.3.3). The alleger also corresponded with the licensee in the Fall of 1987 to discuss concerns about electrical tagging. That con-cern involved two instances during work in an onsite warehouse that was being converted to Unit 2 maintenance offices. The alleger recuested a meeting with the licensee. The alleger contacted the NRC by phone in December 1987 to set up a meeting. . The issues discussed below were received during a January 13, 1985 meetfag with the alleger. These issues were referred to licensee nanage ent for ac-dre s sal . Tne l'censee tcd initiated action on the the alleger's cen:cen$ prior to NA: involvement and, based on audit findings, had initiated :crrec-t've actions. The status of actiens and licensee findings on all it.ves were periccically dis:ussed with the inspector by the licensee, and were svinariltd in an Ma/ 19, 1953 recorardam to the Station Services Superintencent. Sup-
. l e t r.t a l infer.ation on tne origital issues was bta.ned during varicus fol-icw-up telesh:re :0nversctions and during a follow-up inspector netting with '
- ne alleger en June 8, ;5SB. ine licensee contacted the alleger to set up a rec ting, which was schedJled af ter the end of the inspection perioc.
3.1 :: 3.e- Ir:tallation of ;E Ecle Lights in the Sirwiator Sui' ding f arting Lot in July 19s7 L is issue concerned the adecuacy of lighting installed wit'out anchor, bolt tetslates, in the simslater building parking lot. : Licensee revies founc that insta116 tion of the mounting bolts without a template was acceptable. The Contr.ctor A Superintendent, a :ertified professional engineer - civil discipline, used engineering jacgament to direct installation of the meur. ting bolts. A template was not required based on discussions of the installation with the light manufacturer,. who agreed that installation of 4 bolts on a 10 inch bolting circle was acceptable. The licensee concluded that no further actions were _ required and that no corrective actions were warranted. The inspector ooserved the parking lot lights and noted no obvious inadequacies with the in-i stallation. l The licensee stated that the alleger was terminated because of the dis - ' pute the Contractor A Superintendent over the use of the template and after removing the mounting bolts before the cement was poured. The alleger was reinstated after discussions with Contractor- A upper manage-ment. The licensee considered this to be an internal-Contractor matter. and concluded no further action was required.
/.ppendix 8 3 4
TLe inspector noted that this issue and its resolution had no ir.nact on ( nuclear safety. The inspector identified no inadequacies in the etselu- ' tion of this issue. 3.2 !ssue: Coordination of Work Activities to Remove Temporary Security , Lighting a This issue concerned elleged poor control of work activities as evidenced by a job where two contractor groups were assig.ned to do the same work. Licensee review determined that two onsite contractor groups were both assigned to work on temporary security lighting adjacent to a warehouse onsite. The job was first assigned to Contractor 8, who usually works lighting jobs. The start of work by this group was delayed. The alle-ger's employer, Contractor A, was then assigned to do the work by the Su tion Istvices Engineering Department at the request of Security. Knile preparir.g to do the work, Contractor A personnel noted that Con-tractor B was doing the jot; the work order to Contractor- A was then cancelled. The problem of work coordination originated within the Security Deoart-rent. Lit ersee actiors were a: ressed in a memorandum from the Station Ser 'ces :a;eritte tent to the Security Supervisor dated February 1, 10:3, aad f rom tre Ovality Services Supervisor to the Station Service Superin-tendent cated May 19, 1938. Security was cautioned to avoid duplicate w:rk asst;nttets in future jobs. , J The inspector noted this issue and its resolution had no impact on nuc-lear safety. The adequacy of lighting in the protected area has been previously reviewed by the NRC and noted discrepancies were ree?Ived , (NRC Region 1 Inspection Report 50-336/87-20). No inadequacies were 2 identifisc in the resolution of this item.
2.3 Issue
Worker Whole Eody Counts and Termination Exposure Repcrts This wcs a new allegation raised-by the alleger on April 25, 1983. This item invcived two individuals who worked at the Millstone Station in the 1983-1984 time period. Both workers reportedly stated to the alleger that they terminated employment at the site without obtaining an exit whole body count,-and without getting termination exposure reports from the. licensee. The inspector asked the alleger to provide the name and address of both individuals to allow further followup. After conferral with both workers, the alleger identified Individual A, who wore a-respirator while onsite and who_ also had a known cesium uptake. - No information was provided on the second individual. Individual A re-portedly did not want to talk-directly with the NRC and no address was provided. evaluation. This item was turned over to the licensee for review and i
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Appendix 6 4 The licensee provided, for NRC review, exposure Fistory files on Indi- ' vidual A. These showed that he worked at the site periodically from 1974 ' until 1983, with thr last work date onsite being 8/15/83. The licensee stated that Indivir. val A worked at the station during outages on Units : I and 2 and produced termination exposure reports corresponding to each ' work period. The licensee noted that the 7 termination reports were addressed to 5 different addresses used by the worker and noted it was possible the last report was not received at the last address'on file or forwarded in the mail. The licensee stated that another copy of the October 1983 termination exposure report would be provided if requested by Individual A. The inspector-reviewed the last exposure report dated 10/10/83 and cover-ing the period from 5/13/83 to 8/15/83. The inspector noted that the recorded quarterly exposures for all kork periods were low and well within re;.! ate y licits. The inspector noted further that a completed NRC Form 4 cated 5/13/53 was on file and properly reflected the exposure history record. The inspe: tor also noted that the licensee's health physics records show that Individual A properly comoleted the prerequi-site training, redical screenings, and respirator fit testing needed to wear respirators for work in radiological areas at Millstone. Li:e see records shew the results of whole body counts (WBCs) erfc ed for :ncivi:ual A. The last WEC result on file was performed on 5/13/53, wnich was tre e-trance count usually done by the licensee as a matter of pclicy (a.d rot because of an NRC recuirement). The licensee's re-ce rcs c;nfi c. +: t at Ircivicual A lef t the last work assignment at the station without a ter.mination WBC. The licensee concluded, from his review of the esposure history and wurk activities, that a termination KCC was nct needed to meet regulatory requirements based on records that shcw no airDc-ne exposure time (MPC-nours) were accumulated by tr,e indir vicual.
- r.is item is discussed further below.
- 'he inspector ncted that radioactive potassium and Cesium-137 were re-i perted in the kEC results for 7/23/E0 end 3/7/81. Radioactive potassium l is natu-ally occurring and is fcurd in all people. Cesium-137 is not naturally eccur ing and is produced by nuclear fission. The
- inspector noted that the cesiem levels recorded in 1980 and 1981 were 0.315 ncno-curies (NCi) (+) or (-) 3.092 NCi (at two standard deviations) in 1980; and 1.792 (+) or (-) 2,437 NCi in 1981. At these levels, the isotope-was present in trace amounts, just at the limits of detectability end f ar below the action level for follow-up investigation. The inspector noted that subsequent WBC results recorded on 11/6/81, 3/26/82 and 5/13/33 did not show any cesium. The' inspector did not determine the source of the uptake. The inspector noted that the licensee has'no re-cord of an incident report involving Individual A. Thus, the cesium uptake does not appear to be related to work at Millstone.
The licensee provided for inspector review all radiation work permits (RkPs) used by Individual A for work in the Unit l~ and Unit 2 radio-logically controlled areas during the period from May 13 - August 15, L
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l 1083. The inspector reviewed the reports along with the health physics s.>r,ey results showing the radiological conditions in the work areas cf interest. The survey results for airborne activity were reviewec in ( ; particular, as recorded in Air Activity Logt for dates corresponding to the RWPs. The surveys showed that airborne radiological conditions did <l not require the use of respirators because of pre-existing concentrations at the work site, Air activity results were at or below the 3.0E-9 , uS/cc kPC lirit specified in 10 CFR 20 for unidentified isotopes in restricted areas. . The inspector noted f rom the RWP records that most work activity by 16-dividual A in the Millstone 2 containment involved walkdowns, inspectiers, and installation of cables and conduits for neutron detectors and re-sistance-temperature-detectors. Some of the RWPs did require the use of resoirators and Individual A indicated respirators were used. The
.ce M rescirators in these instances appears to have been a precauti:n to s.rotect against possible irgestion of <adica:tive material r..t de air-b:rre curing the course of work such as during core drilling activities >
c+ containTert wllls with general area conta91 nation levels in the range of 50L-200( cisintegrations per minutt (dpm) per 100 sq. cm. Based on it.e rt:ielegical con::itions at the job sites werked by Individcal A, and c:tsider ; that licenset records show that no tGC hrs were rect 'ded for Ir:d i; ..d A, the irspector concludd that a whole body ccunt.us rot-ri: / red of hRC regulaticos. Specifically, no whole body count was necesst*y as part of tne bicassay assessments required by 10 CFR
- 03. ;C 3( a )( 3 ) . -
~
ine inspector reviewed Station Procedure $HP 4970, Internal Exposure Cc troi (Eicassays), Revision 4 dated 4/22/86, which establishes the ' licensee's e M Lsay pro 9rarr and sets.the criteria under wnich W3Cs wil1 De ptrfo ted. SHP 4970 requires a W3C if it is cetermined that a _ limit cf C effe:tive MPC-nou-s is exceeded for a worker. This is consistent wish 10 CF; 20. Additionally, SHP 4907 requires " routine" WBCs for all pe scrrel inued desitretty at the site at the start and end of empl:y .e t, ano at least once per year. These whole bcdy counts are (in part) a 5:reening Masurement used to validate _ the acequacy of other cor.trols establi:te: for work in radiological areas. The performance ~of such W2:5 ' is a licensee administrative practice that exceeds NRC requirements. Eased on ',he above, even though SHP 4907 was not met in this instance, there is no safety significance and there is no violation of NRC. require-rents. The li:ensee stated he would perform a.WBC on Individual A if he req;ested one. The alleger was requested to relay this inforraation' 7 to Individual A. " Tne licensee stated that, since the reported cesium contamination was below their investigation threshold of 20 nanocuries, no further follow-up would have'been taken when it was first noted. The licensee stated' that all contractors are made aware at the start of employment that it ' is expected workers will get a whole body count upon termination,- There is no mechanism in place to enforce this requirement in all cases.- The - F
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Ap;. dix B 6 licensee estimated that about 5% or or fewer workers leave the site with.ut and exit W3C. The licensee feels this is acceptable since, at t an assessment tool, the 95% of the workers who do get the termination whole body counts confirm the success of the respiratory protection pro-gram. The licensee stated that all persons are evaluated per 10 CFR < 20.103(a)(3) when required. The inspector reviewed the inspection record for all three Millstone units and noted that recent NRC Region I reviews .
~
of the internal exposure controls have found the program to be acceptable.. , Minor deficiencies have been noted, but no inadequacies have been iden-tified in the respiratory protection and bicassay assessment function, i During a June 8 followup meeting with the inspector, the alleger provided-additional information about Individual A. The alleger stated that, during Individual A's last work assignment at the site, he was working in containment without a respirator when plant operators turned on a fan which caused ce*tamination to be blown on the voriers. The alleger had no furtter inferration from Individual A as to whether Individual A was found to ha.e contamination on his clothing or skin, or whether he showed positive counts on a nasal smear. Th4s ratter was again reviewed with station Hp personnel, tho stated that an incice-t rec:rt would have been written for any instance involving sL4n co tar 1 ction or a pc ssiole intake of radioactive material. No incident reports involving Individual A are on file. The inspector could net pursue this matter further without actitional specific informaticn directly f ec- !rdivid;al A. The inspector sent cessa;es to Individal A via tne 411eger. Indi.tcual A had not contacted the inspectoa as cf . Jane 13, 195E. Based on the above, the inspector' concluded no'further NRC actions are warranted on this matter. 3.3.. Issue: Categcry 1 Welcing by aa Unqualified Welder } This issue involved a concern that Contractor A, who employc4 the alleger, was using en electrician to perform Category I welding traice the con-tainment. No ot er specific information 6 35 provided as to the name of tne worker, the plant' involved, or the r.ame or type of plant systems affected. Unit 2 was assumed to be the af fected unit since that unit t was in an outage at the time the information was provided. The inspector reviewed activities in the Unit 2 containment on January 23, 1988 for welding by Contractor A. None was identified.. The issue was referred to the licensee for action. Subsequent routine in,spections of Unit 3 and Unit 2 outage activities did not identify welding by Contractor A. Licensee follow-up identified information which partially corroborated the alleged contarns, but-did not show safety significance. Contractor A does not have a Quality Assurance program and is not used by the licensee to perform nuclear sa fety-related work. The. licensee developed a list of work assigned to Contractor A at:all three units , since October 1986. Until recently..most work by Contractor A was per-forn.ed for the-Station Services Engineering (SSE) Department, whose re-
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Appendix B 7 t sponsibilities include non-nuclear projects su:h as grounds maintenance, and maintenance, installation, and modification of outlying buildings, t Ccntractor A was also used in the May - August 1987 period to install ' and modify gates for high radiation areas (HRAs) in all three units, as . discussed further below. During the 1987 - 1988 outages, Contractor A , was used as a general laborer force under plant personnel supe
- vision to (i) move shield blocks around reactor components - Unit 2 contain-ment; and (ii) perform material accountability control in reactor work
areas - Unit 3 containment. No Category I work has been assigned to Contractor A and no safety class welding has been performed by its em-ployees. Contractor A carpenters and electricians have welded in non-Category I work activities. Qualifications for electricians _were based on welding. training obtained during union apprenticeship. Also, Contractor A car-penters we'dt: stif feners on HRA gates to provide added reinfc* cement, wa en reeced. Thtt work was completed with1r. the plant buildings and was contrciled by work orders. One Contractor A electrician performed two ron-Category I welds outside plant buildings: one job involved relocating a security intrusion detection system; the second involved mounting a micr wave a-tenna on the side of the condensate polishing facility. The li:ensee cerciuced thit ccmpletion of these welding activities in ac-cLrdance with general construction work practices was acceptable, ho further h;: a:tions are deemed warranted or are planned on this matter.
'he iaste: tar reviened station adtinistrative requirements to determine n actner any c*rtt: ~ 5 shou'd have been appliec to the work. Th's review .
inclured see:1fic censideration of ACP-QA-2.03A, Non-Category I welding. it.e icentified welding activities did not involve ASME or ANSI code work; the w:rt could not af fect Category I systems; and the work did not in-
,cive ar a:tivity for which a completed weld documentation package woUld be re:u' red c- desired. Based on the above, the requirements of ACP-QA-2.03A were ceered not applicable to welding Contractor A performed. No inace;uaciet were identified.
The inspector reviewed the licensee's actions on this issue end identi-fied no inadecuacies. Inspector review of HRA gates during routine in-spections in all three units has found the gates to.be sufficiently strong to provide an adequate barrira. No inadequate conditions were idertified.
3.5. Issue
High Radiation Area Gate Modifications Tnis issue concerned the controls applied to werk done to modify and install 1alarmf on about 70 high radiation area gates in all three units. The alleger was responsible for completing the work in the March-August 1987 time period. .The alleger stated that an inexperienced engineering technician, newly hired by the contractor, was assigned responsibility for the job. That-individual's designs were reportedly aborted af ter
,. about two weeks work and af ter working on two gates. The job was al-i
Appendix B 8
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legedly perfor ed based on " blackboard" designs withcut writter. guidance and criteria. Proble rs reportedly included lights that were ret needes t and a 56,000 restock charge when another type of light was sele:ted, and the use of alarm bells that ran on 24 vac and required the use of step down transformers. This item was referred to the licensee for review. , The inspector noted that the High Radiation Area (HRA) gates and associ-ated alarms are not nuclear safety-related systems or components. The inspector asked the licensee to specifically address whether installatico work associated with the design not used resulted in radiation exposures which could have ' een avoided if better guidance had been provided. The inspector reviewed the results of the licensee's review of this issue, independently reviewed the detailed design change package under which the work was accomplished, and interviewed the individual responsible - - - for completion of the modifications as the designated plant project engineer (PE). Tne wert invelsed the rodification of esisting gr.tes and the in:tal'atic. of rew gates that provided t,arriers to the entrance to desigrated Hgh radiation areas in the three Millstone units. The work was controlled with c' tailed gs,ir.ince provided in Automated Work Orders and in tne fol-lo,ing plant e:ign cnarge reauests (PD:Rs): f %R l-105-E6, HEA Ga'.e Alarns & Warning Lights (MP1) FZ R 2-004-E7, HRA Gate Alarms & Warning Lights (MD2) FMR 9-D-002, KA Gate Alarms & Warning Lights FX; 03-EE-H2, HRA Wire W sh Gates . I*. licensee is re wired to control access to high radiation areas'as cefired by 10 CFR 20 ard to lock the access ways if the d0se r:tes in-volve radia;icn levels in excess of 1 Rem /hr. The eain purpose of the! PXR was tr act warning lignts and audible ala-ms to existing and new gates to notify personnel that the gates were not properly secured - (loce ec closec) af te passing through the gates. This acticn v.as in rescor se to licensee and NRC-identified concerns that the HRA. gates ere being left oper following access to the rooms. The PDCR was also used to add gates to new areas based on surveys by health physics personnel, and to stiffen existing gates. In addition to the guidance provideo by the PDCRs, the folicwing references also provided written guidance on gate fabrication, modification and installation: Stone & Webster Specification 2199.241-932, Specification for Wire Mesh Doors NUSCO Electrical Installation Specification SP-EE-076 Field Sketches LPKA 120286 and LPU-B 120286 Various AW0s for installation activities during May-August 1957
Appendix B 9 The design changes described in the PDCRs were prepared by an engineering technician in the Station Services Enoineering (SSE) Department and were t reviewed and approved by the licensee's engineering staff for ea:h unit
- as required by station administrative procedures. The review by the unit staff anc, Plant Operation Review Committees determined that the change, ,
while intended to ireprove ccmpliance with the requirements of 10 CFR 20.203(c)(2) and Technical Specification 6.12.1, did not involve an un-reviewed safety question per 10 CFR 50.59 or adversely affect the opera- - tion of safety systems or structures. Power for the circuits would be provided from non-class IE supplies and the gates would not impact seis-mic walls. The function of the alarm circuit was to provide audible and visual inc'ication that the HRA gate was open longer than 10 seconds (the time delay allowed for normal transit without alarms). An override switch was provided to bypass the alarm function for periods when the cite would be left open for extended periods to allow movement of mate-r:als. The t' arm we.;1d also activate, however, if the function switch
.&s n:t returnec to the " auto" position when the gate was closed.
The licensee determined that the original PDCR design was develeped in late 1986 with the intention of using explosion proof alarm lights simi-lar to these already installed on existing gates. The explosien proof lights were ordered from funds available in the 1986 operating budget. The initial circuit design was developed to use, to the extent possible, com onerts already available in station stores, including transformers, 2' VAC alarm bells, and aluminun shield ( ALS) cable. The ALS was chosen to minimite worker radiation e aposure since it would allow installation of the circuit without conduits. Additional bells and relays '.ere or-dered as necessary. As the design change proceeded, it was concluded that the eFplosion proof lights were not need2d and strobe lights were ordereo instead. This action was taken even though there was a 56294 restock charge on the explosion proof lights, because there was still, a net savings in e.x:ess of $10,000. Since the explosion-proof lightsi were never installed, there was no additional exposure required for the job as a result of tne change. The wcrk was originally assigned to Contractor B. The licensee deter-mined that the work was proceeding too slowly and the Station Services Engineering Departrr.ent reassigned the job to Contractor A. The licensee determined that the original circuit design as described in the approved PDCRs would work. Proper functioning was demonstrated by construction and bench testing of a test circuit in the shop. However, contract electricians (including the alleger) recognized that irnprove-ments in the circuit design would reduce the number of cable terminations needed and would result in less ALS cable being installed (estimated at 3 to 12 feet per gate) and thus reduce the work. time required in radi-ation areas. Even though the gates controlled access to HRAs, the typi-cal work area for the gates was the immediate area of the gate and the nearby electrical panel, which were not in "high radiation areas." (Personnel exposure required to do the job is discussed further below.) The modified circuit in its final form included the use of an additional 1
i Appendix B 10 I e instantaneous relay to replace one contact from the gate closure limit switch proposed in the original cesign, inspector interviews with the t project engineer determined that he proposed several interim configura-tions that were found unacceptable during follow up reviews with the alleger. Ultimately, the final circuit design chosen was the one pro- , posed by the alleger. The licensee stated that-the selection of the final design was left to the discretion of the Station Services Engi-neering Department, since unit engineering bad determined that circuit ' changes were conside ed minor in scope and would not impact the conclu-sion of the 10 CFR 50.59 safety evaluation. The licensee reviewed the radiation exposures for the job and concluded they were not excessive. No exposure was incurred on Unit 3. The work on Unit 1 expended 0.835 man-rem for 264 man hours, for an average dose rate of 3.3 mRen/hr. The work on Unit 2 expended 0.710 man-rem for 236 nan hours, for an average dose rate of 3.0 mrem /hr. A tabulation of incivicual expcsures for all contractor personnel who worked on the .icb (wnich incluced esposures for all work during the period and not just the HEA gate job) shcwed persontal exposures were not e,cessive. In-spector review of the tabulation noted that individual quarterly expo-sures we e less than 250 rRem in all cases, except for one incividual w4 th a maximum c3arterly exposure of 440 mrem. The licensee concluded, based on raciation work permit (RWP) records for the installation of the first few ;ates that expcsures were not excessive. Furtner, the licen-see's acministrative exposure limits we e not increased for any worker during the job. Inspector review found to inadequacy in the licensee's cen::.s':r. . The licensee further concluded that, in spite of the additional dose-szvings reali:ec in going from the original to the final circuit cesigr, reaser.aole measures were taken to minimi:e radiation exposures for the; design change. Inese reasures included use of ALS' cable instead cf con'- duits to recute installaticn time, selection of pcwer supplies to mini-mize time spent in radiation areas,_ testing the preliminary design in l the shop and pref abricating and testing materials as much as possible in the shop to minimize installation time in radiation areas, and licen-see supervisory monitoring of work' progress and reassigning the job to another contractor when the first contractor was deemed unacceptably slow. The inspector identified no inadequacies in the licensee's findings. The inspector notet there was a difference between the licensee's and the alleger's version of the job. The alleger stated that the critical circuit-design did not work and modifications were required _on the first two gates modified in the plant. While the inspector did not resolve the different versions, he did note -that based on the low dose expended for the entire job,_ rework of two gates in the field would not change the conclusion that exposures were not excessive. 4
+
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Appendix B 11 Licensee review cnneluded that the engineering technician assigned as project engineer to the job had adequate experience to perfort. ihe work. Tne technician is a contractor personnel employed by Contractor A and \ assigned to a staff position in the SSE Department. The engineering technician worked for 3 years during Unit 3 construction and startup for the architect engineer. The technician gained experience in the elec- ' trical discipline while working with major electrical components and through involvement in switchgear testing. The technician also worked for 1.5 years in the SSE Department working on projects involving the electrical discipline. The licensee did note that the technician had some dif ficulty administrative 1y coordinating the setups necessary for the first HRA gate job in Unit' 1, which involved allowing sufficient lead time to process tagouts and RWPs so that the work could start on time. This dif ficulty stemmed from his lack of experience in processing the administrative controls. The licensee concluded this inexperience did not cause unnecessary radiation exposure. The inspector identified no ina::Equacy with the licensae's conclusion since delays in starting work or in obtaining the prerequisite tagout would not result in radiation exposure. H o'.. e v e r , i n addition to the above described dif ficulties with the licen-s'e's acmit,istrative controls, tht inspector noted that other prcblems occurred in fcilowing the requirements of the tagging procedure, ACP 7.;6:, as cisesssed furtner in Issue 6 below. While the tagging for the hRA gate jobs was found to be done safely, it was not done in full com-cliance with the ACP, as follows: (i) operator-in-attenesnce tagging was perforr'ed by contracter electricians on some occasions, which did not riet the requirements of tagging procedure ACP-QA-2.06A; and, (ii) single tags used for multiple gates (7 primary breakers covered 5 gates each) were processed witho';t using the SF 210 continuatian form required by tre ACP. Further, during interviews with- the contractor technician, the inspector no;ec tht the technician had been responsible for the initid1 preparation of the 10 CFR 50.59 safety evaluatien for the PDCRs, but had not had training in the 10 CFR 50.59 process. The inspector found that the contractor had beccme f amiliar with the administrative requirements by reading the associated aoministrative procedures, in response to inspector inquiry, the licensee stated that contractor personnel are provided on-the-job training in the station administrative rnquirement as needed, and that this training involved reading of the associated administrative procedures. That was not formally documented, l Although the inspector identified no safety issue relative to the HRA , job, and no inadequacies were identified in the completion of the PDCR per the requirements of NE0 3,03, the inspector identified this area as meriting further NRC review to determine the general adequacy of trair.ing provided to contractor personnel on licensee administrative requirements. Even though no safety-related (Category I) work is assigned to the SSE Department, the inspector expressed the concern that the licensee needs to formally document required training to assure contractor personnel l l l l 1 7 W9 y-'n wy --w e,ww wy-7-'- g F *-v-w fr e- e-y a -*eeT
are fully f amiliar with all administrative procedares and require ents they are evpected to follow. This area will be reviewed further on a subsequent routine inspection as a potential elemens of licensee per-( formance. In summary, while followup of this issue did confirm the alleger's statement in part, the licensee concluded that guidance was provided and controls were applied suitable to the activity. Further, reasonable ' ef forts were taken to maintain radiation exposure hours less than the ALARA goal of 1 Rem per unit. The inspector agreed with the licensee's conclusions. The inspector noted that it is neither unacceptable nor unusual for approved desigra to change to reflect improvementt identified in the interaction between design and implementing groups during the design change process. No ur. acceptable conditions were identified during the inspector's review. 2.6. Issue- Ad% rence to Cc trols for Electrical Sv,itchirg part A: Exercising 450 volt breakers without tags or AW3. This issue concerned alleged actions, on one occasion in 1987, by the Contractor A Su;erittercent to ranipulate 480 vei t cir;.u t creakers on s a a rebouse carel in orcer to trouble-sbe:t a crobl e with the air con-c i t i o r. i n g . This actic* was ta6 en witneut a work c-cer or tags to control the activity.
- ?:
r- F:, no a are c se electrical circuit. On a second occasicn in 1957, the alleger processed an AWD and tagging order c c., to rercvc n electrical wire 50 as to cliew installation o' a win-i n a w a r e r.o u s e w a l l . The wire was moved instead tj station r.a?ntenk a n c e p e r s o r
- e l w i t ho.t t a g s o r a n A',,'O . ine alleger was fired during an' cnsuing argument en the control of Contractor A work activities and, he feels, f or f cilev ing ec .inistrative re:;uirerents for wnich he vias held responsible. During this incident, the Contractor A Superintendent al-legedly stated that he cid not come uncer NU or NRC jurisdiction for following aaministrative requirements.
These items were referred to the licensee for follow up and dispositioning and to address the following: (a) assuring work activities are conducted per established procedure controls; and (b) assuring the control of con-tractors is appropriate and that station policies are followed. Part A: Licensee review of the first issue concluded that the Contractor A Superintendent operated 480 volt breakers in the Unit 2 warehouse without a tagout, but a tag was not required for the specific activity. The licensee determined that, on August 17, 1987, the Contractor A Superintendert worked with a representative of a local air conditioning ( AC) company to investigate a report that air conditioning in the main-
l - Appencix B 33 ter.ance shop, recently converted f rom a warehouse, was not operi. ting. The ew tir conditiening unit had been recently connected, in an unre- t lated action to a 4E0 volt distribution panel using a plug-in circuit ' breaker. In this application, the breaker is routinely used as a switri. Other circuits fed from the same distribution panel included welding , machine supplies, which were hard-wired to fixed receptacles. With the AC representative at the cooling unit, the Contractor A Superintendent manipulated one circuit breaker to turn the AC on. The designated AC ' circuit had a temporary label. The unit was left on for the remainder of the day and then shut off using the same breaker. The licensee con-ciudsd that the actions by the Superintendent were proper since no "wori" was done on live circuits and no tagout was required or needed. The inspector reviewed the statement of applicability for ACP-QA-2.06A, the licensee's station tagging procedure, and identified no inadequacies. The inspe:t:r totec that there is a differen:e between the allager's anc the licenset's sersien of the activity. The alleger stated that the status c' tre air ccnditioning unit was not known by the the "ontractor A Sucerintercent when he operated 6 or 7 breakers in an attempt to start the unit. The inspector could not resolve the different versions. The inscector c:*tacted the alleger to obtain any additional specific in-f:r : tion t'at sh0ss activity on August 17th involved "wo"k" on live cir:uits. Lo aeditional information was available to resolve the c"f-ferences. 2* s:a: tor review of this natter did not substantiate ine ll-leger's input. Pa-t E: L':e :e revie , of this issue was d:: rented in recoranda cated 1 U 7D 57, 11c b57 ard 5/19/8E. . The licensee initiated reviews of this issue in -esp:nse to ccrcerns raised by the alleger in a 10/16/87 letter tc the Stati:r See, ices Superir,tendent. Tne licensee's review :enfirmtd essenai facts ir the alleger's staterrent, as follows. On September . E,19E7 car;t te s re:uired an electrical wire to be moved in order tot install a ,,inc:s in an exterior wall of the maintenance shop. Tne al-leger, actug as Co-tractor A electrical foreman, obtained a work order (M2-E7-103K) and "olut" safety t sgs (clearance 1569-87) to perform the work - acting arparently independent of direction from the superintendea.t. The tags were hung on Thursday, September 10 to do the work on Friday, but the work was done on Wednesday by NNECO maintenance personnel. The work was done by NNECO after the Contractor A Superintendent suggested the utility could take the required actions at a cost less than what would be charged by Contractor A. The wire / conduit was moved by NNECO personnel without an authorized work order and by using a modified ver-sien of the " operator-in-attendance" controls specified in tagging pro-cedure A:P-CA-2.06A. (This item is discussed further below ) The acticn to rnove the wire was completed before the alleger's tags were hung on September 10. When the alleger learned on Friday, September 10 that the work was already done, an argun'ent occurred with the Contractor A Super-intendent en the control of work. During this meeting, the Contractor A Superintendent terminated the alleger for insubordination and for
Appendix B 14 4 charging tire for electricians who did net particip2te in the jcb since the only work activity completed was by the alleger to process the AWO and tags. \ The licensee determined that maintenance personnel completed the job on September 9 in about 30 minutes using 3 workers to control the breaker < on Warehouse 4 Lighting panel 28. This power source is fed from the non-safety related Flanders line, Workers were posted at the breaker and within line-of sight of the work area and the breaker to ensure the cir-cuit reroained de energized while the conduit was removed and the wire - was moved down to allow installation of a window. The actions taken met the intent of " operator-in-attendance" tagging permitted by Section 6.1.9 of ACP-QA-2.06A, b9t Tag Log Sheet SF-210 was not used as required by the procedure. The inspector noted that information recorded on SF-210 would have included identification of the equipment covered in the order, its location, and applicable work order. No other specific in-for Ation on the required positicr of the brealer would have been re-guired. The inspectcr noted further that had the job been covered by Tagging Crder 1569-87, as initiated by the alleger, Blue Tags would have been used (which essentially releases the equipment to the person re-sponsible f or the work) and would have allowed the breaker to be posi-tio*ed "ts reavired" by the work party leader. Thus, in the cor.trols ef fected y either the Blue lag or operator-in-attendance methods, the desired positics of the breaker is lef t to the oiscretion of the work party. I . restante te the alle;er's concern, the licensee had the orsite Indus-trial Safety Cesartment (150) review the actions taken in this particular - situation. That review concluded that safety was not compromised and that the assumed risk in the operation was reasonable and controlled. Since ACP-0A-2.C6A was written primarily for the control of inplant equiD rent, and since lets stringent controls than are required by the ACP may be cesirable for work in outlying plant areas, the ISD recom- ' L rended consider 6 tion be given to modifying acministrative procedures for nor plant syster eaintenance work to recogni:e the method used. lo this end, the Station Superintendent issued controlled routing 6926 to address ( tagging in outlying buildings. Actions to draft a new procedure were , I in progress at the end of the inspection period. Notwithstanding the above conclusion on the saf ety of activities en Sep- ' , tember 9, the licensee concluded the' actions were not completed as re-' l quired by station procedures. Corrective actions were taken by the Station Superintencent in a memorandum (HP-11440) issued to the station on January 29, 1988 which reemphasized the need to' follow the require-L ments of ACP-QA-2 06A as the only currently acceptable process fer safety tagging. The inspector identified no inacequacies in the licensee's conclusions or corrective actions. l
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Apptndix B ]$ 4 The licensee evaluated the performance and statements made by the cen-tractor superintendent to assess his attitude toward complying wd th ',' station procedures and policies. The licensee conclude (i that the super-intendent accepts the need to comply with procedures and appears to clearly dif ferentf ate between his authority as superintendent over craf t ' personnel and his relationship to NNECO procedures and supervisory over-sight. The licensee's conclusions were confirmed by the inspector in an Independent interview with the contractor superintendent, To further address contractor adherence with administrative controls. - the Quality Services Department ,,erformed an audit of Contractor A acti-vities. The results were reported in QSD Report 050-88 4243 dated 2/4/88. The audit addressed, in detail, 119 of the 269 work orders involving Contractor A and issued from 1/1/87 to 1/27/88. The audit showed that: the Contractor did not perform work on safety-related equipment, and that the contractor worked on non-safety-related equipment under dire: tion of the Station Services Departeent with the approval of ths Operations Departrent. Also, for 66 work orders, no tags were used when tagging would apparently have been required. Instead, work in all three units on the high radittion area gate modifications was performed with an im-proper use of " Operator-in-attendance" tagging where an electrician as-sure: proper positioning of the C/ feeder circuit breaker at a Danel - in tre vicinity of the work site, but no SF 210 was used. While cc trol of ine work was maintained, the method used did not meet the requirements of the ACp. The licensee also determined that, in some cases where a - werk order indicated no tags were used, one tagout for one gate was used for werk on another gate when a conmon power supply for the alarms was
- invol.ed. Minor administrative problems noted on the completion of the-work order forms included not initiating revisions to tagging reoutre-ments listed as undetermined, and one work order with the tag "verifica-tion" and " cleared" blocks was net. signed of f,
{ Licensee review attributed the tagging problems to improper implementa-tion of the administrative requiremehts based, in part, on a misinter-pretation of the work order format and misunde standing of administrative requirements. Licensee manigement reviewed the work activity associated with the high radiation area (HRA) gate job and concluded that proper administrative controls were generally used and _all work appeared ade-l Quately controlled. However, further actions were deemed necessary to improve contractor understanding of administrative requirement s and oversight of contractor activities by NNECO personnel. " Additional licensee corrective actions-included supplemental training to be provided by Station Services Engineering (SSE) on the proper use of SF-210 for _" operator-in-attendance" tagging. The SSE supervisor, by memo MP-5-GSS-88-6 dated 2/8/88,. was directed to- improve tagout controls. (Corrective actions documented in a-namo dated 2/25/88 included emphasis on operator-in attendance tagging and on work directions;) The SSE Supervisor held a meeting with contractor personnel to review and_ clarify J-j tagging requirements and to address in particular the actions taken on !' a- . _ : an~ _ -, , --- , _ . -,- . .- ,,_ a, . _ _ _ . _ _ ,
Apperdix B 16 I the HRA gate job. Actions were taken to revive the list of $$E perse -el l authe-1:ed to sign work orders, and to limit that function to id200 per- t sonnel effective May 15, 1938. A measure is to be formalized whereby ' S$E supervision will be assured that contractor personnel perform work i in accordance with procedures and policies. Additionally, the Quality l Services Department will revise ACP-QA-2.026. Work Orders, to clarify the requirements on the authorizations needed to sign off " tagging veri-fied" and " work completed" sections of the work order forms. The licen-see is also drafting a ner procedure for the control of tagging opera- { tions in outlying buildings (Controlled Routing 6926). The above actions have either been completed, or were in progress with an expected comple-tion date by July 1988. No inadequacies were identified in the licen-see's conclusions or corrective action plan, based on independent in-spector review and interview of personnel. In surnary, the inspector noted no nuclear or personnel safety cercerns for eitner the specific instance cited by the alleger, or in tr.e HRA gate , tagging discrepancies identified by licensee audit. However, licensee actions were appropriate to assure established work control.and tcgging , rrocedures are soth fully understood and followed. The corrective ac-i tions represent improvements in the licensee's programs to control work a:tivities by a contractor who performs non-Category I wcrk. These-ac-tiens are also appropriate to assure that contractor personnel who might beccre iraolved in Categc y I werk are knowledgeable of accinistrative c:ntrols. For this reason, the inspector will follow completion of the licensee corrective actions during routine inspection.
3.7 1ssue
Control of Contractor Activities This issue invoked the cuestion of whether the centractor superintendent folloned station procedures or wielced undue influence on the site tased on alleged improprieties between the contractor end a licensee supervisor, lhe alleged improprieties' involved matters that were not related to nuclear safety at the site. The information provided by the alleger was referred to licensee nanagement for follow-up.' The inspector asked the licensee to address whether the control of the contractor is appropriate to assure station procedures and policies are followe6. Licensee review determined that oversight and control of the contractor activities is appropriate and that station procedures are generally fol-loved. This conclusion was based on reviews by the lead licensee in-vestigator and on the results of an audit of- the contractor activities. by the NUSCO Quality Services. The exceptions concerned the noted-prob-lees with processing work orders and tagging practices associated with the HRA gate job, and with the job to move the wire in the Unit 2 main-tenance shop. Adequate control is assured through active involvement - by the licensee supervisor bcing at'the job sites on a daily basis. The licensee supervisor is accountable to NNECO management as evidenced by routine interactions which provide appropriate guidance and management direction on activities in the-department. The licensee stated that -L 9 7 yt t--*'9Yp*7-g -t=4-@ ragw-Tie str-rr psmybwy+ty-e mp y N'y--'r-gu-4 yav7*+7 y-.eww- z. -*"gy re-; w--3Th agi+ P if 1't4~- W " - - M -W -T- '"*v WI-*" 1
Appendia B 17 concerns about elleged improprieties in the relationship between the centractor and the licensee supervisor were not substantiated. The lic- ' enses stated that the statements made by the contractor superintencent regarding the need to follow procedures were made during an ar were not indicative of the contractor's observed performance. gument and The in- e spector interviewed the contractor superintendent and neted that the views he empressed regard adherence to station procedures and policies as important, especially in his role as supervisor. , The inspector had no further questions on these concerns. 4 9 I
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d UNtTED STAfts j [ NUCLEAR RCULATORY COMMIS$lON L p CEDONi
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[ 476 ALLENDALE ACAD KING OF PRUSSIA, PENNSYLVANIA 164ot OCT 111989 Dockat No. 50-336 . E. ' Northeast Nuclear Energy Company
- ATTN: Mr. E. J. Hroc:La '
Senior Vice Presid(nt - Nuclear Engineering and Operations Group P. O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
Inspection Report No. 50-336/89-13 ' This letter refers to the special allegation team inspection conducted during the period July 10 through 21, 1989, by Mr. J. P. Durr and other tnembers of the regional an! headquarters staff. The inspection was conducted at the Millstone Nuclear Generating Station, Unit No. 2 and consisted of document reviews, personnel interviews and observation of ongoing activities including equipment installations. The results of the inspection are documented in the enclosed inspection report. The preliminbry results were discussed with you and members of your staf f on July 21, 1989. Several of the activities inspected were apparent violations of NRC requirements and your operating license. of Violation enclosed as Appendix A to this These are set forth in the Notices letter.
, The violations have been categorized by severity level in accordance with the " General Statement of
- Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (tr.forcement Policy). You are required to respond to this letter followirig the instructions in Appendix A.
i The foregoing referenced violations are symptomatic of other underlying problems
. hat need your attention. Your employees should be trained in and encouraged to utilize your existing formal corrective action systems or your allegation processing program to resolve nuclear safety issues such as those that are addressed in this report. We have found that Northeast Utilities has generally used reasonable efforts and a conservative approach to resolve identified safety issues. However the inspectica results indicate a need for more aggressive actions on your part to identify potential employee concerns and create.an atmosphere conducive to the reporting and discussion of those~ concerns. There-fore, in your response to this letter, you should address the actions' that you ,
have taken and plan to take to Improve the two way communications between your employees and management. In the future, w plan to direct allegations of the kind discussed in this report to you for risolution. We will monitor your , allegation processing program to assure ourselves that it.is responsive and. thorough, in accordance with Section 2.790 of the NRCs " Rules of Practice,". Pat ; 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclosure will be placed in the NRC Public Document Room. g3f))&-~ -
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mortheast kuclear' Energy Co. ~h 4 The responses directed by this letter and the enclosed Notice are not subje to the by thePaperwork ciegranceReduction procedures of1980, Act of the Of PLfice of Management and Budget as required 96-511, t Sincerely, ' i g4yd (d . ' 3$ Marvin V. Hodges. Directop D! vision of Reactor Safety
Enclosure:
NRC Inspection Report 50-336/89-13 and Notice of Violation cc w/ encl: V. D. Romberg, Vice President, Nucicar Operations S. E. Scace, Station Superintendent D. O. Nordouist, Director of Quality Servtces R. B.
- 9. M. Miller, Kacich, Manager, Generation Facilities Licensing Station Superintendent, Hadcam Nect Oerald Garfield, Esovire Public Docucint Room (POR) local Public Document Room (LPOR)
Nuclear Safety Information Center (NSIC) NRC Resident inspector State of Connecticut i 3
113 ' Fire fighting procedures and ground fault isolation procedures (see C.3.15
" Panel C26/RPS Ground" for a discussion of ground fault procedures) require '
that certain loads be removed. Errors in the listing of these loads in OP2388 (a copy of which is kept in the control room) will make these procedures more dif ficult and time consuming to execute. However, it must be clearly understood , that these procedures do not involve operation of emergency core cooling or reactor trip systems or their supporting systems to initiate protective action. However if they lose some of these electric systems may be initiated due to their " fail safe" design power. Conclusion it was substantiated that there were some errors in drawings for electrical panels and the drawings did not always reflect at-built conditions. These errors ; are in the process of being corrected. However, it was not substantiated that these errors involved nuclear safety related equipment and that there was a danger to public health and safety, licensee claims blanket exemption to itthe was also not National substantiated Electrical Code. that the Violation of SP-EE-076 is not substantiated since this standard is not applicable to Unit 2. C.6 Reactor Trip Circuit Breakers (related to C.3.41) On April 14, 1989, the employee issued a memorandum to the NRC Resident Inspector, stating that he overheard a conversation between an elec.trician and his foreman. The electrician stated that 7 out of the 8 RTCD failed the high risk testing. The employee later explaineo that when he said the breakers failed he meant that the _ breakers could not be closed remotely. Oiscussion + The employee's concern as stated in the April 14, 1989 memorandum had been ' fully addressed by the Resident inspector in NRC Report No. 50-336/89-09. , e Section 6.0 of that report listed and discussed the maintenance history of each RTB f rom January 1,1987 through April 19, 1989. The fact that the RTB's cannot be closed remotely is net a safety issue. The details are discussed in allegation C.3.41.1. Conclusion Although the employee's statement is true that the RTBs could not be closed remotely during the high risk test, this is not considered a safety issue. C.8 Improper Installation of " Blue Taos" Allegation Two different blue tags were issued by two different jobs to be placed on two different components in the same system. This is a violation of tagging procedures, a violation of tagging policy, and a potential safety problem.
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l 114 01scussion t During the inspection, one of the allegers notified the inspector of a new ' concern not previously identified to the NRC. He stated that he was authorized to place a blue tag to do maintenance on a sewage discharge pump switch when , he discovered a blue tag had already been placed on the pump breaker for work whiCh was being done by Station Services Department maintenance. He considered i l this to be a violation of tagging rules and a potential safety concern to '! electricians. A blue tag is one which allows only the person who signed for the tag to operate the tagged component. This allows the component to operate while tags, performing certain maintenance without constantly installing and removing , The tagout in cuestion concerned a sewage discharge pump which is outside the scope of NRC regulatory authority. However, the tagging system in question is used station wide for all equipment including safety related equipment. For this reason, the safety issue relative to this allegation was reviewed by the inspector. The inspector discussed this allegation with the alleger, a shif t supervisor, and senior management of the Unit 2 Operations Department. The inspector also reviewed ACP-QA-206A, Station Tagging, Revision 14. Electricians for Unit 2 maintenance were assigned to perform maintenance on sewage pumps per AWD U2-89-7621. They were authorized to hang blue tags on the pump switches. During the process, the electricians found blue tags already installed on the sewage pump breakers in a different location which were for AWD U2-89-69P2. This job was being worked by Station Services. However, the electricians stopped the job untti the situation could be resolved. Upon review - of the circuntances, the inspector cetermined the following: (1) ACP-Ot-2.06A, Section 6.1.17, states in part, "A tag of no other color may be attached to a switch or device bearing a blue tag, and only one blue - tag may be attached to one switch or device at one time, thereby permitting clearance to only one person". Hanging blue tags on separate devices in the same system or circuit is not specifically prohibited but it does not make sense to do so as concurrent jobs could interfere with , each other. (2) Shop personnel performing the work must sign on to the tagout verifying its adequacy. This is a backup to operations hanging the tag. Although l the originally installed blue tag was hung at a dif ferent location, it was noticed by the electricians who stopped the job. The backup system appears to have worked. (3) Two dif ferent departments, Station Services and Unit 2 maintenance were authorized to work on the same component. This conflict was not identified by the shop PMS planners nor by operations personnel prior to authorizing the Unit 2 work. This was caused by _ the fact that (1) although there are meetings among Unit 2 departments to coordinate work, there is apparently little coordination with Station Services which does work at all three units; (2) the PMS AWD system is not equipped to 1
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115 readily identify this kind of conflict; and (3) Operations personnel do not review other tagouts before authorizing new tagouts. They depend on t persons hanging the tags to notify them of conflicts. In this case the *
" conflicting" tags were on different components and not picked up by operations personnel.
(4) There does not appear tc be a formal policy not to hang blue tags on different components in the same system or circuit. However, Operations ' personnel stated that this should not be deliberately done (particularly for adjacent components) as it does not make sense. (S) The alleger felt strongly that this was a potential industrial safety issue . since an electrician may receive an electrical shock by mis-operation of another component in the system. The inspector's evaluation was that. the probability of shock from an internal power source was low, since blue tags create situations where you can expect power. The most that could happen is that power has been shut of f by opening of a Circuit breaker further upstream. However, the application of external sources of voltage such as a resistance check would present a problem if work is not Coordinated. The insoector discussed the above issue with Unit 2 Operations Department management. They noted that ACP-QA-2.6A is a station procedure and any changes made to it would require the agreement of all three Units. Unit 2 agreed to evaluate the following issues and bring any proposed resolutions to station management: (1) Determine on different if there are any potential safety problems in hanging blue tags . components in the same system or circuit if they are signed out to different individuals or jobs. (2) Determine if ACp-QA-2.06A needs to be changed to prohibit more than one ; blue tag on dif ferent components in the same system if these blue tags are ~ assigned to dif ferent jobs or different individuals. It appears to be acceptable to have two blue tags on dif ferent components if they are signed out to the same individual for the same job if that individual wants that extra degree of control- , (3) Evaluate whether or not there needs to be a mechanism in place to attempt to resolve conflicting tagouts before the tags are hung. Item (1), (2), and (3) above are collectively considered an unresolved item penoing resolution by the licensee and further NRC review (50-336/89-13-12). , Conclusion This allegation is partially-substantiated in-that blue tags for different jobs were authorized to hang on different components in the same system and that there is no positive mechanism for identifying this conflict prior to hanging the tags. It was unsubstantiated that this was a violation of company rules as ACP-QA-2.06A does not prohibit the alleged actions. While some licensee
l 126 1 l representatives concurred that it did not make serse to hang blue tags on dif it, ferent components in the same system there was no formai policy prohibiting ( it also could not be substantiated there is a significant safety issue, ' As noted above, there is sufficient questions raised by the alleger that this should be resolved by licensee management. e 9 y
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- UNITED STATES a
NUCLEAR R$0VLArcRY coMMIS$10N
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CEOi!N I
*. 1 478 ALtENDALE ROAD e...e KING OF PRUSSIA. PENNSYLVANIA 194o6 0CT 2 01989 ,
Docket Nos: 50-245; 50-336 ' Northeast Nuclear Energy Company , ATIN: Mr. Edward J. Mroczka Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06101-0270 Gentlemen:
Subject:
NRC Region I Combined Inspection Report Nos. 50 245/89-17; and 50-336/89-17 This refers to the routine resident safety inspection conducted by P. Habighorst of this of fice on July 18 - September 5,1989 and July 27 - September 5,1989 at the Millstone Units 1 and 2, respectively. The inspection consisted of observation of activities, interviews with personnel and document reviews. The results of the inspection are described in the NRC Region Inspection Report enclosed in this letter and were discussed with Messrs. H. Haynes, J. Stetz, and J. Keenan of your staff at the conclusion of the inspection. Your cooperation with us is appreciated. Sincerely, . 4
- "III y Edwar C. 'enz nger ' Thief ,
-Projects Bra No. 4 Division of Reactor Projects
Enclosure:
NRC Region I Combined Inspection Report Nos. 50-245;. and 50-336/89-17 cc w/ encl: W. D. Romberg, Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licensing D. O. Nordquist, Director of Quality Services S. E. Scace, Station Superintendent D. B. Miller, Station Superintendent,. Haddam Neck-Gerald Garfield, Esquire-Public Document Room (PDR)- Local Public Document Room (LPDR). Nuclear Safety Information Center (NSIC) NRC Senior Resident Inspector State of Connecticut Qlj. { . - -
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i i U.S. NUCLEAR REGULATORY COMMISSION l REGION I ( l Report Nos.: 50-245/89-17: 50-336/89-17 '! Docket Nos.: 50-245: 50-336 License Nos.: OPR-21: OPR-65 .. Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, CT 06101-0270 Facility Name: Millstone Nuclear Power Station, Units 1&2 Location: Waterford, Connecticut Dates: Unit 1, 7/18/89-9/5/89; Unit 2, 7/27/89-9/5/89 Reporting Insnector: P. J. Habighorst, Resident Inspector Inspectors: W. J. Raymond, Senior Resident Inspector P. J. Habighorst, Resident Inspector A. A. Asars, Resident inspector, Haddam Neck R. J. Paolino, Lead Reactor Engineer, Division of Reactor Safety R. W. Winters, Reactor Engineer, Division of Reactor Safety D. C. Lew, Resident inspector, Oyster Creek ' L. M. Kolonauski, Reactor Engineer, Division of Reactor Projects Approved by: /M Donald R. Haverkamp, Chief w -M /o/f/ MT Reactor Projects Section 4A ( Date Division of Reactor Projects Inspection Summary: Combined Inspection Report Nos. 50-245/89-17 and 50-336/89-17. Areas inspected: Routine NRC resident and specialist inspection at Millstone 1 [106 regular hours, 9 backshift hours, and 5.5 deep backshif t hours), at Hill-l stone 2 (142 regular hours, 21.5 backshif t hours, and 4.5 deep backshift), of plant operations, surveillance, maintenance, previously identified items, com-mittee activities, evaluation of licensee self-assessment, and Plant Incident ! Reports (PIRs). l l gq watoc& 1
. 2 Results:
- 1. General Conclusions on Adequacy, Strength, or Weakness in Licensee '
Programs Steam generator tube plug engineering reviews, corrective actions, and < planning for repairs were thorough and ccmprehensive, reflecting a licensee program strength in the area of engineering and technical ,, support.
- 2. Violations One licensee identified violation was reported involving failure to evaluate conditions adversa to safety with respect to two hydraulic con-trol units. No notice of violation was issued. (See section 5.8)
- 3. Unresolved items Twa nty-one open environmental qualification items were closed: ten at Unit I and eleven at Unit 2. (Section 3.)
At Unit 1, an unresolved item concerning seismic verification and oper-ability of hydraulic control units was opened. (Section 5.8) At Unit 2, one unresolved item was identified regarding implementation of previous commitments to NRC Bulletin 83-03, Check Valve Failures in Raw Water Cooling Systems of Diesel Generators. (Section 5.5) A second unresolved item was identified regarding lack of design specifi-cations, and the implementation and control of design for the emergency diesel generator saturable transformers. (Section 8.0) 1 l
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. 8 The.11censee reviewed all valves which were tested in the reverse direction and evaluated the conservatism of the test. The results of the review were submitted to the NRC in Mty 1988. Addftionally, an \
exemption request for the affected atmospheric control system valves was submitted to the NRC for approval. The licensee concluded that the reverse direction testing of some atmospheric control system ' valves (valves AC-3A, AC-3B, AC-5, AC-6, AC-7, AC-9. AC-II, and AC-12) was less conservative than direct testing because some con-tainment pressure boundaries, including actuator shaf t seal, body-to-bonnet joints and double 0-ring flange seals, were-not tested. Since reverse direction testing does not test these boundaries and the system configuration does not allow direct testing of these valves, the licensee has proposed in its exemption request to re-verse the valves. Reversing direction of the valves and use of test-able double 0-ring seals would adequately test all the sof tware of-the valves. The licensee recognized that although the valves' soft-ware would be tested, the reversal of these valves would result in non-conservatism in the testing of the seat leakage. . The basis for accepting the non-conservative leak; testing of-the seats is-the his-torical data that concludes that the valve seats are equally leak-tight in either direction. The inspector concluded that the basis for the proposed reversal of the valves is weak. The conservative testing of valve seat leakage should not be eliminated. Additionally, reversing the valves would result in containment pressure tending to unseat the valves under postulated accident conditions. This item remains unresolved.pending the final disposition of the exemption request.
- 4.0 Facility Tours The inspector observed plant operations during regular and backshift tours. I of the following areas:
Control Room Vital Switchgear Room Diesel Generator Rooms Turbine Building Intake Structures Enclosure Building ESF Cubicles Control room instruments were observed for correlation between channels, ' proper functioning, and conformance with Technical Specifications. Alarm conditions in effect and alarms received in the control room were dis-cussed with operators. -The inspector periodically reviewed the night order log, tagout log, plant-incident report (PIR). log, key log, and by-pass jumper log. Each of the respective logs was discussed with opera - tions department staff. No inadequacies.were ncted. m-m+- - -~ :=- T \- < - -- V t- e w w-w m-c e- e e e e-rr-----*'e,r----mvw-rweewe-w-'e+ e w -w- =w erv-w e- w-*e*----Tw'r'ves' --w
l l 9 During an inspection tour on July 17, the inspector verifled proper com-pletion of equipment tagging activitie, at Millstone 1 under switching and Tagging order M97-89-1. This review was completed to verify that the y diesel generator and service water systems were properly restored to service following maintenanri. No inadequacies were identifieo. During an inspection tour on August 28, the inspector verified proper com-pletion of equipment tagging activities at Millstone 2 under tagging orders 2-2414-89, 2-2437-89, 2-1899-89, and 2-1700-89. The reviews in-ciuded corrective maintenance activities on the '8' reactor building com-ponent cooling water pump, and vital AC chillers, No inadequacies were noted. During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and to verify correct communication and equipment status. No inadequacies were noted. 5.0 Plant Operational Status Reviews ' 5.1 Review of plant Incident Reports (PIRs) - Units 1 and 2 The plant incident reports (PIRs) listed below were reviewed during the inspection period to (1) determine the significance of the events; (ii) review the licensee's evaluation of the events; (iii) verify that licensee response and corrective actions were proper; anc, (iv) verify that the licensee reported the events in accordance with applicable requirements. The Unit 1 PIRs reviewed were numbers 89-57, 89-60, 89-61, 89-62, and 89-64 PIR 89-62 involved a problem with the gas turbine generator control system. Review of this matter is described in section 10.1 as followup of AW0s 89-8459 and 89-9071. The Unit 2 PIRs reviewed were numbers 89-87, 89-88, 89-89, and 89-90. ' PIR's 89-67 and 89-88 involved problems with the 'A' emergency diesel generator. Review of this matter is described in section 8.0. No inadequacies were identified. 5.2 NRC Bulletin 89-01, Failure of Westinghouse Steam Generator Tube Mechanical Plugs - Unit 2 Numerous plants have experienced primary water stress corrosion crack-ing (PWSCC) and leaks of Westinght,use steam generator tube mechanical plugs. On February 25, 1989, North Anna Unit 1 experienced a mechanical plug failure following a reactor trip during a feedwater isolation transient. The plug failure caused a 75 gallon per minute (gpm) primary to secondary leak ratt. The failure mechanism involved a full circumferential severance of the top portion of the plug from the body of the plug. The top portion of the plug was propelled up the length of the tube by primary system pressure to a point just above the u-bend transient poin't where it impacted and punctured the outer curvature of the tube. The top portion of the plug subsequently impacted and dented an adjacent tube.
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[, 8, UNITED STATES [ TW k NUCLEAR REOULATORY COMMISSON I
'e f CEQl2N I
[ 475 ALLENDALE ROAD KING OF PRUSSIA. PENNSYLVANIA 164o6 JAN I 91990 Docket Nos. 50-245; 50-336 Northeast Nuclear Energy Company ATTN: Mr. Edward J. Mroczka Senior Vice President - Nuclear Engineering and Operations , P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region I Combined Inspection Nos. 50-245/89-25; 50-336/89-23 This refers to routine resident safety inspection conducted by Mr. P. Habighorst and other; of this of fice during October 21 - December 4,1989, at the Millstone Nuclear power Station, Units 1 & 2, Waterford, Connecticut of activities authorized by NRC License Nos. DPR-21 and OPR-65 and to discussions of our findings by Mr. Habighorst and Mr. Dempsey with Messrs. J. Keenan and J. Stetz of your staff at the conclusion of the inspection. Areas examined during this inspection are described i' the NRC Region I inspection report which is enclosed with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, measurements, and document reviews. The enclosed inspection report (Sections 3.6, 5.1.2, 5.1.3, and 6.2) discusses two events involving check valve failures in safety-related applications on Millstone 1 and 2. The events highlight the need to improve your program to assure continued check valve operability and acceptable performance. We are aware of your initiatives to upgrade procedures related to check valves.
- Please let us know within 60 days f rom receipt of this letter of your plans and schedules regarding improvements of check valve performance.
On November 8 and December 4, 1989, the resident inspector presented various concerns (See Appendix A of the enclosure) from employees at Millstone station to your staff for resolution. Further review of your actions to address employees concerns will be included in subsequent routine inspections. In accordance with 10CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and the enclosure will be placed in the NRL Public Document Room. x l mg3oo3@^
%x.
, Ne theast Nuclear: Energy Company -2 ' JAN' l 9 1990
, .The-response requested by this letter is not subject to the clearance procedures of- the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511, c
-Your cooperation with us in this matter is appreciated.
Si rely, ,
/
/ .
/1L14 / -'
E ward C. Venzinger, Projects Branch No. Olvision of Reactor o cts
Enclosure:
NRC Region I Combined Inspection Report Nos. 50-245/89-25; 50-336/89-23 cc w/ encl: W. D. Romberg, Vice President, Nuclear Operations R. M. Kati ch, Manager, Generation Facilities Licensing D. O. Nordquist, Director of Quality Services S. E. Scace, Station Superintendent J. P. Stetz, Unit 1 Superintendent J. S. Keenan, Unit 2 Superintendent Gerald Garfield.. Esquire Public Document Room (PDR) local Public Document Room (LPDR) Nuclear Saf ety Information Center (NSIC) .; NRC Senior Resident Inspector ' State of Connecticut k 4 I 4 1- -- --
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U.S. NUCLEAR REGULATORY COMMfS$30N REGION I Report Nos.: 50-245/89-25 and 50-336/89-23 4 Docket Nos.: 50-245; 50-336 License Nos. OPR-21; OPR-63 \. Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270
/
Facility Name: Millstone Nuclear Power Station, Units 1 and 2 Inspection at: Waterford, Connecticut Dates: October 21 - December 4,1989 Reporting Inspector: P. J. Habighorst, Resident Inspector Inspectors: W.J. Raymond, Senior Resident inspector, Millstone Station P.J. Habighorst, Resident inspector, Millstone Unit 2
- 0. A. Dempey, Resident Inspector, Millstone Unit 1 G. Vissing, f roject Manager, Project Directorate 1-4, Office of Nuclear Reactor Regulation Approved by: o / dva ([/P/40 Donald R. Haverkamp, Chief (/ Date Reactor Projects Section 4A Division of Reactor Projects Inspection Summary: Combined Inspection Reports Nos. 50-245/89-25; 50-336/89-23 Routine NRC resident inspection of plant operations, Areas Inspected:
surveillance, maintenance, previously identified items, engineering / technical support, committ<e activities, evaluation of licensee self-assessment, security, allegations, and radiological controls. Results:
- 1. General Conclusions on Adequacy, Strength or Weakness in Licensee Programs The overall operations control during the observed plant evolutions were effectively implemented and is considered a strength. (Section 3.3.2)
A strength was noted in problem identification of the seismic _ supports for the service water strainers, classification of the associated Unusual Event, and control of plant activities during che Unusual Event-at Unit 2. (Section 3.5.1) Two events involving check valve failures in safety relateu applications on Millstone 1 and 2 highlight the need to improve the program to assure continued check valve operability and acceptable performance. (Sections 3.6, 5.1.2, 5.1.3, and 6.2)
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22 health physics actions to reduce the noble gas were application ; of a nitrogen purge en 1.he pressurizer, containment venting, and ' random airborne samples, t 'l The inspector reviewed the sensitivity of the PCM-18 personnel contamination monitors. The detectability of the monitors are ' 5000 dpm/100cm2 for a two-inch diameter location; and low levels , of distributed contamination (i.e. <1000 dpm/100cm2) over the entire body. During the early part-of the outage, licensee ; management decided to let contractor force personnel leave the ' site after alarming the PCM-18, based on: personnel contamination surveys with an HP-210 detector revealing less than 1000dpm/100cm2; showers of select individuals to determine if contamination is inhaled and deposited internally; and assurance that the workers revisit the PCM-1B monitors the following day, and the subsequent results were less than the alarm setpoint. The licensee provided a plant distribution memo (MP-S-5244) on November 3, to identify the health physics actions, and address health concerns regarding personnel contamination monitor alarms. The inspector called the concerned individual, base ' on the investi-gation results. The individual had no further que.' ions and was satisfied. This item is closed. 4.4 Su- ary Licensee review of the root cause determination for LER 89-20 was sufficient with extensive followup. Sufficient centrol of personnel monitoring during the Unit 2 outage was noted. 5.0 Maintenance / Surveillance (IP 62703/61726/92702) , 5.1 Observation cf Maintenance Activities - Units 1 and 2 Maintenance activities were reviewed to determine the scope and nature of work done on safety related equipment and to verify the activities were completed in accordance with established l procedures and plant equipment controls. Proper implementation of safety related tagging was also verified during inspector review of maintenance activities. Tagging associated with switching orders 1-2098-89, 1-2101-89 2-2886-89, 2-2877-89, 2-2880-89, and 2-2951-89, and by pass jumper evaluations 89-67, 89-68, 89-69, 89-70, 89-74 and 89-75 were reviewed and found l acceptable. Maintenance and testing activities associated with the following were reviewed to verify (where applicable) procedure compliance and proper return to service, including operability testing.
23' ' AWO M1-89-11733, "A" FRV Failed Open AWO M1-89-11757, "B" FRV Packing Leak AWO M1-89-11744, "A" Feedwater Pump Discharge Check Valve y AWO M1-89-12732,
-- "B" FRV Minimum Flow Valve Air Supply AWO M1-89-12722,
-- "A" FRV Minimum Flow Valve Air Supply AWO M1-89-12723, "C" FRV Minimum Flow Valve Air Supply AVO M1-88-10381, Perform IC-414A, CRD Module Instrument Calibration ,
AWO M1-88-06342, PDCR 1-74-86: Extend Existing LLRT Connections on Electrical Penetrations AWO M1-89-12936, "A" Recirculating Pump MG Set Speef Controller Not functioning Properly AWO M1-89-12948, Replace K1038 and K1048 Relays "B" Recirculating Pump MG set Speed Control AWO M1-89-10179, Repair Breaker fer 1-MS-6 at MCC-101-AB-2 AWO M1-89-11246, "B" Fuel Pool Meat Exchanger Tube Leek AWO M1-89-11983, IRM 16 Voltage Pre-Amplifier AWO M1-89-12444, "A" FRV Failed Open AWD M1-89-12607 Emergency Diesel Generator - Inspect Right Side Fuel Control Rod per SIL-A-22 pDCR l-89-?2, Assembly Feedwater Discharge Check Valve Seat Retoining Modification
- PDCR l-89-33, FFV Minimum Flow seismic Backup Air Supply AWD M2-89-11543, Pressurizer Heater Inspection AWO M2-89-06672, Auxiliary Feedwater Check Valve Repair AWO M2-88-11232, Vital Inverter Capacitor Replacement AWD M2-89-12229, Service Water Strainer Support Installation AWO M2-89-13005. Valve 2-FW-12A Electrical Equipment Qualification Walkdown AWD M2-89-11824, Repair of 2-FW-48 AWD M2-89-11825, Repair of 2-FV-4A No inadequacies were identified.
for additional inspector followup. The items below were selected 5.1.1 AWD M1-89-11744, "A" Feedwater pump Discharce Check Valve Failure on October 19, 1989 The "A" feedwater regulating valve (FRV) was removed following draining'and cooldown of the header. A 7/8 inch bolt was found under the "A" FRV seat. -The licensee determined that the bolt came from the "A" feedwater pump discharge check valve FW-2A, Licensee inspection per AWD 89-11733 identified no major damage to the "A" FRV; there was minor scoring on the cage and on the hemispherical head of the FRV. The damage did not affect either. seating surface on_the dual seat Copes Vulcan' valve. Licensee corrective action was to lap the FRV seats and re-assemble the valve. The valve was subsequently stroke tested satisfactorily. t I
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/[ ~ AEGION I 474 ALLENDALE ROAD KING OF PMUSSIA. PENNSYLVANIA 1 Hot MAP 0 e 1930 4 Docket No; 50-336 Northeast Nuclear Energy Company '
ATTH: Mr ' Edward J. Hroczka Senior Vice_ President - Nuclear Engineering and Operations ' P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region I Inspection No. 50-336/89-24 ___ i This refers to the routine resident safety inspection conducted by Mr. p. Habighorst of this of fice on December 5,1989 - January 19, 1990, at' M1ilstone Nuclear Power Station, Unit' 2, Waterford, Connecticut, of activities authorized by NRC License No. DPR-65 and to the discussions of our findings-held by Mr. Habighorst with Messrs. S. Scace and J. Keenan of your staf f at the. ' conclusion of the inspection. Areas examined during this inspection are. described in the NRC Region-I inspection report which is enclosed with this letter. Within these areas. the inspection consisted of observation of activities, interviews with-personnel,- measurements and document reviews, Based on the results of this inspection,- it appears -that_ one of your activities was not conducted in full compliance with NRC requirements, -as set _ forth in the. Hotice of Violation enclosed hertwith as Appendix. A. The licensee-identified violation involved operation of the facility in cold shutdown for a period of ;- 29 hours without an operable containment gaseous and par _ticulate. radiation - l monitor. You reported this event in Licensee Event Report (LER) 89-009. ' This violation has:been categorized by severity level =ia-accordance with the
" General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Par. 2, Appendix C (Enforcement Policy). You are required to-respondito. this i
letter and in preparing your response, you should follow the instructions in Appendix A. , In review of the apparent violation, NRC inspection identified two deficiencies r not recognized or reported in LER 89-009;. 2pecifically, the failure to identify an additional technical specification limiting condition for operation appli-cable- to the- event; and secondly, an inaccurate time interval for the out-of-service radiation monitors. Our independent review determined that aj plant vulnerability existed based on plant decay heet rate and reduced i inventory operation at the time of the event. Therefore, in your. response to this letter, you are requested-to address LER accuracy and your assessment of plant vulnerability' during the event,
. _ - . - _- _ _ - =
Northeast Nuclear Energy Company 2 With regards to matters discussed in section 8.3 of the enclosed inspection t report regarding LER 89-011, we are concerned about the apparent untimely ' notification and reporting of the incorrect. location of the air supply check valve for a service water isolation valva. The deficiency was identified on , September 6,1989, but the event was not reported until January 5,1990. You are requested to respond to this concerr.. Please submit to this office, within thirty days of the date of this letter, a written explanation in reply, ' including; (1) your assessment of when the service water system deficiency should have been reported and (2) the steps which have or will be taken to assure timely notification and reporting of events or conditions as required by 10CFR 50.72 and 10CFR 50.73. This response may be appended to your response to a similar report request for an event at Millstone Unit 3 as documented in inspection 50-423/89-23. The response directed by this letter and the accompanying notice is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL-96-511. Your cooperation - .th us in this ratter is appreciated, Sincerely, (( Qf y Q Edward C. Wenzinger, Chief i Projects Branch No. 4 Division of Reactor Projects ('[
Enclosures:
I
- 1. Appendix A, Notice of Violation '
- 2. NRC Region 1 Inspection Report No. 50-336/89-24
! cc w/encls: W. D. Romberg, Vice President, Nuclear Operations S. E. Scace, Nuclear Station Director, Millstone Station J. S. Keenan, Nuclear Unit Director, Millstone Unit 2 D. O. Nordquist, Director of Quality Services R. M. Kacich, Manager, Generation Facilities Licensing Gerald Garfield, Esquire Public Document Room (PDR) Local Public Document Room (LPDR) Nuclear safety Information Cer.ter (NSIC) NRC Senior Resident Inspector State of Connecticut l _
\ - 1-U.S. NUCLEAR REGULATORY COMMISSION t REGION I Report No. 50-336/89-24 Docket No.: 50-336 License No. DPR-65 Licensee: Northeast Nuclear Eneroy Company P.O. Bor 270 Hartford, Connecticut 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2
- nspecticn at: Wate-ford, Connecticut I r s p e c t i e r.
Concucted: Decerber 5, 1939 - January 19, 1990 Rep 0-tir; Inspector: Peter J. Habighorst, Resident Inspector Inspectors: William J. Raymond, Senior Resident Inspector, Millstone Peter J. Habighorst, Resident inspector, Hillstone 2 William Oliveira, Reactor Engineer, Operational Programs Section, Division of Reactor Safety Approved by: a4[/ . M 3/MYO Donald R. Haverkamp, Chief Date Reactor Projects Section 4A f Division of Reactor Projects Inspection Sun. mary: Inspection on December 5, 1989 - January 19, 1990 (Inspection Report No. 50-336/89-24) Areas Ir.spected: Routine NRC resident inspection of plant operations, radio-logical controls, maintenance / surveillance, engineering / technical support, security, and safety assessment / quality verification including committee activities and Licensee Event Reports. Within certain of these areas, the inspection included review of licensee actions in response to allegations and previously identified items.
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- __ - - - - - - - - ~
Results:
- 1. General Conclusions en Adeovacy, Strength or Weakness in -
Licensee Programs inadequate timeliness of a reportability evaluation and an example of a itcensee event report (LER) inaccuracy were noted in review of LER 89-011-00 and LER 89-009, respectively. (Section 8.3) < Extensive licensee efforts were not successful in maintaining sufficient reliability of the steam jet air ejector radiation monitor. A significant amount of corrective maintenance was noted without achieving continual reliabiitty of the monitor. (Section 4.3.1) The establish. ment of a pilot precictive maintenance program for rotational equipment was noteworthy. (Section 5.4)
- 2. Violations Within the scope of this inspection, a cited licensee-identified violation was noted, which' concerned operation of the facility in cold shutdown for a period of about 29 hours without an operable containment radiation gaseous and particulate monitor.
(Section B.3) One non-cited licensee-identified violation was identified for the event documented in LER 89-011-00. (Section 8.3, page 25)
- 3. Unresolved Items Five previously unresolved items were closed, and two items were updated and remain open (Sections 5. 3.1. , 6.1.1. , 6.1. 2. , 8. 2.1. , 8. 2. 2. , 8. 2. 3. ,
and 8.2.4.). ' Four unresolved items were opened regarding: (1) pre planned alternate monitoring method for high range noble gaseous monitor (Section 4.3.2.); (2) modification configuration control process as it relates to incorpor-ating vendor information (Section 8.2.4); (3) the consequence of failure to test control room annunciators (re: LER 89-008-00) (Section 8.3, page 20); (4) revision of LER 89-009-00 (Section 8.3, page 23); and, (5) timeliness of reportability evaluations (re: LER 89-011-00) (Section 8.3, page 25).
- 4. A11ecation Followup A worker concern regarding the capability of the steam jet air ejector radiation monitor to detect a tube leat was reviewed. (Section 4.3.1.)
11 l
2 NRC Activities s Routine resident inspection involving 131 regular hours, 9 backshift hours, and 2 deep backshift hours. 3.0 Plant Operations 3.1 Control Room Observations Control room instruments were observed for correlation between channels, proper functioning, and conformance with technical-spect-fications. Alarm conditions in effect and alarms received in the control room were discussed with operators. The inspector periodi-cally reviewed the night order log, tagout log, plant incident report (pIR) log, key log, and bypass jumper log. The following tagouts were reviewed for acceptability and implementation: 2-3'O "F-20 Heat Exchanger," 2-2-90 "C Instrument Air compressor," 2-4241-89
" Pressurizer Vent Valves 2-RC-424, 2-RC-425," 2-3022-89 " Spare Ba ttery Cha rger," and 2-4193-89 " Charging system valve '2-CH-507." No inade:.;2:ies we e m e:. Each of the respective logs was discussed with operation department staf f. No inadequacies were noted.
3.2 Plant Tours The inspector observed plant operations during regular and backshift tours of the following areas: Control Roon Containment Vital Switchgear Room Diesel Generator Room Turbine Building Intake Structure i Enclosure Building ESF Cubicles
)
During plant tours, logs and records were reviewed to ensure I compliance with station procedures, to determine if entries were correctly made, and to verify correct communication and equipment status. No inadequacies were noted. 3.3 Review of plant Incident Reports l The plant incident reports (PIRs) listed below were reviewed during the inspection period to (1) determine the significance of the events; (11) review the licensee's evaluation of the events; (iii) verify the licensee's response and corrective actions were proper; ar.d, (iv) verify that the licensee reported the events in accordance with applicable requiremeats, if required. The PIRs reviewed were: 89-132, 'B' service water header leak 89-136, reactor protection system channel 'O' power supply failure
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*\-***** - 478 ALLENDALE ROAD' KINQ OF PRUSSIA, PitNNSYLVANIA 194o0 JU!,' 2 81990 Docket No. 50-3 Northeast Nuclear Energy Company ATTN: Mr. Edward J. Mroczka Senior Vice President - Nuclear Engineering and Operations ' '
P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region i Inspection No. 50-336/90-09 This Mr. refers to the routine resident safety inspection conducted by P. Habighorst of this office on April 17 - May 29, 1990 at Millstone Nuclear Power Station, Unit 2, Waterford, Connecticut of-activities authorized Mr.NRC by Licensewith Habighorst No. Mr. OPR-65 and to the discussions of our findings by J. Smith of your staff st the conclusion of the inspection. Areas examined during this inspection are described.in the NRC Region I inspection report which is enclosed with this letter. Within these areas, the inspection consisted measurements, and document of observation reviews. of activities, interviews with personnel, Bosed on the results of this inspection, it appears that one of your activities was not conducted in full compliance with NRC requirements, as set forth in the Notice of Violation enclosed herein as Appendix A. The violation involved tne s improper performance of surveillance required by technical: specifications ~on the main station batteries. This violation has-been categorized-by severity - level in accordance with the " General Statement of Policy and Procedure for NRC ~ Enforcement Actions," 10 CFR Part 2, Appendix 'C (Enforcement Policy,1990 . You are required to respond to this letter and in preparing-your response), you-should follow the instructions in Appendix A. .While the safety significance of _ the violation is low, we are concerned that proceduras were not followed on numerous occasions and that a supervisor accepted out of specification results without an adequate documented evaluation. Please give particular attention to these matters in your response. . Your cooperation with us is appreciated. n rely, l . Edward C. WeAzinger, Chie Projects Branch No. 4 ~ Division of Reactor Projects
Enclosures:
- 1. Aopendix A, Notice of Violation
- 2. NRC Region I Inspection Report No. 50-336/90-09
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l U.S. NUCLEAR REGULATORY COMMISSION REGION I L Report No.: 50-336/90-09 Docket No.: 50-336 License No. OPR-65 Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 ( Inspection at: Waterford, Connecticut Dates: April 17 - May 29,1990 Reporting Inspector: Peter J. Habighorst, Resident Inspector Inspectors: William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident inspector Thomas Moslak, Resident Inspector, Three Mile Island Douglas Dempsey , Resident Inspector, Millstone 1 i
, ,/ '
[j Approved by: b 441 M onald R. Haver h/J T / 90 ief ITa te ' + ReactorProjectg4m) St on 4A Division'of Reactor Projects - Insoection Summary: 1_nspection on April 17 1990 - May_29, 1990 Insoection Recort No 3 0-336f90-09 Areas Inspected: Routine NRC resident inspection of plant operations, surveillance, maintenance, previously identified items, engir.eering/ technical support, committee activities, periodic reports, licensee event reports, and allegations. Results: ' See Executive Summary Aco7I W s t __- - - __ _ - - - - - - - - - - - ~
I 1 Executive Summary I t plant Operations . Review of licensee actions during the reactor trip concluded the licensee's response was adequate. NRC inspection of outage activities identified an experienced and ef ficient operations support organization. .- The inspectors will review licensee's corrective actions to prevent recurrence of emergency diesel generator exhaust pipe lagging fires. Radiological Controls Good health physics assistance and control were noted; specifically, in steam generator plenum work. Surveillance and Maintenance Review of routine maintenance identified no noteworthy findings. Surveillance activities were generally performed in accordance with established test con-trols. A violation was identified regarding the improper performance of sur-veillance required by the technical specifications on the main station bat-teries. Specifically, On March 7, 1990, uncertified contractor personnel per-formed a main station battery surveillance while not under direct observation of certified test personnel; levels of battery cells not meeting the procedure acceptance criterion were not documented; and water additions required by the procedure were not performed. On March 14, 1990, the amount of water added to individual battery cells was not documented. Encineerino and Technical Support Stean generator eddy current testing was well implemented with an adequate inspection scope. The licensee's decision to inspect and locate the source of primary to secondary leakage displayed a good regard for plant operational safety. Security Routine review in this area identified no noteworthy findings. Safety Assessment / Quality Verification Routine review in this area identified no noteworthy findings. I
\
7 DETAILS 1.0 _S_urmary of Facility Activities Millstone Nuclear Power Station Unit 2 (Millstone 2 or the plant) operated at rated 4:00 p,mthermal on Aprilpower
- 26. from the beginning of the inspection period until ,
The licensee commenced a planned downpower to 10 percent rated to remove a workman's rag located in the main generator cool-ing air f an for the alterex (alternator-exciter collector ring). The rag was sucked into the f an housing on April 24 during preventive maintenance activity on the collector ring. The turbine was off-line at 8:00 p.m. and reactor power was stabilized at 7-10 percent. The plant returned to 100 percent rated at 12:32 p.m. on April 28 af ter removal of the obstruction in the alterex. On May 8 at 12: 49 a.m. the operators manually tripped the reactor based on decreasing water level in the No. I steam generator (see report detail 2.3). On May 10, the licensee decided to place the plant in cold shutdown based on elevated primary-to-secondary leakage rate calculations. The plant was in cold shutdown at 5:01 a.m. on May 11. The outage was used to conduct eddy current testing on both steam generators to identify and cor-rett the leakage. At the end of the inspection period, the plant was being maintainec in cold shutdown. NRC Activities The inspection activities during this report period include 162 hours of inspection during normal activity working hours. In addition, the review of plant operations was routinely conducted during periods of backshif ts - (evening shif ts) and deep backshif ts (weekend and midnight shif ts), in- ; spection coverage was provided for 21 hours during backshifts and 7 hours during deep backshif ts. On May 10,17, and 18 the resident inspector, senior resident inspector, and regional Division of Reactor projects Branch Chief met with the electad officials of the municipalities of Waterford, Montville, New London, and East Lyme. The meeting agenda included the NRC inspection activities at Millstone, and planning for an upcoming public meeting-(scheduled for July 25 in East Lyme). 2.0 plant Operations 2.1 Control Room Observations Control room instruments were observed for correlation between channels,- proper functioning, and conformance with Technical Specifications. Alarm conditions in effect and alarms received in the control room were discussed with operators. The inspector
-periodically reviewed the night order log, tagout log, plant incident report (pIR) log, key log, and bypass jumper log. Each of the respective logs was discussed with operations department staff. The
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(: , 2 l l inspector reviewed and verified tag outs: 2-715-90, 2-870-90, 2-749-90, 2-631-90, 2-631-90, and 2-896-90 and bypass jumpers ' 2-90-27, " Inverter 2 Relay RC-1", and 2-90-29, " Shutdown Cooling Suction Valve Interlock". No inadequacies were noted. 2.2 plant Tour _s The inspector observed plant operations during regular and backshift tours of the following areas: Control Room Containment Vital Switchgear Room Turbine Building Diesel Generator Room Enclosure Building Intake Structure ESF Cubicles During plant tours, logs and records were reviewed to ensure ccmpliance with station procedures, to determine if entries were correctly made, anc to verify correct com:nunication and equipment status. No inadequacies were noted. 2.3 Manual Reactor Trip On May 8 at 12: 49 a.m. , the control room operators manually tripped the reactor f rom 100 percent of rated power based on a decreasing level indication in the No. I steam generator. All plant safety systems functioned generally as designed during the transient, ar.i the trip was uncomplicated. At 4:04 a.m. the control room operators closed the main steam isolation valves (MS1Vs) in preparation for isolating the main condenser for planned mainienance work activity on the low pressure feedsater heate s. During closure of the MSIVs, both main steam atmospheric dumps failed to it.11y open resulting in ' main steam safety valves opening for a short duration. The licensee-reopened the MSIVs at 4:PO a.m. and began troubleshooting activities on the atmospheric dump valve control circuitry (see report detail 4.3). Licensee post-trip revier concluded the cause of decreasing level in the latingNo.valve. I steam generator was a f ailure of the No.1 feedvater regu-The licensee maintenance department disassembled the valve anc identified (see report that the valve stem had separeted from the plug detail 4.1.1). Feedwater system presisre upon separation of tho valve stem and plug forced the valve to close, and thus.re-suited in decreasing level in the No.1 steam generator. The following lists the chronology of critical plant parameters during the manual reactor tr.ip; i
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-Docket No. 50-336 Northeast Nuclear Energy Company 3,,
~ATTH: Mr.-Edward J. Mroczka Senior Vice: President - Nuclear Engineering and Operations P.O. Box 270-Hartford, Connecticut 06141-0270 Gentlemen: ;
Subject:
NRC Region I Inspection No. 50-336/90-14 This' refers te. the routine resident safety inspection conducted by P. Habighorst,-J. Stewart, and W. Raymond of this office on July 12 - August:21, 1990,' at Millstone Nuclear Power Station, Unit 2, Waterford,. - Connecticut, of activities authorized by NRC License'No. OPR 65 and-to the-discussions;of- our findings by Mr. Habighorst and Mr. Stewart with-- ,- Mr. J.-.Keenan of your staf f at the conclusion of the inspection.: Areas examined during this inspection are described in thel NRC Region ! in-1 spection report which is enclosed with this letter. ' Within these areas, the; inspection consisted.of observation of activities, interviews with' personnel, ' and document reviews. Overall operation of the_ facility' continued to be satis-- factory. Your cooperation with us is appreciated.. l E sincerely,= 0.91CiL15;S"?;0 T/ - m GY!ARD C.1EtEGER Edward C.;Wenzinger, Chief' Projects. Branch No.: 4 Division of Reactor Projects L
Enclosure:
- NRC ' Region I Inspection. Report No.; 50-336/90-14 i
9&oow/OV~~ . OFFICIAL RECORD COPY- IR'MII12 90 214- , 9/14/90-n V
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t . 3 U.S. NUCLEAR REGULATORY COMMISSION (^ REGION I Report No.: 50-336/90-14 . , Docket No.: 50-336 License No. DPR-65 Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 Inspection at: Wate* ford, Connecticut Dates: July 12_- August 21, 1990
-Reporting Inspector: J. Scott Stewart, Project Engineer, Division of Reactor Projects Inspectors: William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector J. Scott Stewart, Project Engineer, Division of Reactor Projects R. Paolino, Senior Engineer, Division of Reactor Safety Approvec by: 8 d/ eeu (9//P/ft)
Ecnalc R. Haverkamp, Chie(/ Date ?
-Reactor Projects Section 4A Division of Reactor Projects Insoection Summary: Insoection on July 12 - August 21, 1990 i
Insoection Report No. 50-336/90-14 Areas Inseetted: Routine NRC resident and specialist inspection of plant operations, maintenance / surveillance, safety assessment / quality verification, and employee concerns. Results: See Executive Summary I' i 9ofocewl0Y,,- L [, [ i t L f
g . . 1 DETAILS ( 1.0 Summary of Facility Activities The plant operated at Nil power throughout the inspection per N . During the pericd, plant-personnel were preparing for the Unit 2 outwe, which will commence on Septimoer 15, 1990. One of the preparatory tasks com-pleted was new fuel receipt and inspection. Major tasks scheduled for the outage include: replacement of the main service water header piping, moisture separator tube bundle replacement, and installation of fc&ctor coolant system mid-loop level instrumentation. NRC Activities The inspection activities during this report period included 100 hours of
- inspection during both normal and backshift working hours. Inspection activities includea plant r . rations, maintenance, security, and sur-veillance.
On July 25, the Millstone resident staff as well as a numoer of both regional and heacouarters management and staff personnel conducted a public meeting to discuss NRC activities at the Millstone site. Approximately 15 memoers of the public as well as a number of local officials and Millstone staff attenced the meeting. 2,0 Plant Ooerations 2.1 Control Room Observations Control room instruments were observed for correlation between. cnannels, proper functioning. and conformance with Technical ' Specifications. Alarm concitions in effect and alarms _ received in the control room were discussed with operato-s. The inspector periodically reviewed the night order log, tagout log,: plant incident report (PIR) log, key log, and bypass jumper log. Each of the respective logs was discussed with operm tions department staff. On July 25, the inspector determined that jumper bypass tag 2-90 h on containment radiation monitor RM-8262 identified the condition of
" leads lif ted" when the leads to which the tag was attached appeared to be connected (RI-A-90-0118). .The unit 2 shift supervisor was in-formed and prompt action was taken by the crew to verify that the tag was incorrect and that there were no immediate safety consequences.
The shif t supervisor promptly informed the Instrument-and Controls department and action was taken to correct the deficiency. TheLtag in question was cleared and replaced with two jumper bypass tags which accurately reflected that-the flow control valve in.the system was disengaged from the valve controller and the control.ler was ' deenergized. The flow control valve had been in manual control for some time and the plant has instituted periodic checks to verify the
2
\
system flow. . initially, the flow conicol valve controller leads were . lif ted to disengage the flow control . valve and tag 2-90-17 accurately \ reflected the condition. At a later date, the controller was de-energi.ted and the controller leads were reattached, however, the _ tag was not updated to reflect this condition. Neither system operability ' nor personnel safety were affected by the discrepancy and this issue is considered closed. ,. No other discrepancies were noted. 2.2 Plant Tours The insoector observed plant operations during regular and backshift tours of the following areas: Control Room Containment Vital Switcngear Room Diesel Generator Room Turoine Builcing Intake Structure Enclosure Guilding ESF Cubicles Durirg plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly mace, ano to verify correct communication and equipment status. No significant observations were noted. During tne irspection, a cuestion was raised as to the ability of a penetration thru-wall fire barrier number A-4/T-1 to meet its requirement to prevent spread of a fire. An NRC regional specialist visiting the site on an unrelated inspection accompanied a utility engineer on a walkoown of the penetration._ It was concluded that the penetration met the requirements for the given seal design. The : penetration is. filled with grout from the cable vault side and passes - tnrough a 12-inch thick concrete block wall using 2 inches of damming material. The seal design requires 8 inches of grout and I inch of camming material, hence, the minimum requirements were exceeded. 2.3 Stanc-by Reaciness of Engineered Safety Features System and l System Walkcown
- l. During the inspection period, two engineered safety feature (ESF) j systems were reviewed to verify system operability. The systems l- reviewed uere auxiliary feedwater and control room ventilation. The review included proper positioning of major flowpath valves, proper operation of- indication and controls, and visual-inspection for proper lubrication, cooling, and other conditions. References used were:
Final Safety Analysis Report ' Plant instrument and piping diagrams (P& ids) 25203-26005, and 25203-26027 l l l
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NOV 0 5 W c Docket No. 50-336
-t Northeast Nuclear Energy Company ATTN: Mr. Edward J. Hroczka ,
Senior Vice President - Nuclear Engineering and Operations P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region I Inspection No. 50-336/90-18 This letter transmits the results of the routine safety inspection conducted by Mr. P. Habighorst of this office on August 22 - October 1,1990, at Millstone Nuclear Power Station, Unit 2, Waterford, Connecticut, of activities authorized by NRC License No. DPR-65. Our findings were discussed by Mr. Habighorst with Mr. J. Keenan and others of your staff at the conclusion of the inspection. Areas examined.during the inspection are described in the NRC Region I inspection report which is enclosed with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, and document reviews. Based on the results of this inspection, it appears that certain of your activities were not conducted in full compliance with NRC requirements, as set forth in the Notice of Violation and in the Notice of Deviation enclosed here-with as Appendices A and B. The violation involves the unavailability of the : reactor vessel level monitoring system during reduced inventory operation, and the deviation involves a failure to implement a commitment for the spent fuel 3 pool boraflex coupon surveillance program. You are required to respond to this @ letter, and in preparing your response, you should follow the instructions in .5 Appendices A and B. [C With regard to operational events that occurred during and shortly after this inspection, we are concerned by the number of problems noted with personnel performance, as identified in tagging errors, procedural non-adherence, and inadequate configuration controls. The examples are highlighted in report i details 2.3.1, 2.3.2, 2.5, and 7.1. Another example of poor personnel per-l formance that occurred after the inspection period involved the dropping of the incore instrumentation plate. We request that you respond to this matter within 30 days of your receipt of this letter. In your response, you should l address these personnel performance problems and describe the actions taken l individually and collectively to prevent recurrence.
//hb/W" 0FFICIAL RECORD COPY IR MILL 2 90 0001.0.0 11/29/80 C, h
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O O t Northeast Nuclear Energy Company 2 g3g I ( The responses requested by this letter are not subjet.t to the clearance 'I procedures of the Office of Management and Budget, as required by the Paperwork. Reduction Act of 1989, PL 96-511. , Your cooperation with us is appreciated. Sincerely, Original Signed By: Edward C. Wenzinger, Chief Projects Branch No. 4 Division of Reactor Projects
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. Appendix B, Notice of Deviation
- 3. NRC Region I Inspection Report No. 50-336/90-18 cc w/encis:
W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Services R. M. Kacich, Manager, Generation Facilities Licensing S. E. Scace, Station Director, Millstone J. S. Keenan, Nuclear Unit Director, Millstone Unit 2 Gerald Garfield, Esquire Public Document Room (PDR) Local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) ! NRC Senior Resident Inspector State of Connecticut OFFICIAL RECORD COPY IR MILL 2 90 0002.0.0 11/29/80
v w- , t U.S. NUCLEAR REGULATORY COMMISSION k REGION I ' Report No.: 50-336/90-18 Docket No.: 50-336 License No, OpR-65 Licensee: Northeast Nuclear Eneroy Company p.0. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 Inspection at: Waterford, Connecticut Dates: August 22 through October I, 1990 Reporting Inspector: Peter J. Habighorst, Resident Inspector Inspectors: William J. Raymond, Senior Resident Inspector Peter J. Habighorst, Resident Inspector Gu(S.Vissing,NuclearReactorRegulation,PDI-4 Approved by: bd/ s ///[b Donald R. HaverKamp, Chi p Date Reactor Projects Section 4A Division of Reactor Projects Inspection Summary: Inspection on Aucust 22, 1990 - October 1, 1990 Inspection Report No. 50-336/90-18
- Areas inspected
Routine NRC resident and specialist inspection of plant-l operations, surveillance, maintenance, previously-identified items.. engineering and / technical support, committee activities, licensee event reports, special reports, Results: See Executive Summary l 1
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. . , 17-In conclusion, based on the fact that the loop seals have been removed from the PSV piping and testing is completed at conditions-resembling actual use, anomalies as described in Information Notice \
89-90 are not present for Millstone 2. No inadequacies were noted in the inspection. t 4.5 Steam Generator Level Disclepanc,y_ On August 11, 1990, the licensee entered technical specification (TS) requirement 3.3.1.1 for channel 'O' stean generator level on the No. I steam generator. Entrance into the TS action statement, which was based on the channel 'O' level being slightly outside the acceptance criterion during the performance of daily ~ channel checks as detailed in procedure OP 2619-1, required that the channel be placed in bypass. The acceptance criterien requires that all channels agree within 4% level indication. _On August 20_, the licensee completed the repair of the channel and cnannel 'O' was restored to operability after successfully completing the daily channel check. Licensee troubleshooting of the level transmitter identified a malfunctioning voltage to current converter. According to manufacturer information this converter contributes 0.25% of the instrument loop accuracy. Troubleshooting also identified a shif t (less than 4%) in transmitter output. Items reviewed during the inspection included licensee troubleshooting activit1ts, problem identification, level channel accuracies, dynamic level response during transients, and operability requirements of the channel. Based on a review of transmitter cross comparisons for-dynamic response during the August 27 reactor trip, successful completion of the cnannel check, and corrective maintenance activities, the in- , spector concluded that the level transmitter was now operable. At ' tne end of the inspection period, the licensee was conducting the required calibrations to further assess the transmitter and to determine what additional followup actions were necessary. The licensee's actions to address the steam generator transmitter level problem were acceptable. 4.6 Previously Identified Items 4.6.1 (Closed) Unresolved Item 88-24-05: Inadeouate Application of Acceptance Criteria for Heise Gauge Calibration This item was opened based on an inspector noted discrepancy in the licensee application of the acceptance criteria for procedure I/C 1104A "I&C pressure Test Gauges Calibration." Paragraph 2 of I/C 1104A defines the acceptance criterion for Heise gauges as +/-0.1% of
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P O BOX 270 k ' J N7)[.$'CE. w w %, . . . ,w... HARTFORD. CONNECTICUT 06141-0270 (203) 665-5000 t December 3, 1990 Docket No. 50-336 A09066 Hr. E. C. Venzinger, Chief Projects Branch No. 4 Division of Reactor Projects U. S. Nuclear Regulatory Commission Region I 478 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Venzinger:
Hillstone Nuclear Power Station, Unit No. 2 RI-90-A-136 Ve have completed our review of an allegation concerning activities at Hillstone Unit 2 (RI-90-A-136). As requested in your transmittal letter, our response does not contain any personal privacy, proprietary, or_ safeguards information. The material contained in this response may be released to the public and placed in the NRC Public Document. Room at your discretion. The NRC letter and our response have received controlled-and limited distribution on a "need to know" basis during the preparation of this response. Based upon our-request on October 25, 1990, Region I personnel extended the due date for this-response to December 3 1990. The basis for our extension request was the competing demands for time on personnel involved in these matters and ti.. thd ongoing refueling outage. In order to place this issue in perspective, the following chronological summary may be helpful. On April 18, 1990, troubleshooting activities under AVO H2-90-4154 identified a problem associated with proper operation of the flow control valve (FCV) of the RM 8262 system. During this activity, it was noted that the stem of the j FCV vas loose and that the FCV.needed replacement. AVO H2-90-04311 was written to replace the FCV. assembly. At this time, bypass jumper 2-90.17 vas installed to allow manual flow control, and the radiation monitor 8262 was restored to service. I I os_e h/hb
4 Hr. E. C. V:nzingtr, Chief U. S. Nuclear Regulatory Commission [ A09066/Page 2 December 3, 1990 During the in-place testing process associated with the replacement activity c* for the new FCV, leakage was_noted at the_ threaded connection on the valve body. The new valve was then removed from the system, and the old valve was reinstalled. It was then determined that the threads _on the replacement valve , had been inadequately tapped during the tine of manufacture. The threads were then fully tapped and the valve leak tested on the bench. The activity of ' tapping the threads and subsequent bench testing was not documented in AVO H2-90-14311. The second in-cycle outage occurred and activities on this work ordet vere delayed. The valve was then reinstalled in the system on June 28, 1990, and the system was returned to service. This was done without the completion of a leak test required by the AVO inspection plan. The supervisor and the individual assigned to this work order then vent on vacation. During subsequent , activities, it was noted that this work order was still open and the inspection plan incomplete. The leak test was performed satisfactorily on August 8, 1990, and the AVO accepted by Operations. During the period of June 29, 1990 through July 18, 1990, poor flow performance of the RM 8262 system was investigated by I&C and Maintenance. These activities were unrelated to the flow control valve. Allegations Numerous work orders and design changes have been performed on radiation monitors RM 8262 and RM 8123 during the period June through August'1990. The following items have been identified: : Item 1 2
- a. Vork on the flow control valve, including replacement, commercial grade dedication, and retesting was conducted under AVO H2-90-04311 vhile the AVO only authorized work on the system bypass valves.
Response
AVO H2-90-04311 was clearly written to replace tha flow control valve. The statement that it only authorized work on the bypass valve is inaccurate.
-b. The flow control valve failed an initial 60 psi leak check, but this-failure was not dispositioned.
Response
The disposition of the initial failure of the-leak test was performed but not properly documented. This-vas caused by ineffective communications l between the specialist conducting the work and his supervisor. The leak L test performed as part of the acceptance of the replacement valve during work order close out is properly documented on the AVO. l I
Mr. E..C. V:nzingar, Chief U. S. Nuclear Regulatory Commission I .A09066/Page 3 December 3, 1990
- c. There was not adequate communication between Maintenance and Operations to t s-ensure that proper actions were conducted when the monitors were inoperable.
- 1) Prom July 23, 1990 to August 4, 1990, neither monitor was operable, and compensatory actions were not taken. (The inoperability was due to in-progress system design change work).
Response
The statement that neither RM vas operable during the period of time between July 23, 1990 to August 4, 1990 is not accurate. A review of operator rounds indicates that RM 8123 was out of service July 27,- July 30, and August 2 to August 4, 1990 to upgrade the type of flov indicating switch. RM 8262 was operable during this time.
- 2) On July 26, 1990, RM 8262 vas out of service to change filters and RM 8123 was out of service due to a failed low flow alarm, but no Technical Specification Action Statement was logged, and no compensatory measures were taken. (Please address separately from 1).
Response. On July 26, 1990, both radiation monitors 8123 and 8262 vere in service. A normal particulate filter change took place at 1400 hours. This process was done in series. It was also done with prior communication with and authorization from Operations. This activity does not require any entry into Technical Specification Action Statements.;-therefore, none vere logged into, and no compensatory actions vere taken. ,
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- d. Drawings have not been updated as a result of the above design change work.
Response
PDCR MP2-90-032 was written to modify the flov indicating svitches for seven radiation monitors by replacing the Magnahelic flow indicating switch with a Photohelic flov indicating switch. Only one of these monitors, RM 8123, has had the modification made. No DCR has been processed at this time. 'The need to process DCRs at the time the system was returned to service was discussed with plant engineering management. As no operations critical drawing was affected, no'immediate changes were i deemed appropriate at that time. Item 2 i AVO M2-90-08033 installed a bypass jumper (jumper card #3) without , documentation. (No bypass jumper log entry and no PDCE). The tag was cleared l on August 31, 1990. I i
Mr. E. C. Mensinger, Chief U. S. Nuclear Regulatory Commission
'} A09066/Page 4 December 3, 1990 Response (
[ Background - AVO H2-90-08033 vas written to correct a report problem with the Metrascope position indication for CEA 18. During the work activities associated with the AVO, a broken edge connector was identified at pin location #1 on circuit card #3. The #1 pin connector was repaired by adding a ' piece of vire. It was identified later that the #1 pin connection was not used by the circuit. The edge connector was cleaned and indication was restored.] No specific questions were posed by this item. The addition of the small length of vire as a repair to the broken pin connector is not considered a plant modification. The use of a bypass jumper or design change administrative control was not considered necessary. Ve observe that there are some rather significant discrepancies between the allegations described above and the facts as ve understand them, as substantiated by the documentation which exists at the Hillstone site. This situation results in resources being expended on matters that have limited, if any, significance to them. Our review and evaluation have concluded that when taken either singularly or collectively, the t.llegations present no indication of a compromise of nuclear safety. Please contact my staff if there are any other questions on these matters. Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Y/ 81 < E. J. Mp'czka Senior Tice President cc: V. J. Raymond, Senior Resident Inspector, Hillstone Unit Nos. 1, 2, and 3 l f l l l
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The NRC Region 16fnce has completed its followup of the concerns that you brought to our attention on July 13,1990, and August 30,1990, alleging (1) an inadequate Are barrier, (2) inappropriate action related to oil leakage from the 'A' reactor coolant purnp (RCP), and (3) inadequate action on the licensee's part to minimize radiation exposure during the addition of oil to the reactor coolant pump. With ,= gard to issue (1), we determined that the subject fire barrier was adequately constructed and that your allegation was unsubstantiated. Our review is documented in inspection report 50 336/9014, section 2.2 (attached). With regard to issue (2), we determined that your allegation was substantiated because the
" A" RCP oil collection facility required some improvements. We are tracking licensee action to improve oil collection. Our review was documented in inspection report 50 336/9016, section 4.3 (attached).
l With regard to issue (3), we found that licensee action to irb ement ALARA principles was adequate, although some actions to improve work coordinawr, and exposure control activities for RCP oil addition were implemented followir.;. a cost jtsb review. Your allegation in this case, was unsubstantiated. Our review was documented in inspection repon 50 336/90 25.
- section 7 (attached).
We consider all of the abose items closed. We appreciate you in, ng us of your concerns and feel that ue have been responsive to those concerns. Should yw have any additional , questions regarding these matters, please call me collect at (215) 337 5225, i 'erely,
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Enclosures:
As stated bec: M. Perkins (2) RI-90-A 137. RI 88-A 0003(826)
- 1. Stewart W. Raymond s w sos e k
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* .50 4 3%/fe -# V a g system flow, initially, the flow control valve controller leads were t ,
lifted to disengage the flow control valve and tag 2-90-17 accurately reflected the condition. At a later date, the controller was de-energized and the controller leads were reattached, however, the tag , was not updated to reflect this condition. Neither system operability . nor personnel safety were affected by the discrepancy'and this issue is considered closed. No ether discrepancies were noted. 2.2 Plant Tours The inspector observed plant cperattens during regular and backshift tours of tre following areas: . Control Room Containment ' Vital Switchgear Room Diesel Generator Room Turtine Builcing Intake Structure Enclosure Building ESF Cubicles During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly race, and to verify correct communication and equipment status. No significant observations were noted. During the i su : tion, a question was raised as to the ability of a penetration tnru wall fire barrier nuecer A-4/T-1 to rneet its requirement to prevent spread of a fire. An NRC. regional specialist visitirsg the site on an unrelated inspection accompanied a utility , engineer on a walkdown of the penetration. It was concluded that the ' penetration reet the equirements for the given seal design. The penetration is' filled with grovt from the cable vault side and passes ' tnrough a 12-inch thick concrete block wali using 2 inches of damming material. The seal design reovires 8 inches of crout and 1 inch of camming rnaterial, hence, the minimum requirements were exceeded. 2,3 stand-by Readiness of Engineered Safety Features System and - System Waltecwn During the inspection period, two engineered safety feature (ESF) systems were reviewed to verify system operability. The systems reviewed were auxiliary feedwater and control room ventilation. The~ L review included proper positioning of major flowpath valves, proper l- operation of indication and controls, and visual inspection for. - proper lubrication, cooling, and other conditions. References used were:
-- Final Safety Analysis Report
-- Plant-instrument and piping diagrams-(P& ids) 25203-26005, and 25203-26027 L
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The inspector observed health physics controls and neutron source survey results, loose parts control on the spent fuel pool bridge, and surveillance testing of four poison boxes. The inspector t reviewed the i ndor's procedures and associated safety analysis. Questions on the safety evaluation were presented to cognizant Itcensee engineering personnel and were appropriately dispositioned. < Conclusion The control and implementation of neutron logging in the high density spent fuel pool poison boxes was completed efficiently. !*spection report detail 5.2 documents the blackness testing results the signi-ficance of the findings and the licensee's~ followup actions. 4.3 'A' Reacter Coolar.t Pump Muter Oil testece On Asgust 15 and August 27 the Itcensee added oil to tne upper reservoir of the ' A' reactor coolant pump (RCP) motor. The oil was added as a result of a downward trend in oil level indication since early August. A total of approximately six (6) gallons was added on the above cates. The reservoir has a ceDacity of approximately 120 gallons, ine licensee identified the cause of the oil loss to be leakage from
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a one inch fitting at the rear of the lift pump package. Approxi-rately 50'. of tre oil lost was collected in the oil collection tank, the renaining oil accumulated at the rear of the lift pump package, on the pump insulatian, and on the motor frame. Daring the current refueling outage, the licensee replaced the suscect fitting. The RCP oil lif t pump package will be inspected prior to start up during oil system operation at normal temperature , witt, tre motor operating. - 10 CFR 50 Appendix R Section !!!, Subpart 0 requires an oil collection system to be installed such that failure of a lube oil system will riot lead to a fire during normal and design basis events. Further, the collection system shall be capable of collecting lube oil frem all potential pressurized and unpressurized leakage sites in the RCP oil system. Leakage points to be protected shall include the lif t pump and piping, overflow lines and flanged connections on the oil reservoirs, in spite of the oil leakage in August, the upper reservoir did not'go below a point in level requiring operator action to protect the RCP motor. No other indication of inadequate performance of the RCP motor was noteJ during this time interval.
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b3 33/fa.f o 16 3 In summary, oil leakage out of the ' A' RCP motor upper reservoir was verified to collect both around the pump, and in the oil collection ' tank. 1he inspector vertfied oil around the pump during the early part of the cycle 10 refuel outage. The leakage in August of 1990 differed from past instances when the A pump lost oil, in that most ' of the oil lost was collected in the collection sumps during the prior events, The leakage point was identified, and licenste corrective actions are ongoing to correct the current problem. The inspector questioned the acceptability of the oil collection facility relative to the performance objectives prescribec in 10 CFR 50 Appendix R Section Ill, subsection O. The inspector' utoncern, based on the August 1990 leakage, is whether the oil col M iion facility is sufficient to prevent a fire hazard during RCP operation. This itet is unresolved pending licensee actions to improve the oil collection ior the RCP rnoter such that leakage from areas around the oil lif t pute package is adequately contained (50 336/90-18-06). 4 . t. hRC Information Hetic? 89 90. " Pressurizer Safety lift Setpoint $hift" The inspection consisted of follow-up of licensee actions related to NRC Information Notice 89-90. This information notice advised licensees cf rotential problems resulting from operating pressurizer safety' valves (PSV) in an environeett dif ferent from that used to estaolish the p5V lift setpoints, in October 1989 Westinghouse infor-red its plant c ners of a potertial deviation of the PSV set pressure from ine American Society of Mechanical Engineers ( ASME) section III and plant technical specification requirements for plants having loop seals upstrea, et the PSVs. Millstone 2 was originally constructed with loop seals between the pressuri:er and safety valves. In Asgust 1983, the licensee reteved the loop seals based on previous incustry results of the Electric Po.,er Researcn Institute (EPRI) study in 1982. The licensee removes the P$V and sends it to Vyle laboratories for testing and overhaul durinj each refueling outage. The Wyle Laboratories procedure for testing the PSV was reviewed together with the test results from the last surveillances done in February 1989. The valve trust undergo three consecutive acceptable lift setpoint tests at normal operating conditions for acceptance. The valve is subjected to leak tests prior and after the lif t setpoint tests.
. _ . _ _ . _ _ . . ~ _ _ _ _ . _ . __ _ __ _ _ ._ ' W11Wro-p( ,
s 10 t personnel regarding radiation protection coverage requirements. The permits principally served to inform workers of protective clothing ! requirements and dostmetry use requirements. The licensee has recogni2ed this matter as an area for enhancement. The licensee is revising the radiation work permit to include specific radiological controls coverage requirements. The inspector noted that the licensee has provided memoranda with expanded guidance to radiation protection personnel regarding - radiological controls reovirements for Unit I steam generator work activities. This was considered a good initiative.
-There is no procedure that provides guidance regarding installation, i operation, and surveillance of engineering controls (e.g., portable ventilation systems) used to minimize airborne radioactivity. The licensee has developed and is reviewing a draft procedure.
-Tne licensee's procedure for use of the Delmonox Breathing Air Supply Systerr contains an illegible graph that is to be used for determi-nation of proper air pressure to workers. Also, the graph appears to specify a breathing air hose length that is:not permitted. The id:ensee initiated an immediate review of the matter. Subsequent -
irspection revie. of work activities where the breathing air supply was being used indicated air pressure and hose lengths were correct. 4
-Tne licensee installed general area radiation survey meters (ARMS) on the Unit 2 steam generator platforms to alert personnel in the event that a hot particle was inadvertently removed f rom the generators during eddy current testing. The alarms of the ARMS were '
set at different alarm set-points (above background radiation levels). Also there were no periodic surveillances of the ARM and alarm set points to ensure they were working properly. The licensee I initiated a review of this matter. l 7.0 Worker Concerns (RI-90-A-137 item 2.b.) i l 7.1 General On August 22, 1990, a worker contacted NRC Region I and expressed I concern that the total radiation exposure received during a recent. oil addition to the A Reactor Coolant Pump was much higher than espected and,that no steps have been taken to reduce total radiation exposure during the oil addition,
7.2 Findings
The inspectors met with cognizant licensee personnel and discussed the addition of oil to the A Reactor Coolant Pump (RCP). The
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inspector. reviewed applicable documentation including radiation l l.
SD-136/90. ? $ i 11 s surveys and post-job critiques. The last addition was made on August 15, 1990. The licensee's ALARA personnel expected that the oil addition would result in an accumulated exposure of between 0.8 person-rem to 0.9 person-rem. This was about the exposure sustained , when oil was last added on October 12,1989(0.871 person-rem). The cumulative exposure estimate did not require a documented ALARA review. A pre-job meeting was held. At the pre-job meeting estimated radiation dose rates were discussed as well as activities to be performed, estimated stay time and heat stress requirements. The need to stay in low dose rate wait areas was discussed. Because of high radiation dose rates in the area and heat stress concerns, the oil addition was to be completed by three crens. The first crew was to remove the deck grating above the A RCF, install a ladder to the oil reservoir fill area, stage tools and leave. The second crew was to go down the ladder and fill the oil reservoir. A third crew was to assist. However, apparently through miscommuni-cation or error, the second crew removed the oil fill tube and removed the ladder when exiting the area. Since the original plan was to leave the oil fill tube in place so that edditional oil could be acoed if needed, a re-entry into the area to re-install the fill tube was rieeded. The total cumulation exposure as a result of re-installing the fill tube was 1.36 person-rem as ccmpared to the original estimate of ' between 0.8 and 0.9 person-rem. As a result of the problems encountered a post-job critique was held on August 16, 1990. The critique identified four recommencations whicn were subsecuently documented in a memorandum 19 the Unit 2 Maintenance Manager to address the problems encountered. An action request was issued by the Station Director on September 4, 1990, to review the exposure control and ALARA options for RCP oli addition at power. 7.3 Conclusion The inspector concluded that due to venknesses in pre-planning and or personnel error in failing to follow initial plans additional exposure was sustained by personnel to fill the oil reservoir of the A RCP, The inspector also concluded that the licensee recognized weaknesses in the performance of the task and initiated corrective actiens to review and improve exposure control activities for RCP oil addition.
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o 60*1%l9o.g$ l 12 F t The following corrective actions were noted:
-As discussed above, the Station Olrector issued an action item '
to review and improve RCP and additions. This occurred about ' two weeks af ter the event.
-A post oil addition critique was held the day following.the oli addition. Recommendations for corrective action were documented in an August 21, 1990 remorandum f rom the ALARA r.cordinator to the Unit 2 Maintenance Manager. "
-The oil leak on the ARCP was located and repaired.
-ine Meinse ance Foreman overseeing the oil addition was counselee regarding the breakdown in communication
-ine licensee initiated design reviews to change out hard piping anc install flexible piping for the RCP oil system to preclude leaking joints. ,
Essed on the above, the concern that no steps were taken to reduce total raciation exposure curing TCP oil addition is not substan- , tiatec. Th$s concerns is closed. 9.0 Exit Feeting The ir.spector met with licensee representatives (denoted in Section 1) on - October 5, 1990. Tne inspector summarized the purpose, scope and fincings of the intoection.
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s 476 ALLENDALE ROAD kJNQ oF PRUSSIA PENNSYLVANIA tf AM M' 1 7 1991 ( , Docket No. 50-336; RI-90-A-137;RI-90-A-175; Northeast Nuclear Energy Company ATTN Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and operations P.O. Box 270 Hartford, Connecticut 06141-0270 Dear Mr. Mroczkat This refers to your letters dated December 3, and December 7, 1990 (attached). Thank you for informing us of the results of your review of the subject allegations as documented in your letter. We are satisfied that descrepancies identified during the-maintainence activities described in the letter were minor in nature and have been corrected. Further, we-find no evidence of any programmatic deficiencies; therefore, we agree with your conclusion that none of these issues present any indication of a , comprumise of nuclear-safety. We consider these matters closed., A copy of_this letter with the attachments _is being placed in_the ' Public Document Room. We appreciate your cooperation in these matters. E; C. Wens Af r Chief Reactor Pr Branch 4 enclosures; As stated 7f/b/98D6W Q
cc v/encls W. D. Romberg, Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licensing. S. E. Scace, Station Director, Millstone Public Document Room (PDR) . Local Public Document Room (LPDR)
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{ , (f ' Public Document Room (PDR) Local Public Document Room (LPDR) HRC Senior Resident Inspector i gjl,h ' State of Connecticut ) k, y bcc w/encls: ) Allegation rile ( H th Q con .dr.7r D es) # l- to .g./57 J. Stewart p q - /*, i h ___ n: u _, 75= Ally e b4 fuir Al' 9o -st~t W C.w.H5.44 l 3
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~ Docket No. 50-336 A091D Mr. E. C. Venzinger, Chief l Projects Stanch Nu. 4 Division of Reactor Projects ,
U. S. Nuclear Regulatory Commission ' Region I 478 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Venringer:
H111 stone Nuclear Pover Station, Unit No. 2 RI-90-A-137 _ Ve'have completed our reviev of an allegation concerning activities at Hillstone Unit 2 (RI-90-A-137). As requested in your transmittal letter, our response does not contain any personal privacy, proprietary, or safeguards information. The material contained in this response may be released to the publie' and placed in the NRC Public Document Room at your discretion. The NRC letter and our response have received controlled and limited distribution on e "need to know" basis during the preparation of this response. . . Issue 1 During vork under AVO-H2-90-08697, the wrong fuses vere tagged when tagging l 2-MS-190B. Please explain. l , Response Vhen 2-MS-190B vas tagged for the work order ref erenced, the appropriate operations procedure was used to identify and tag out of service the power supply listed for this valve operator. When the electrician assigned to this job began vorking on the valve operator, he found that another circuit supplying tht valve operator vas energized. - The Operations; Department vas informed of this and investigated this situation. Thav found that another power supply..which previously had not been identified as supplying power to this valve operator, was indeed energized. This power supply (which controls the valve operator'in the proportional control mode) was subsequently tagged, and the work authorized by the work order completed. l 1
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r Hr. E. C. Venringer. Chief UI S. Nuclear Regulatory Commission A09123/Fage 2 December 3, 1990 Unit 2 Engineering then performed an investigation to ensure that all power t supplies to these particular valve operators vere identified in the appropriate procedures. The affected procedures have been revised to reflect the existence of the second power supply. Issue 2 A tagout of the "G" Demineraliter Inlet Motor-Operated Valve failed to remove 125V from the limit svitch contacts. P&lD 25213-32810, Sheet 20. does not show 125V to the limit svitch contacts. The Maintenance Supervisor for Hillstone Unit 2 vas informed of this discrepancy by the NRC Resident Inspector on August 23, 1990. Please discuss the alleged weakness of the plant dravings and the resulting tagout inadequacy. No response to the implied personnel safety issue is requested.
Response
The power supply for the motor on the "C" CPF Demineralizer inlet motor-operated valve operator was tagged by Operations to support a work order. The problem identified on the vork order stated that the " valve does not close." The work to be performed was " adjust or replace torque switch." Breaker 2-CND-V72 vas the only power supply listed in documents located in the control room. The work vas completed with the original tagouts. The tegout was proper for the work to be performed. No additional tagging was requested or needed to deenergize t'e 125V supply to the Jimit switch contacts in order to safely complete the troubleshooting and repair procedure. The valve in question is part of a vendor supplied package associated with the ' condensate polisher on Hillstone Unit 2. The circuit in question has been located on vendor drawings for this equipment. The circuit is part of the control system for the demineralizers and was not identified during the original development of the tag out list for these valves. The Unit 2 Engineering Department vill be reviewing the viring diagrams for each of the demineraliser motor-operated valves installed with this backfit to determine if any other anomalies exist. When this work is completed, control room dravings vill be revised as needed to identify the existence of other sources of 125V supplies to the valves, and the Operations Department vill make any necesrary changes to valve tagging procedures. Consideration vill also be
- given to labeling fuses which supply these circuits. In the interim, j operations personnel are avare of the potential for a 125V supply to be energized in these valves and vill confer vith Maintenance personnel to ensure that the power supplies are located and deenergized prior _to the conduct of any work on these valves. This concern has pointed out a potential personnel safety issue which has been addressed.
l Issue 3 Lights in the turbine building 45 ft. level have no P&ID. (No response to this issue is required).
Hr. E. C. Venringer. Chief , U. S. Nuclear Regulatory Commission A09123/Page 3 December 3, 1990
Response
t In order to provide a clarification to thin issue, the following discussion is provided for your information. The lighting circuit for these lights is indeed not shown on any electrical viring schematic drawing. There is a series of lighting distribution drawings (25203-350XX) which reflect the approximate physical locations of lamp fixtures and the circuits which supply them. These drawings provide sufficient detail regarding the source of power to the lights to allov safety tagging of the power supply for these circuits when these lights are deenergized for relamping and other mainter.ance activities. For the lights in the turbine building at elev. 45', Draving 25203-35038. Sheet 6, indicates that lighting panel L10 supplies these lights from several circuits. No further action is considered necessary, and none is planned. Issue 4 The on-call electrician on August 27, 1990 could not be contacted at 2:30 a.m. because the phone list in the control room was out of date. Please reviev and updato your on-call phone list.
Response
The on-call phone list has been updated. In this particular situation, the on-call electrician had recently sioved and changed telephone numbers. The phone list was in the process of being updated at the time of this event. The ' on-call electrician was carrying a radiopager and could have been contacted via this device. Control Room personnel did not attempt to contact the on-call individual in this manner, but rather contacted another electrician who vas not on-call via the telephone. : After our reviev and evaluation, ve find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety. Nonetheless, ve are pleased that the issues requiring attention on our part have been identified so that they can be corrected. Please contact my staff if there are further questions on any of_these matters. Very truly yours, NORTHEAST NUC1. EAR ENERGY COMPANY . ( A htl E. J. Mrk'ska e G/ Senior fice President cct V. J. Raymond, Senior Resident Inspector, Hillstone Unit Hos,1, 2, and 3
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HARTf ORD. CONNECTICUT OPD 0210 L t Ta 03.".il'.LT.',0~ 8 * *'5 6000 t i e December 7, 1990 Docket No. 50-336 A09TD Hr. E. C. Venringer. Chief Projects Branch No. 4 Division of Reactor Projects U. 5. Nuclear Regulatory Commission Region I 475 Allendale Road King of Frussia. Pennsylvania 19406
Dear Mr. Venzinger:
Hillstone Nuclear Fover Station. Unit No. 2 RI-90-A-17$ Ve have completed our review of an allegation concerning activities at Hillstone Unit 2 (RI-90 A-175). As requested in your transmittal letter. our-response does not contain any personal privacy.-proprietary, or safeguards information.- The material contained in this response may be released to the public and placed in the NRC Fublic Document Room at your discretion.' The NRC ' letter and our re=ponse have received controlled and limited distribution on a-
"need to know" basis during the preparation of this response.
Issue 1-On september 19.'1990, the on-call electrician was assigned duties as electrical coverage for reduced inventory condition while on shutdown cooling and drained to the centerline of the hot-leg for norsle dem installation. Reduced inventory coverage is a dedicated position as is the on call electrician. The on-call electtician vould be required to respond to station emergencies,
- s. Are these multiple assignments contrary to company procedures? Please explain.
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1 Hr. E. C. Venringer. Chief U. S. Nuclear Regulatory Commission A09127/Page 2
- December 7, 1990
Response
During shutdown periods with reduced reactor coolant inventory. H111 stone ' Unit 2 has established an on-site emergency response team of one electrician and five mechanics whose task is to establish containment integrity by closing the containment equipment hatch and the personnel access door in the event of loss of shutdovn cooling. The team provides 24-hour coverage and is required to respond and complete the tasks within two hours. Each of the team members carries a radiopager while on site to ensure that the team members can be contacted in a short period of time. Assignment to this team is not an emergency plan on-call assignment, and as such, there is no multiple on-call assignment issue. This assignment is not contrary to company procedures. Issue 2 On September 15. 1990, an electrician was assigned to disconnect the "A" heater drain pump and hang grounds on both the 4160V switchgear and locally on the pump motor. During verification of the AVO and the tagging. the following allegedly discrepant conditions vere identified
- 1. The wrong procedure was referenced in the AVO vibration data procedure rather than thu grounding of metal-clad switchgear procedure.
- 2. The motor heaters were not tagged.
- 3. The " refuel-heat" drain transfer vas tagged in the vrong position.
- 4. The AVO did not allov for grounding of the switchgear even though this is -
a necessary first step to disconnecting the motor. Please address these discrepant ennditions. If tagging discrepancies are identified, please discuss whether or not this is a recurring condition that . may require corrective action. Combined Response to Items 1 and 4 In revieving the work order for this job. it was noted that procedures referenced in the Procedures cautions section of the automated work order (AVO) vere not applicable for the work to be performed. This was caused by the fact that various procedure numbers may automatically appear in the procedure fields of the AVO. These numbers appear as a function of the equipment identification number used in the creation of the AVO. These procedure numbers appear as a convenience to aid in writing the AVO and are for information only. The Maintenance Supervisor and the Job Supervisor are responsible for reviewing the AVO for accuracy prior to the commencement of work, ap was done in thiS case.
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Mr. E. C. Venringer. Chief
. U. S. Nuclear Regulatory Commission' A09127/Page 3 r December 7, 1990 The original " Job Description" section of the vork order did not provide instructions to install grounds on the switchgear prior to disconnecting the motor leads. This error was brought to the attention of the Maintenance Supervisor by the Job Eupervisor prior to starti.ig the work. The Maintenance t Supervisor made pen and ink changes to the work orddr, cortecting the error and indicating that grounds were to be installed on the switchgear prior te proceeding with the vork described in the AVO. This revi9v and correction ,
process is consistent with the provisions of the governing station procedure. The changes made corrected the error, and the work order was correct and
- complete before the vork vas initiated.
Response Item 2 The motor heaters for the "A" heater drain pump were not initially tagged for this AVO. The heaters did not need to be tagged to safely perform the grounding of the heater drain pump side of the " refuel load center / heater drain pump transfer switch" (NA 105). After the safety tags vere placed and the AVO released for this work, the Job Supervisor requested that the "A" hwater drain rump motor heaters be tregtAd. Tags vere placed at that time. The reason the Job Supervisor requested tagging of the motor heaters was in anticipation of having to remove the entire A" heater drain pump at some later time. Response Item 3 The " refuel load center / heater drain pump transfer switch" vas not tagged in the wrong position. The heater drain pump transfer svitch vas tagged in the
" heater drain pump" position and never changed. Another tag was initially made out for the transfer switch in the " refuel load center" position. That tag was not needed, was never placed on the transfer switch, and was destroyed.
After our review and evaluation, we find that none of these issues taken ' either singularly or collectively present any indication of a compromise of nuclear safety, Ve appreclote the opportunity to respond and explain the basis for our actions. Please contact my staff if there are any further questions on any of these matters. Very truly yours, NORTHEAST NUC15.AR ENIkGY COMPANT Z A/ E.J./hoczka g/ Senior Vice President cci V. J. Raymond. Senior Resident Inspector, N111 stone Unit 140s.1, 2, and 3
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Projects Bianch Nu. 4 Division of Reactor Projects , U. S. Nuclear Regulatory Commission Region I 470 Allendale Road King of Prussia, Pennsylvania 19406 Dear Hr. Venfingert i Hillstone Nuclear Power Station, llnit No. 2 RI-90-A-137 _ Ve have completed our review of an allegation concerning activities at Hillatone Unit 2 (R1-90-A-137). As requested in your transmittal letter,_our response does not contain any personal privacy, proprietary, or safeguards information. The material contained in this response may be released to the The NRC- , pu'olic and placed in the NRC Pubile Document Room at your discretion. ' letter and our response have_ received controlled and-limited distribution on a #
"need to knov" basis during-the preparation of this response, j i
Issue 1 During vork under AV0-H2 90-09697, the vrong fuses vere tagged when tagging J. 2-MS-190B. Please explain. l o
-Responge;
- When 2-MS-190B vas tagged for the work order referenced, the appropriate:
operations procedure vas used to identify:and tag out of service the power. , supply listed for this valve operator. When the electrician: assigned to this job began verking on the valve operator. he found that another circuit supplying the valve operator vas-energized. The Operations Department _vas informed of-_this and investigated this-situation. They found that.another , power supply, which previously had not been identified as supplying-power to l-this valve operator, was indeed energized.- This pover supply (which controls ' the valve operator in-the proportional-control mode) was. subsequently tagged,: H . and the work authorized by the vork order completed.-
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r 1 ( e 6-Mr. E. C. Venringer. Chief U'. S. Nuclear Regulatory Commission A09123/Page 2 December 3, 1990 Unit 2 Engineering then performed an investigation to ensure that all power c supplies to these particular valve operators vere identified in the , appropriate procedures. The affected procedures have been revised to reflect the existence of the second power supply. , 1ssue 2 A tagout of the "G" Demineraliser inlet Motor-Operated Valve f ailed to remove 125V f rom the linit switch contacts. P&ID 25213-32810, Sheet 20, does not shov 125V to the limit svitch contacts. The Haintenance Supervisor for Hillstone Unit 2 vas informed of this discrepancy by the NRC Resident Inspector on August 23, 1990. Please discuss the alleged weakness of the plant drawings and the resulting tagout inadequacy. No response to the implied personnel safe., issue is requested.
Response
The power supply f or the motor on the "G" CPF Demineralizer inlet motor- The operated valve operator was tagged by Operations to support a work order. problem identified on the work order stated that the " valve does not close." > The work to be performed was " adjust or replace torque svitch." Breaker 2-CND-V72 was the only power supply listed in documents located in the control room. The work vas completed with the original tagouts. The tagout was proper for the work to be performed. No additional tagging was requested or needed to deenergize the 125V supply to the limit switch contacts in order to safely complete the troubleshooting and repair procedure. The valve in question is part of a vendor supplied package associated with the ' condensate polisher on Millstone Unit 2. The circuit in question has been located on vendor drawings for this equipment. The circuit is part of the control system for the demineralizers and was not identified during the original development of the tag out list for these valves. The Unit 2 Engineering Department vill be reviewing the viring diagrams for each of the demineralizer motor-operated valves installed with this backfit to determine if any other anomalies exist. Vhen this work is completed, control room drawings vill be revised as needed to identify the existence of other sources of 125v supplies to the valves, and the operations Department vill make any. necessary changes to valve tagging procedures. Consideration vill also be In the interim, given to labeling fuses which supply these circuits. Operations personnel are avare of the potential for a 125V supply o be energized in these valves and vill confer with HaintenanceL persont,'1 to ensure that the power supplies are located and deenergized prior to the conduct of any work on these valves. This concern has pointed out a potential personnel safety issue which has been addressed. Issue 3 Lights in the turbine building 45 f t. level have no P&ID. (No response to this issue is required).
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Mr. E. C. Venzinger, Chief U'. S. Nuclear Regulatory Commission A09123/Page 3 December 3, 1990 g
Response
In order to provide a clarification to this issue, the following discussion is provided for your information. The lighting circuit for these lights is indeed not shown on any electrical viring schemstic draving. There is a series of lighting distribution drawings (25203-350XX) which reflect the approximate physical locations of lamp fixture? and the circuits which supply them. These drawings provide sufficient detail regarding the source of power to the lights to allow safety tagging of the power supply for these circuits when these lights For are the lights in deenergized for relamping and other maintenance activities. the turbine building at elev. 45', Draving 25203-35038, Sheet 6, indicates No that lighting panel L10 supplies these lights f rom several circuits. further action is considered necessary, and none is planned. Issue 4 The on-call electrician on August 27, 1990 could not be contacted at 2:30 a.m. because the phone list in the control room was out of date. Please reviev and update your on-call phone llat.
Response
The on-call phone list has been updated. In this particular situation, the on-call electrician had recently moved and changed telephone numbers. The The phone list was in the process of being updated at the time of this event. on-call electrician was carrying a radiopager and could have been contacted via this device. Control Room personnel did not attempt to contact the on-call individual in this manner, but rather contacted another electrician ' who was not on-call via the telephone. After our reviev and evaluation, ve find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety. Nonetheless, ve are pleased that the issues requiring attention on our part have been identified so that they can be corrected. Please contact my staf f if there are further questions on any of these matters. Very truly yours, l i NORTHEAST NUCLEAR ENERGY COMPANY I $M ?. / ! E. J. Hracika J/ Senior ice President I cci V. J. Raymond Senior Resident Inspector,. Hillstone Unit Nos,1, 2, and 3 l l L
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i December 7, 1990 Dorket No. 50-336 A09076 Mr. E. C. Venringer, Chief Projects Branch No. 4 Division of Reactor Projecta U. S. Nuclear Regulatory Cc,mmission Region I 475 Allends.le Road King of Prussia, Pennsylvania 19406
Dear Hr. Venringer:
Millstone Nurlear Power Station, Unit No. 2 RI-90-A-144 Ve have completed our reviev of an allegation concetning activities at Hillstone Unit 2.,'ll-90-A-144). As requested in your transmittal letter, our response does not contain any personal privacy, proprietary, or safeguards information. The material enntained in this response may be released to the , public and placed in the NRC Public Document Room at your discretion.- The NRC letter and our response have received controlled and limited distribution on a "need to know" basis during the preparation of this response. Based upon our request on October 25, 1990, Region I personnel extended the due date for this response to December 7, 1990. The basis for our extension request was the competing demands for time on personnel involved in these matters and the then ongoing refueling outage. Item 1 Steam Jet Air Ejector (SJAE) radiation monitor, RH 5099, does not appear to be collecting vater. The monitor response may have decreased. The Plant Equipment operator (PEO) has not emptied the collection bottle-for ten days (as of September 14, 1090), although normally, four quarts of water are emptied per day. The monitor has been reading 2,000 cpm and has decreased to 1,000 cpm. f h /$%
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s Mr. E. C. Vensinger, Chief U. S. Nuclear Regulatory Commission A09076/Page 2
- December 7, 1990 Background Information The SJAE RH monitors the gaseous activity in the SJAE exhaust. The system g includes a moisture removal system that dttes the gas going to the detector e pig. The output of the moisture removal system is manually removed from collection bottles as part of the Plant Equipment Operator rounds. The moisture content of the gas exiting the SJAE is largely dependent upon the performance of the SJAE drain system. Vhen thin SJAE drain system is operating effectively, there is essentially no moisture to be removed. Vhen .
the SJAE is not being effectively drained, the moisture content in the SJAE exhaust increases substantially. As a result, the amount of moisture removed by the RM moisture collection system increases as well. If the moisture is not effectively removed from the gas, it collects in the detector pig and causes the detector to fail. Excessive moisture in the SJAE exhaust has been a long-standing problem with the operation of this system. A replacement rhulation monitoring system that is external to the process piping is currently being designed (Project Assignment 89-053). This vill allow reliable radiation monitoring without compromise by the moisture level of the SJAE exhaust. This allegation questions the response of the RH during the time period of September 4 through 14 of 1990. During this period, the unit operated at 100% pover. During the previous week, the unit had undergone a startup that included a reviev'of the valve alignment of the SJAE drain system. Changes in the valve alignment at that time significantly improved the moisture removal effectiveness of the SJAE. The result of this improved drain performance was the lov output of moisture from the RH moisture removal system. .This lov output of moisture is a normal response for the RH system when these conditions exist. The activity level of the gas at the SJAE is dependent upon the amount of the , gaseous activity of the RCS, the amount of primary to secondary leakage, and 3-the amount of air inleakage into the condenser. During this time period, the RM readings varied between 1,000 and 2,400 cpm. Chemistry samples indicate that SJAE activity was 2 E-5 uci/ml to 6 E-5 uct/ml respectively. This shows a close correlation betveen actual sample values and RH response,
- a. Should Operations have determined there to be a problem?
Response
The RH performance during this time period does not represent a problem. The response of the RM was correct for the plant conditions present. _The secondary side conditions, including the SJAE activity levels, are appropriately monitored by Operations. This includes review by Plant Management and Chemistry staffs on a daily basis. L
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4 Nr. E. C. Usnringar, Chief it. S. Nuclear Regulatory Commission A09076/paFe 3 L December 7, 1990
- b. Have appropriate enmpensatory measures been taken by Opetations?
Response '. , No additional actions were necessary.
- c. Have evaluations been done to ensure the monitor is operating properly or is reliable?
Response
Yes. The RH's performance is reviewed on a daily basis by the Operations, Management, and Chemistry staffs. Vhen the monitor was removed from service at the next outage, it was verified to be dry and in proper working order. The long-term reliability issue is being addressed by the replacement of the RH vith a non-intrusive system, which.is not subject to moisture damage. Short-term reliability of this RM has been enhanced by the upgrading of the moistute removal system capacity and additional Operations' awareness of the need for optimum performance of the SJAE drain system.
- d. If problems have been noted, what actions have been taken to evaluate, compensate, track, and resolve the problems? Have all appropriate trouble reports or other documents been completed or filed?
Response
Ve currently plan to replace the RM vith an improved design in 1991. ' The RH vas calibrated during the outage as part of its normal maintenance. Bypass jumpers remain in place that have increased the moisture removal capability of the system. All work has been conducted in accordance with
- station procedures.
- e. Have Operations and I&C personnel performed the appropriate duties with regard to the radiation monitor? (The Shift Supervisor ~and an I&C supervisor have been made aware that there may be a problem.)
Response
Yes, appropriate duties vere performed. f. What actions are planned to ensure continued reliability and performance?
Response
Short term - Additional attention has been placed by Operations on the proper operation of the SJAE drain system. Performance reviev activities-vill be continued by Operations. Management, and Chemistry staff perr.onnel. A temporary SJAE radiation monitor is being used to. provide diverse indication. l l_ , , _ . - . - ~ - -
N Mr. E. C. Venringer, Chief
- U. S. Nuclear Regulatory Commission A09076/Page 4 December 7, 1990 ,
reliable system being designed and procured at1,ong term - The RH this time. Item 2 L his concerns within a 14-day administrative requirement.The one of , alle raised by the alleger on August 23, 1990 and involvedThe concern three was issues 1. use of the " completed by" block for prerequisites, 2. another channel, during testing, anda procedure statement about required acti -
- 3. break times.
Items (1) meeting. and Item (2) (3) waswere discussed not discussed. at a September 5,1990 I&C Department Background Information A Unit 2 I6C Department meeting was held on August material was covered. 23, 1990. A variety of Numerous questions vere fielded from the attendees, n. What is your policy in this regard?
Response
Pertinent questions asked during Department meetings for nswer must be researched are addressed. which an a . process. No time requirements exist for this information and discussing issues in an open interest to the entire Department. environment which are of , b. Does the complaint have validity, and if so, why vasn't the employee's concern addressed? yRonse The issue does not have validity. The questtor (i.e. Item 2 of this allegation) the Departmentvas asked Manager bytime. at that a member of the I&C Department and addressed the need for specific follov-up to this question.The Department Manager did not note The verbal response provided by the Department Manager to Item 2 of this allegation discussed the Unit 2 I&C procedure guidance regarding surveillance testing . _- ~
s <. Mr. E. C. Vanzinger, Chief U. 5. Nuclest Regulatory Commission A09076/Page 4
- December 7, 1990 L.ong term - The RH is planned to 'ue replaced in 1991 vith the more reliable sys.em being designed and procured at this time.
t Item 2 The alleger has complained that an ILC Supervisor failed to respond to one of , his concerns within a 14-day administrative requirement. The concern was raised by the alleger on August 23, 1990 and involved three issues:
- 1. use of the " completed by" block for prerequisites,
- 2. a procedure statement about required actions if RPS pretrips come in on another channel, during testing, and
- 3. break times.
Items (1) and (3) vere discussed at a September 5, 1990 I&C Department meeting. Item (2) vas not discussed. Background Informat f orj A Unit 2 I&C Department meeting was held on August 23, 1990. A variety of material was covered. Numerous questions were fielded from the attendees,
- a. Vhat is your policy in this regard?
Response
4 Pertinent questions asked during Department meetings for which an answer must be researched are addressed. No time requirements exist for this process. The Department meeting forum is used as a means of sharing information and discussing issues in an open environment which are of . interest to the entire Department.
- b. Does the complaint have validity, and if so, why vasn't the employee's concern addressed?
Response
The issue does not have validity. The question (l'.e., Item 2 of this allegation) was asked by a member of the I&C Department and addressed by the Department Manager at that time. The Department Manager did not note the need for specific follov-up to this question. The verbal response provided by the Department Manager to Item 2 of this allegation discussed the Unit 2 I&C procedure guidance regarding surveillance testing. _ . . ;, _ . _ . , _ _ .. _ -. a
Hr. E. C. Benziaget. Chief U. S. Nuclear Regulatory Commission A09076/Page 5 6 December 7, 1990 Item 3 A worker was assigned to troubleshoot a problem with the control circuit for t the turbine bypass valves under AVO 90-09684 on September li, 1990. 'Jork had previously been performed on this system under AVO 90-06498 and AVO 90-07792. This work identified deficiencies with PT 4300 which vere to be corrected , during the outage.
- a. The initial AVos may not have been documented in the appropriate loop folder. Is there a requirement to do so? Vhy vasn't this requirement met?
Response
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There is no station requirement to maintain a loop folder. Vork activities that are performed are required to be documented on the AVO document. Unit 2 ILC has a Department Instruction 1.10 that defines the contents of the department loop folders. It includes the I&C specialists' and technicians' responsibility to maintain a brief handwritten vork history in the loop folder. It is expected that all Unit 2 I&C personnel vill keep the loop folder information up to date. This allows a more efficient votk history review than can be obtained from the PMMS system or the Nuclear Plant Records Facility. AVO M2-90-06498 was written to troubleshoot a problem associated with a reactor regulating system alarm. Since this was written against the system ID, there is no loop folder to record information in. The AVO that governed the specific calibration activity on PT 4300 was updated with the necessary information. All Unit 2 I&C Department expectations were met with respect to loop folder entries,
- b. An orange sticker was not affixed to the instrument PT 4300 contrary to .
Department instruction causing a redundant trouble report to be issued. Is there a requirement to affix an orange sticker when a deficiency is _ identified with an instrument? Vhy wasn't that requirement met?
Response
ACP-0A-2.02C, "Vork Orders", contains step 6.1.2 that states, "Vhere possible, place problem report tags / stickers on the component or other conspicuous location, which indicate the problem has been reported." The requirement was met for each trouble report. Note that the AVO problem description reflected the problem as it was seen by the Operations staff. Vork order tags vere appropriately hung, as indicated on the associated AV0s. The location of the tag is typically at the location that is conspicuous to the Operations Department staff. In this case, it was properly located in the control room by the alarm vindow that is associated with the reactor regulating system.
D Hr. E. C. Uenzinger, Chief U. S. Nuclear Regulatory Commission A09076/Page 6 L December 7, 1990 Note the seennd AVO, H2-90-09CB4, is not a " redundant" trouble report. The first AVO, H2 90-06498, identified the alarm condition as being caused by the pressure transmitter and rerommended its calibration. The second s AVO describes an additional problem, that of the recent decrease in the Trei value as well as the previously existing alarm condition. The information that PT 4300 had already been identified as needing calibration was included in the job description section of AVO H2-90-09684. The work activities of this AVO resolved problems associated with both the recorder and the pressure transmitter viring. AVO H2-90-07792 performed the calibration activity. After our review and evaluation, ve find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety, Ve appreciate the opportunity to respond and explain the basis for our actions. Please contact my staff if there are further questions on any of these matters. Very truly yours, NORTilEAST NUCLEAR ENERGY COMPANY P-E. J.proczka Senim- Vice President cct V. J. Raymond, Senior Resident inspector, Millstone Unit Nos. 1, 2, and 3 1 l l c l' i i
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~ f REOloN I 6 [ 475 ALLENDALE MoAD KING oF PMUSSIA, PENNSYLVANIA 1HM dAU ! 7 1991 \. Docket No. 50-336; - ' RI-90-A-137;RI-90-A-175; Northeast Nuclear Energy Company ATTH: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations i P.O. Box 270 ~ Hartford, connecticut 06141-0270
Dear Mr. Mroczka:
This refers to your letters dated December 3, and December 7, 1990 (attachad). Thank you for informing us of the results of your review of the subject allegations as documented in your letter. We are satisfied that descrepancies identified during the maintainence activities described in the letter were minor in nature and have been corrected. Further, we find no evidence of any programmatic deficiencies; therefore, we agree with your conclusion that none of these issues present any indication of a ' compromise of nuclear safety. We consider these matters closed. A copy of this letter with the attachments'is being placed in the Public Document Room. We appreciate your cooperation in these matters. e ; E. C. Wenz gr Chief , Reactor Pr Branch 4. enclosures; As stated hl lY h 1%
e i cc v/encls 3 -- W. D. Pomberg,'Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licensing S. E. Scace, Station Director, Millstone Public Document Room (PDR) C Local Public Document Room (LPDR) 11RC Senior Resident Inspector i State of Connecticut . 9
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i OFFICIAL RECORD COPY G cc v/encis: W. D. Romberg, Vice President, Nuclear Operation- t R. H. Kacich, Manager, Generation Facilities Lit'ising ' S. E. Scace, Station Director, Millstone Public Document Room (PDR) . , Local Public Document Room (LPDR) NRC Senior Resident Inspector State of Connecticut bec w/encis: Allegation rile (with concurrences) R l- 70 -A-/5 7 F m c.w.s A ry ar-4
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s n December 7, 1990 Docket No. 50-336 A09127 ' Mr. E. C. Venzinger. Chief Projects Branch No. 4 Division of Reactor Projects
'J. S. Nuclear ReFulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
;ar Mr. Venzinger:
Millstone Nuclear Pover Station, Unit No. 2 RI-90-A-175 have completed our reviev of an allegation concerning activities at iillstone Unit 2 (RI-90-A-175). - As requested-in your _ transmittal letter, our response does not contain any personal privacy, proprietary, or. safeguards-information. The material contained in this response may:be_ released to the- , _public and placed in the NRC Public Document-Room at your: discretion. The'NRC- '
.le.ter and our re=ponse have received controlled and_ limited distribution on-a "need to know" basis during the preparation of this -response.
( 1ssue 1 l L _On September 19, 1990, the on-call electrician was assigned duties as electrical coverage for reduced inventory condition while on shutdown cooling and drained to the centerline of - the hot-leg for nozzle dam' installation. Reduced inventory coverage is a dedicated position as is the on-call electrician. The on-call eleettician vould be required to respond to' station-emergencies. L a.' Are these multiple assignments contrary to company procedures? Please-p explain. , S ~ osS412 ptv sat
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_ Mr. E. C. Ven:inger, Chief 1 U. S. Nuclear Pegulatory Commission A09127/Page 2 l December 7, 1990 l 6
Response
i During shutdown periods with reduced reactor coolant inventory, Millstone t Unit 2 has established an on-site emergency response team of one ) electrician and five mechanics whose task is to establish containment integrity by closing the containment equipment batch and the personnel , l access door in the event of loss of shutdown cooling. The team provides ! 24-hour coverage and is required to respond and complete the tasks within i two hours. Each of the team members carries a radiopager while on site to ' ensure that the team members can be contacted in a short period of time. Assignment to this team is not an emergency plan on-call assignment, and as such, there is no multiple on-call assignment issue. This assignment is not contrary to company procedures. , Issue 2 On September 15. 1990, an electrician was assigned to disconnect the "A" heater drain pump and hang grounds on both the 4160V switchgear and locally on the pump motor. During verification of the AVO and the tagging, the following allegedly discrepant conditions vere identified:
- 1. The vrong procedure was referenced in the AVO vibration data procedure rather than the grounding of metal-clad switchgear procedure.
- 2. The motor heaters vere not tagged.
3 The " refuel-heat" drain transfer vas tagged in the vrong position. 4 The AVO did not allow for grounding of the svitchgear even though this is a necessary first step to disconnecting the motor. Please address these discrepant ennditions. If tagging discrepancies are identified, please discuss whether or not this is a recurring condition that may require corrective action. Combined Response to Items 1 and 4 In reviewing the work order for this job, it was noted that procedures referenced in the Procedures Cautions section of the automated work order (AVO) vere not applicable for the votk to be performed. This was caused by the fact that various procedure numbers.may automatically appear in the procedure fields of the AVO. These numbers appear as a function of the-equipment identification number used in the creation of the AVO. These procedure numbers appear as a convenience to aid in writing the AVO and are for information only. The Maintenance Supervisor and the Job Supervisor are responsible for reviewing the AVO for accuracy prior to the commencement of vork, as was done in this case.
p . Ucnzinger, Chief p . Nuclear Regulatory Commission *
\ A00127/Pago 3 i
. Dec=ber 7, 1990 L
'6 The original "Jnb Description" section of the work order did not provide instructions to install grounds on the switchgear prior to disconnecting the motor leads. This error was brought to the attention of the Maintenance Supervisor by the Job Supervisor prior to starting the work. The Maintenance Supervisor made pen and ink changes to the work order, correcting the error
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and indicating that grounds were to be installed on the svitchgear prior to proceedir.g with the work described in the AVO. This review and correction ' process is consistent with the provisions of the governing station procedure. The changes made corrected th.e error, and the work order was correct and complete before the vork vas initiated. Response Item 2 The motor heaters for the "A" heater drain pump were not initially tagged for this AVO. The heaters did not need to be tagged to safely perform the grounding of the heater drain pump nide of the " refuel lot.d center / heater drain pump transfer svitch" (NA-105). After the safety tags vere placed and the AVO released for this work, the Job Supervisor requested that the "A" heater drain pump motor heaters be tagged. Tags vere placed at that time. The reason the Job supervisor requested tagging of the motor heaters was in anticipation of having to remove the entire "A" heater drain pump at some later time. Response Item 3 The " refuel load center / heater drain ;ap transfer svitch" vas not tagged in the wrong position. The heater drrin pump transfer svitch was tagged in the
" heater drain pump" position and never changed. Another tag vas initially made out for the transfer svitch in the " refuel load center" position. That -
tag vas not needed, was never placed on the transfer svitch, and was destroyed. Af ter our review and evaluation, ve find that none of these issues taken < either singularly or collectively present any indication of a compromise of nuclear safety. Ve appreciate the opportunity to respond and explain the basis for our actions. Please contact my staff if there are any further questions on any of these matters. Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
/
E.J.fttoczka y Senior Vice President cc: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 l
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*=.** KINO oF PAVSstA, PENNSYLVANIA 18404 Docket No. 50 245; 50 336; 50-423 License No. DPR-21; DPR 65; NPF-49 EA No. 90 219
(' Northeast Nuclear Energy Company , ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations P.O. Box 270 Hanford, Connecticut 06141-0270 Gentlemen: .
Subject:
NRC Region I Combined Inspection Nos. 50-245/90-20, i 50-336/90-22, and 50 423/90-20 l This letter transmits the NRC repon of our routine safety inspection that was conducted by I Messrs. D. Dempsey P. Habighorst, and K. Kolaczyk of this of6ce on September 18 - Nosember 15. October 2 - November 15 (and continued December 3-13 to evaluate funher one of the signi6 cant issues described herein), and October 16 - November 15,1990, for Millstone Units I,2, and 3, respectively. At the conclusion of the inspection the findings were discussed by the above inspectors with Mr. S. E. Scace and other members of your staff. Areas examined during the inspection are described in the NRC Region I inspection report w hich is enclose,d with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, and document reviews. < Based on the results of this inspection, two apparent violations were identi6ed at Millstone. , Unit 2 and are being considered for escalated enforcement action in accordance with the
" General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Pelicy),10 CFR .$an 2, Appendix C (1990). The apparent violations involve the loss of conta+. ment integnty control as a result of personal errors and are discussed in the Unit 2 cperations and outage sections of the enclosed repon, Specifically, the apparent violations invohe the loss of containment integrity due to the inoperability of the containment purge valve isolation system, and in a separate event, the loss of containment integrity via the No. I steam generator atmospheric dump valve. Accordingly, no Notice o.f Violation is presently being issued for these inspection nndings. Please be advised that the number and characterization of apparent violations described in the enclosed inspection repon may change as a result of funher NRC review.
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Nonheast Nuclear Energy Company E26E
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An enforcement conference to discuss these apparent violations at Millstone Unit 2 has been scheduled for January 15, 1991. The purposes of this conference are to discuss the apparent violations, their causes and safety significance; to provide you the opponunity to point out any errors in our inspection repon; to provide an opportunity for you to present your C proposed corrective actions; and to discuss any other infornation that will help us determ ' the appropriate enforcement action in accordance with the Enforcement Policy. You will be advised by separate correspondence of the results of our deliberations on this matter. No ' response regarding these apparent violations is required at this time. The enclosed report addresses your performance during the recent refueling and maintenance outage on Millstone 2. Overall, we found the control of outage activities to be good, with effective management of planned activities and aggressive followup of problems, ne thorough evaluatior, of unplanned events, the extensive support by corporate engineering and vendors to disposition of these issues, and the effective interface between site and corporate engineering were notable strengths. Your assessment of the personnel performance aspe these events was requested in our letter to you dated November 5,1990, enclosing Inspection Report 50 336/90-18. . Nomithstanding the above conclusion regarding generally good performance, we noted a number of events anributed to personnel error, that apparently resulted from pro:cdure quality and adherence weaknesses. Funber, the failure to satisfactorily complete a critical step during the replacement of in core instruments that resulted in the dropping of the inco instrument support plate was signincant (see section 9.2 of the enclosed repon). Our assessment was that the lift tool installation error resulted from a con 6ination ofinadequacies in procedure details, personnel experience, and supervision of the work activity. The event demonstrates the need for greater diligence in the review process for "tried and proven" procedures to eliminate any over-;eliance on personnel experience for ethical activities. Based on the reiults of this inspection at Millstone Unit 1, certain of your activities appeared to be in violanon of NRC requirements, as specified in the Notice of Violation enclosed herewith as Appendix A. We are concemed about the violation because it involved the operation of Millstone I with non-conservative setpoints on the steam jet air ejector radia monitor. You are required to respond to this violation and should follow the instructions spccined in Appendix A when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrente. Afar reviewing your response to Appendix A, includiug your proposed corrective actions and the results of future inspections, the NRC will deteriaine whether funher NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements. In addition, cenain of your activities at Millstone Unit 2 appeared to be in deviation from your written comtritments, as specified in the Notice of Deviation enclosed herewith a Appendix B. We are concerned about the rieviations because they involved the failur 2
4 DEC 2 013N'
- Northeast Nuclear Energy Company reactor protection channels and to operate the loose parts monitor in accordace with Final Safety Analysis Report commitments. You are requested to respond to these deviations, and should follow the instructions specified in Appendix B in preparing your response.
In accordance with 10 CFR :1.790 of the NRC's Rules of Practice," a copy of this letter and \f its enclosures will be p! aced in the NRC Public Document Room. I The responses directed by this letter and the enclosed Notices are not subject to the clearance procedures of the Office of hitnagement and Budget a.s required by the Paperwork Reduction Act of 1980, Pub. L No. 96.511. I Your cooperation with us is appreciated. Sincerely, ,
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O' Wm ~ gh fes ' Hehl'
}frector ,
pj ision of Reach, rTrojects i Enc!csures:
- 1. Appendix A, Notice of Violation
- 2. Appendix 8, Notice of Deviadon '
- 3. NRC Region 1 Combined Inspection Report No. 50-245/90-20; 50-336/90-22 50-423/90-20 cc w/encis:
W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Direcwr of Quality Senices
- R. bl. Kacich, blanager, Generation Facilities Licensing S. E. Scace, Station Director, Millstone H. F. Haynes, Nuclear Unit Director, hiillstone Unit 1 J. S. Keenan, Nuclear Unit Director, Millstone Unit 2 C. H. Clement, Nuclear Unit Director, hiillstone Unit 3 Gerald Garfield, Esquire Public Document Room (PDR)
Local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) NRC Senior Resident Inspector 1 Sta:e of Connecticut 3 a a .-- m.n- . .
1
, q U..S. NUCl. EAR REGULATORY COMMISSION REGION I j
- l 50-245/90 20; 50 336/90-22; 50-423/90-20 !
Report No.: ! I Docket No.: 50-245; 50-336; 50-423 t ;! License No.: DPR 21; DPR-65; NPF-49 .,, " Licensee: Northeast Nuclear Energy Company P. O. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Units 1,2, & 3 Inspection At: Waterford, CT Dates: September 18 November 15,1990 (Millstone 1) October 2 November 15, and December 3 - 13,1990 (Millstone 2) October 16 - November 15,1990 (Millstone 3)- Reporting Inspectors: D. A. Dempsey, Resident Inspector, Unit 1 P. J. Habighorst, Resident Inspector, Unit 2 K. S. Kolaczyk, Resident Inspector, Unit 3 p Inspectors: W. J. Raymond, Senior Resident inspector ' D. A. Dempsey, Resident Inspector, Unit 1 P. J. Habighorst, Resident Inspector, Unit 2 K. S. Kolaczyk, Resident Inspector. Unit-3
- 3. S. Stewart, Senior Project Engineer .,
A. Vegel, Reactor Engineer Approved by: skb - /2/pJ/9a l Donald R.11averkamp, Chief [ Date l
. Reactor Projects Section 4A Division of Reactor Projects L lorection Sumimtty Report 50-245/90-20; Report 50-336/90-22; Report 50-423/90-20
..J' Arm InsMr. tid; Routine NRC resident inspection of plant opeTatiens, radiological controls, maintenance, surveillance, security, outage activities, licensee self assessment, and periodic reports.
l Besults; See Executive Summary
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m. 13-3,6 ImpropqLTaning Concern - Unit 2 On October 19, the inspector reviewed a concern regarding improper equipment isolation controls. Specifically, no local tar,s were hung on motor. t operated valves 2 RC-403 and 2 RC-405 (pressurizer reliefisolation) during ' maintenance work. The valves are considered to be boundary isolation valves, as defined in procedure ACP-QA 2.06A, section 6.1.8 and, thus should have been tagged locally. No local tags were hung. The inspector noted that ACP-QA 2.06A, section 6.1.8, requires that, in addition to normal equipment tagging, local operators of motor and pneumatically operated valves be tagged when the valve is used as a system isolation boundary point. If the local operator is in a high radiation area, pincement of safety tags is left to the discretion of the shift supervisor, senior control operator, or job supervisor. The inspector reviewed the tagouts for valves 2 RC 402 and 2 RC-404, and aork orders M2 90 09844, M2-89 05344 and M2 90-09843. Trie purpose of the review was to determine when the tagouts were accomplished and wh:n work was initiated on the piessurizer power operated relief valves. The valves uere released for maintenance on September 22, and October 13. The tagout review confirmed that appropriate remote work control tags had been placed, but no local tags had been placed on the motor operators for valves 2 RC 403 and 2 RC-405. On September 22,1990, at approximately 6:30 pm, the pressurizer manway was removed. This established a vent path during the time in which maintenance occurred and obviated the need to maintain boundary valve protection. Inspector review and discussions with health physics personnelindicated that - access to both the pressurizer block valves and power operated relief valves - require high radiation area access controls. Actual radiation levels at the motor operated bh>ck valves constitute a radiation area, but a worker had to traverse a hot spot field of about 8.0 rem / hour to gain access to the relief ultes. Inspector discussions with the job supervisor Indicated that he was aware of the tag sequence and that access to the four valves required high radiation controls. The job supervisor did r.ot feel that hanging a boundary tag on the associated block valves was iequired. h h
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l 14 On September 22, during release of work order M2 89 05344, a time existed during which the pressuriter manway was still installed and locally tagging a boundary valve would have been useful; however, the area was controlled as a high radiation area. Therefore the discretion exercised to not hang local tags was acceptable per the ACP-QA-2.06A. Conclujon ,, The inspector found that procedure ACP-QA-2.06A permits the exercise of discretion conceming hanging boundary tags in high radiation areas. The inspector concluded that the discretion exercised by the job supervisor was appropriate. No unsafe conditions were identified. 3.7 EcI)utr Attentiveness to Duty - Unit 2 The NRC resident inspector office inspected a concern that in two separate events licensee workers were reportedly found asleep while on duty. The first incident concerned a plant equipment operator (PEO) working in the Millstone 2 contaimnent on September 16, who allegedly was found asleep three times, and was aroused the last time by the operations supenisor. The second incident reportedly occurred around October 20 and involved a fire watch who was found asleep in the Millstone I cable vault. NRC followup of the events could not substantiate the fire watch concerns, and only partially substantiated.. the PEO concern as described below. 3J 1 Plant Ecuinment OIvarator Perfettpine valve Testine.-Efjt_2 -
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The inspector interviewed the Unit 2 operations supervisor, the Unit 2 plant equipment operator, and an operations person. All interviewees agreed upon the ongoing activities at the time; the date, location, and ? , individual involved. The activities involved containment penetration local leak rate testing. The time was between 7:00 - 8:00 pm on - September 16, and tne location of the work was the ground elevation inside containment. The Unit 2 operations supervisor observed thc.individua.1 during setup' activities for local leak rate testing on September 16. 'The supervisor did not observe the individual to be inattentive to duty; only that the. iodividual.was sitting down and leaning against some cloth material. The supervisor did not see any need in discipline the individual. However, he did inform the PEO's shift supervisor that the resting position he was in was not appropriate to the situation. The inspe: tar ' interviewed the plant equipment operator who stated that he was - attentive to' duty and recognized during activities that he should present . a more active position. V
17 The licensee has declared an Unusual Event on five occasions since January 1989i On three occasions in which ESF systems were declared inoperable, and - shutdown was initiated but not achieved, the licensee declared an unusual event and notified the NRC pursuant to 10 CFR 50.72. On one occasion involving c the feedwater codlant injection systern, the technical specification limiting condition for operation was not exceeded and shutdown was not required, In e this case, the licensen declared a " general interest even!' pursuant to its agreement with the state of Connecticut and notificd the NRC in accordance with 10 CFR 50.72. The inspector concluded that the licensee is classifying . events involving loss of an ESF function properly, and in accordance with NRC requirements. This item is closed. 4.0 Radialecical Controls 4.1 Postine and Control of RadiologicaLAms - All. Units During plant tours, posting of contaminated, high airbome radiation, and high radiation areas was reviewed with respect to boundary identification, locking requirements, and appropriate hold points. The inspector had no significant observations. 4.2 Radochemistr3Jfspline, Unit 2 On or about September 27,1990, an authorized work order (AWO) was initiated to allcw a vendor to pump out and clean a numb r of oli water separator sludge tanks (sewers). After pumpdown of the number 3 tank, a radiochemistry sample of the removed sludge was taken and some trace amounts of Cs 138 and Co-60 were identified. The contents were pumped to a 7, ' truck and the truck was decontaminated. The waste is in storage and will be processed as radioactive material. Plant personnel have initiated a plant incident report to investigate the source of the low level contamination and to ensure adequate controls are in place to prevent unmonitored releases from the oil water separctor tanks. The inspector had no further questions regarding this licensee activity. The inspector concluded that licensee actions were appropriate. A
61 evidence of cracked welds, missing bolts, loose fittings or damaged brackets. The lift pole and ICI plate threads were in good condition, with the exception that one-third turn on the lift pole starting thread was damaged.- L \ With the exception of C 16, no damage was observed on the ICis. Thimble C-16 had a 16-inch longitudinal split along one of the four fluted sections of the , detector sheath. The fluted section keeps the ICI centered within the instrument tube. The ICI detector will remain centered in the tube and its operation will not be affected by the damaged section. In the unlikely event that the thimble tube separated during the operating cycle, it would remain captured in the fuel assembly guide tube and would not become a loose part. If the damaged ICI failed to function, the remaining 44 detectors provide adequate margin to the minimum number required by TS 3.3.3.2 to support plant operation. Based on the engineering reviews and examinations, the licensee concluded that the ICI plate drop caused no damage that would affect adversely reactor safety or prevent continued reactor operation. [nmector Reviews and Conclusions The inspector reviewed the videotapes of the ICI and UGS structure, interviewed personnel involved with the examinations, and reviewed the engineering evaluations of the consequences of the drop. The inspector noted that the lift tool installation error occurred as a result of a combination of inadequacies in the associated procedure, familiarity of the personnel with the job, and supervision of the work, The event constituted a licensee failure to assure the satisfactory completion of a critical step in the refueling sequer.2. The inspector noted that the error is one of several , personnel performance issues that have occurred during the refueling outage. This NRC concern was addressed to the licensee for action and response in NRC inspection report 50-336/90-18, and will be followed as part of that inspection. The inspector concluded that licensee inspections, engineering and reportability evaluations, end conclusions were proper. The licensee's followup assessment of the event and its causes was extensive and thorough. Engineering support to evaluate the consequences of the event was good. 9.3 Len.sf Containment Intecrity Durine Fuel Movement Drgiption of Event On October 2, reactor refueling operations were in progress, with fuel
62 mowment ongoing in the containment and in the spent fuel pool. During refueling, containment integrity is established.to mitigate the potential consequences of a postulated accident involving the dropping of an irradiated fuel bund!e. To satisfy containment integrity requirements, the equipment hatch must be installed, at least one door of the personnel air lock must be \ closed, and penetrations either must be secured or capable of automatic isolation. rhe licensee had established containment integrity to satisfy the , requirements of technical specification 3.9.4 as a prerequisite for refueling. Plant operators were also preparing to drain steam generator #1 (SG#1). The operators were using Step 5.1.1 of OP 2316A, Main Steam System, to e,;tablish a drain vent path usir.g the atmospheric dump valves (ADV). The operator followed step 5.11.6.6 of the procedure to open the SG#1 dump valve. Opening the dump valve also required clearing of a safety tag. The SG#1 dump valve was tagged closed on 9/25/90 per clearance M2-2129 90 when the steam generator manway was opened to support steam generator maintenance activities. The tagging order stipulated that the atmospheric valve had to be kept closed (along with several other valves) at the direction of the shift supervisor for containment boundary protection. This control _ was reenforced by a caution in OP 2316A, which stated that the dump valve should not be opened while performing core alterations in order to assure that technical specification 3.9.4 requirements were met. The supervisory control room operator on duty on October 2 was aware that the secondary manway was open and of the operating procedure caution, but failed to recognize that clearing the tag to open the ADV was prohibited under existing p! ant conditions and would violate containment integrity.
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The dump valve was opened at 6:45 pm on October 2 to support the draining evolution. The vent path was opened for about I hour and 5 minutes, when, at7:50 pm, the duty outage coordinator, a shift supervisor, and a senior reactor operator (SRO), noted the open status of the ADV, The SRO immediately notified the shift personnel that containment integrity requirements were not satisfied. Refueling activities were suspended and, by 8:00 pm, the j ' ADV was closed, reestablishing containment integrity, l Fuel handling logs and records (ENG Form 21008-1, page 9 of 73) show that - l a single fuel bundle had been moved during the time when containment l integrity was compromised. Fuel bundle N-45 was inserted in' core location T-l ! 7 at 6: 42 pm. As the next move in sequence, fuel bundle K 25 was moved from core location T-9 and inserted in the north upender at 7:10 pm. No further fuel movement occurred from then until refueling activities were halted
63
.c at7:50 pm, as reactor engineering personnel investigated a problem with a hoist limit switch and proces:ed a temporary procedure change to OP 230312 to revise a bridge coordinate.
( The licensee initiated plant information repcrt 90109 to document the event and evaluate the incident. The event was reported to the NRC as required by ' 10 CFR 50.73 (a)(2)(i)(B) as licensee event report (LER) 9018 dated November 1,1990. Cause of Event i Licensee review attributed the cause of the event to personnel error. The open manways would have established an adequate vent path for the draining activities and obviated the need to open the dump valves, Inspector reviews-noted that the status of the steam generator manways was covered during shift turnover and briefings. Discussion with the operator indicated that he was - aware of the procedure and tagging requirements but failed to appreciate the consequence of opening the dump valve. The operator focused on the draining - evolution and failed to recognize that opening the ADV was prohibited under the existing plant conditions and would violate containment integrity, l L l l-L l l 1 I I
s . 4 - 64 e n 1.icensee Actions and Evaluations , Upon discovery of the violation, actions were taken'immediately to meet the , y requirements of TS 3.9.4. The licensee's assessment was that there was no actual impact on worker or public safety at the time since no radiological l, source term existed during _the 75 minute period in which containment integrity z, was compromised. In order to prevent necurrence of the event,'the followirig actions were taken:- ' c (i) the caution in OP 2316A on use of the ADVs was moved from step X to Y, 4 to place it closer to the instruction where the operator takes the action to open : the valves as part of the drain down evolutioni and, (ii) operations supervisors - were counseled regarding the need for greate'r attention to detail during the - ' performance of extensive maintenance work and changing plant conditions.
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The inspector reviewed the licensee's responses and determined that they ' 4 adequately addressed the root cause. The licensee's evaluation of the event was provided in LER 90-18. The inspector reviewed the evaluation with licensee personnel As no fuel handling accident occurred during the event, there were no actual technical consequences. The licensee completed an additional assessm_ent of the potential , consequences had a fuel drop accident occurred,- During the 75 minutes when containment integrity was lost, the actual fuel handling inside containment took place for 25 minutes and involved the movement of one fuel bundle from the ' core to an upender. The dump valve is an eight inch diameter, air operated. valve'(reference drawing 25203 26002). The licensee determined that the valve was manually _ _ opened two turns off its seat for the draining evolution,~ which was calculated . ' to be 1/2 inch of valve travel, and resulted in an opening of 0.087 square feet. ' Using offsite information system (OFIS) data to review containment pressure from 5:00 pm to 9:00 pm on October 2, the licensee noted that containment I pressure was positiv_e at about 2.0 inches of water, and, further, was constant during that period indicating that the open dump valve had no' apparent affect on the containment boundary. Nonetheless,- the licensee conservatively i ' assumed, for the purpose of the assessment, that the positive pressure would : have resulted in flow out of the containment during a postulated fuel handling M accident.' The calculated flow rate from the containment under the prevailing conditions would be 300 cubic feet per minute (cfm), o The licensee compared _the consequences of the postulated event under the above conditions with the FSAR analysis for a fuel handling accident. The' [ ' FSAR analysis assumes a fuel decay time of 72 hours, whereas the actual fuel - decay time on October 2 was 16 days, thus the potential source term is reduced - l M b e -vvr
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4 65 signi0cantly. Further, the FSAR analysis assumes that the containment purge valves are open initially and would remain open for 10 minutes during the event, which would result in a release to the enviromnent at a flow rate of 32,000 cfm. The calculated 300 cfm discharge rate would result in a e s signincantly reduced release rate. The licensee determined that the FSAR analysis remains bounding Pnd that an event under the conditions prevailing on October 2 would be much less signincant than that analyzed. The inspector , reviewed the licensee's calculations, analyses and assumptions. Insoeetor Reviews and Conclusinas The inspector noted that the personnel error by the operator is one of several personnel performance issues that have occurred during the refueling outage. This NRC concern was addressed to the licensee for action and response in NRC inspection report 50 336/90-18. Further a number of issues discusset! in this report further suggests a problem with attention to detail in carrying out of operating activities in accordance with regulatory requirements and licensee procedures. The failure to maintain containment integrity during fuel movement as required by TS 3.9.4 is an apparent violation of containment integrity technical speci5 cations (50 336/90-22-06), 9.4 Base Plate Anchor Boh Corrosian Procram to Evaluate Seismic Cateeory 1 Supoorts Initial NRC review of this issue was documented in Region I inspection report 50 336/90-82, Section 3.4.2, which considered the licensee's dispositioning of degraded anchor bolts on the "C" reactor building closed cooling water 1 (RBCCW) heat exchanger in 1989. . This item was reviewed during this - inspectica period to evaluate the actions taken since the 1989 outage and in progress caring the present outage to address the potential support degradations, During intervie vs with site engineering personnel, the inspector noted that the licensee had pre eiously identined the potential for anchor bolt corrosion and the need to address the concern generically, particularly in light of the experience with the RBCCW beat exchangers. The corrcsion mechanism and the location of bolt wastage resulted in significant loss of material with attendant loss of margin to the bolt design strength, with few obvious external indications of corrosion or degradation. Indirect evidence of underlying corrosion included cracked grout or rust _weepage on or around the support base plates.
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*"** fl0V - 01G00 Docket No. 50-336 t Mr. Edward Mroczka Senior Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company P.O. Box 270 Ilartford CT. 06141-0270 Dear Mr. Mroczka; The U.S. Nuclear Regulatory Commission recently received a number of allegations concerning activities at Millstone 2. Details of these issues are enclosed for your review and followup. We request that die results of your review and disposition of these matters be submitted to Region I within 30 days of receipt of this correspondence. We request that your response contain no personal privacy, proprietary, or safeguards information so that it can be released to the public and placed in the NRC Public Document Room. If r.ecessary, such information to be withheld shall be contained in a separate correspondence and the affidavit required by 10CFR 2.790 must accompany your response if proprietary or like information is included.
The response requested by this letter and the accompanying enclosures are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. Your cooperation with us is appreciated. Please address any questions that you may have regarding these issues to Mr. Scott Stewar at (215) 337-5232, or Mr. Donald Haverkamp at (215) 337-5120, Sincerely, bl eas Edward C. Wenzinger, Chief - Reactor Projects Branch 4 Enclosure 1, Allegction RI-90-A-0180, Enclosure 2, Allegation RI-90-A-0202. cc w/ encl: W. Raymond, SRI f , f/ W . Wo3
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- 1. a. Wide range nuclear instruments were not operable on October 9,1990 as required to support refueling operations because; i) "A" channel spikes periodically. This is a-long standing problem that has not been resolved, the I&C technicians have " banged on" the channel to stop the spiking.
- 11) "C" channel cable has been-damaged and.this damage has ,
affected readings on the channel. The channel has
" low TR" readings on the cable, i ii) a PDCR to change out the channels continues to be open and until closed and signed off, the channels cannot be operable,
- b. the I&C technicians have been under pressure'to allow the (above) discrepant conditions to continue to exist with the channel considered operable to allow fuel alterations to occur.
Please in your discussion of the above issues, provide any Plant Operations Review Committee determinations concerning operability of WRNIs for core alterations. Y V
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- 2. The " owner" of Wide Range Nuchar Inst;ument procedure, SP-2417H, was not consulted for a recent procedure change processed to support outage activities. This is contrary to I&C department policy, i
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[_t . (- RI-90-A-0202 Enclosure 2 Issue 1 Authorized Work Order M2-90-00579 is a one page AWO for annual preventative ' maintenance (PM) on various Limitorque operators. A note on the AWO says that the performance of the PM will not affect EEQ boundaries. However, ACP 2.16, page 21, Item 0 states that all maintenance work or EEQ examinations be. documented on 3 page AW0s.
- 1. Was the one-page AWO appropriate for this maintenance ' item? Were there proper EQ reviews?
- 2. Were single page AW0s appropriate in the past to ensure EQ requirements were satisfied? (If a review of single page AW0s is conducted, please discuss the sample size and effort to-ensure that the sample is representative).
- 3. ACP-QA-2.16 w'as revised on September 11, 1990 to require that maintenance on EEQ equipment be documented on 3 page AW0s, (Reference MM-90-214, dated Novemoer 6, 2990) Why was this revision required?
- 4. Are motor operated valve cover gaskets replaced or are torque switch settings changed using single page AW0s? If so, is this satisfactory to ensure EEQ requirements?
Issue 2 Authorized Work Order AWO-H2-90-12648 required electrical retermination of valve 2-MS-1908. t-l 1. How and for how long was the termination that needed to be redone ! incorrect? j 2. Was there a-safety impact'due to the original deficient termination? l 3. What were the circumstances that caused the A40 to be prepared? (i.e. l How was the deficiency discovered? What was the cause'of the deficiency?)- l .. l 4. Was there a QC hold point'or similar review that should have prevented the deficiency in' the original termination? 1. I o
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Subject:
A'l on Concerning Hillstone Unit 2 , The NRC Region I office has completed its followup in response to the concerns you brought to our attention on November 1,1989. You expressed dissatisfaction with the NRC followup .and disposition of allegations, specifically the NRC i inspection teams disposition of your concerns relating to the SP-EE-076 Electric ' Code and the use of electrical metallic tubing (EMT), as documented in * - inspection report 50-336/89-13. The NRC has subsequently reviewed your concerns again and concluded that the evaluation and disposition of your concerns in inspection report 50-336/89-13 was correct. The inspection team determined that $P-EE-076, as a specification, does not apply to Unit 2. It is a specification developed by Stone and Webster and used in the construction of Millstone Unit 3, the plant for which Stone and Webster was the architect and engineer. $P-EE-076 has been distributed to the other-nuclear generating stations operated by the licensee as a reference and general guidance document. The inspection team reviewed licensee internal memoranda that promulgated the use of SP-EE-076 at these plants, and no discrepancies , I were noted. l The inspection team also concluded that your allegation that EMT is used in l applications other than lighting installations at Millstone Unit 2 is I substantiatwd, These applicatier.;, however, are clearly allowed for and f controlled by the applicable specification, Bechtel drawing 25203-34001, and, I therefore, pose no safety hazard to the plant. I l We appreciate your informing us of your concerns and feel that our actions in ' l this matter have been responsive to those concerns. Should you have any additional questions, or if I can be of further assistance in this matter, please call me collect at (215) 337-5120. Sincerely, in, fit M Donald R. Have kamp, hief
..-Allegation File RI-88-0003 Reactor Projects Section 4A It. Raymond l
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*** KING oF PRusslA. PENNSYLVANIA 19406 Docket No. 50 245; 50 336; 50-423 ,,
License No. DPR 21; DPR 65; NPF-49 \ EA No. 90 219 Nonheast Nuclear Energy Company ATTN: hir. E. J. Mroczka Senior Vice President - Nuclear Engineering and Operations P.O. Box 270 Hartford, Connecticut 06141-0270 Gentlemen:
Subject:
NRC Region I Combined Inspection Nos. 50 245/90 20, 50-336/90 22, and 50-423/90-20 This letter transmits the NRC repon of our routine safety inspection that was conducted by Messrs. D. Dempsey. P. Habighorst, and K. Kolaczyk of this office on September 18 - November 15, October 2 - November 15 (and continued December 3-13 to evaluate funhet one of the significant issues described herein), and October 16 - November 15,1990. for Millstone Units 1,2. and 3. respectively. At the conclusion of the inspection the findings were discussed by the above inspectors with Mr. S. E. Scace and other members of your staff. Areas examined during the inspection are described in the NRC Region I inspection report. l which is enclose,d with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, and document reviews. : l
- i. Based on the results of this inspection, two apparent violations were identified at Millstone
! Unit 2 and are being considered for escalated enforcement action in accordance with the
" General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Polic,0,10 CFR Part 2, Appendix C (1990). The apparent violations involve the loss of conta5 ment integrity control as a result of personal errors and are discussed in the Unit 2 l operations and outage sections of the enclosed report, Specifically, the apparent violations l involve the loss of containment integrity due to the inoperability of the containment purge valve isolation system, and in a separate event, the loss of containment integrity via the No. I steam generator atmospheric dump valve. Accordingly, no Notice of Violation is presently being issued for these inspection findings. Please be advised that the number and characterization of apparent violations described in the enclosed inspection report may change as a result of funher NRC review.
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. DEC 2 61990 1 Northeast Nuclear Energy Company. !
An enforcement conference to discuss these apparent violations at Millstone Unit 2 has been l scheduled for January 15, 1991. The purposes of this conference are to discuss the apparent - violations, their causes and safety significance; to provide you the opportunity to point out any errors in our inspection report; to provide an opportunity for you to present your U
.l proposed corrective actions; and to discuss any other information that will help us determine the appropriate enforcement action in accordance with the Enforcement Policy. You will be advised by separate correspondence of the results of our deliberations on this matter. No '
response regarding these apparent violations is required at this time. The enclosed report addresses your performance during the recent refueling and maintenance outage on Millstone 2. Overall, we found the control of outage activities to be good, with effective management of planned activities and aggressive followup of problems. De thorough evaluation of unplanned events, the extensive support by corporate engineering and - i vendors to disposition of these issues, and the effective interface between site and corporate engineering were notable strengths. Your assessment of the personnel performance aspects of ' l these events was requested in our letter to you dated November 5,1990, enclosing Inspection Report 50436/9018. j Notwithstanding the above conclusion regarding generally good performance, we noted a number of events attributed to personnel error, that apparently resulted from procedure quality and adherence weaknesses. Further, the failure to satisfactorily complete a critical step during the replacement of in-core instruments that resulted in the dropping of the incore instrument support plate was significant (see section 9.2 of the enclosed repon). Car assessment was that the lift tool installation error resulted from a combination of inadequacies in procedure details, personnel experience, and supervision of the work activity. The event demonstrates the need for greater diligence in the review process for "tried and proven" procedures to eliminate any over reliance on personnel experience for critical activities. J Based on the results of this inspection at Millstone Unit 1, certain of your. activities appeared l to be in violation of NRC requirements, as specified in the Notice of Violation enclosed i
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herewith as Appendix' A. We are concerned about the violation because it involved the_ operation of Millstone I with non conservative setpoints on the steam jet air ejector radiation monitor. You are required to respond to this violation and should follow the instructions- l speci0ed in Appendix A when preparing your response. In your response, you should i document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to Appendix A, including your proposed corrective actions and i the results of future inspections, the NRC will deterinine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements. l In addition, certain of your activities at Millstone Unit 2 appeared to be in deviation from your _ written commitments, as specified in the Notice of Deviation enclosed herewith as Appendix B. We are concerned about the deviations because they involved the failures to tes 2
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l \' Nonheast Nucien Energy Company N reactor protection channels and to operate the loose parts monitor in accordance with Final Safety Analysis Report commitments. You are requested to respond to these deviations, and should follow the instructions speci6ed in Appendix B in preparing your response, In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room, f The responses directed by this letter and the enclosed Notices are not subject to the clearance procedures of the Of6ce of hianagement and Budget as required by the Paperwork Reductio Act of 1980, Pub. L. No. 96.511. Your cooperation with us is appreciated. Sincerely, G jj s
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, les V. Hehl' }rrector ,
) ision of Reach, r'?rojects
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. Appendix B, Notice of Deviation
- 3. NRC Region 1 Combined inspection Report No. 50-245/90-20; 50 036/90-22; 50-423/90-20 cc wiencis:
W. D. Romber;;. Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Services
- R. bl. Kacich, hianager, Generation Facilities Licensing S. E. Scace, Station Director, hiillstone H. F. Haynes, Nucler Unit Director, hiillstone Unit 1 J. S. Keenan, Nuclear Unit Director, hiillstone Unit 2 C. H. Clement, Nuclear Unit Director, hiillstone Unit 3 Gera'd GarGeld, Esquire Public Document Room (PDR)
Local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) NRC Senior Resident inspector State of Connecticut 3 l i
P U.S. NUCLEAR REGULATORY COMMISSION
' REGION 1 Repon No.: 50-245/90 20; 50-336/90-22; 50-423/90-20 s
Docket Noa 50-245; 50 336; 50 423 , License No.: DPR 21; DPR-65; NPF-49 Licensee: Nonheast Nuclear Energy Company P. O. Box 270 Hanford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Units 1,2, .3 Inspection At: Waterford, CT Dates: September 18 November 15,1990 (Millstone 1) October 2 - November 15, and December 3 - 13,1990 (Millstone 2) October 16 - November 15,1990 (Millstone 3) Reporting Inspectors: D. A. Dempsey, Resident Inspector, Unit 1 P. J. Habighorst, Resident Inspector, Unit 2 K. S. Kolaczyk, Resident inspector, Unit 3 Inspectors: W. J Raymond, Senior Resident Inspector D. A. Dempsey, Resident Inspector, Unit 1 P. J. Habighorst, Resident Inspector, Unit 2 K. S. Kolaczyk, Resident Inspector, Unit 3 J. S. Stewart, Senior Project Engineer . A. Vegel, Reactor Engineer Approved by: ek me.c -- /2/rJ/90 Date Donald R. Haverkamp, Chief f Reactor Projects Section 4A Division of Reactor Projects L In1,. ection Summary: Report 50-245/90-20; Repon 50-336/90-22; Report 50-423/90 20 Areat Inspected: Routine NRC resident inspection of plant operations, radiological controls, maintenance, surveillance, security, outage activities, licensee self-assessment, and periodic ( reports. l Results: See Executive Summary i A lol0 M W,.
13 3.6 Improper Tareine Concern - Unit 2. On October 19, the inspector reviewed a concern regarding improper equipment isolation controls. Specifically, no local tags were hung on motor. c operated valves 2 RC-403 and 2 RC-405 (pressurizer relief isolation) duiing - maintenance work. The valves are consioered to be boundary isolation valves,- as defined in procedure ACP-QA 2.06A, section 6.1.8 and, thus should have ,4 been tagged locally. No local tags were hung. The inspector noted that ACP-QA-2.06A, section 6.1.8, requires that, in addition to normal equipment tagging, local operators of motor and pneumatically operated valves be tagged when the valve is used as a system isolation boundary point. If the local operator is in a high radiation area, placement of safety tags is left to the discretion of the shift supervisor, senior control operator, or job supervisor. The inspector reviewed the tagouts for valves 2 RC-402 and 2 RC-404, and work orders M2 90 09844, M2-89-05344 and M2 90-09843. The purpose of the review was to determine when the tagouts were accomplished and when work was initiated on the pressurizer power operated relief valves. The valves were released for maintenance on September 22, and October 13. The tagout review confirmed that appropriate remote work control tags had been placed, but no local tags had been placed on the motor operators for valves 2-RC 403 ' and 2 RC-405. On September 22,1990, at approximately 6:30 pm, the pressurizer manway was remosed. This established a vent path during the time in which maintenance occurred and obviated the need to maintain boundary valve protection. , inspector review and discussions with health physics personnel indicated that access to both the pressurizer block valves and power-operated relief valves require high radiation area access controls. Actual radiation levels at the motor operated block valves constitute a radiation area, but a worker had to : traverse a hot spot field of about 8.0 rem / hour to gain access to the relief - valves. Inspector discussions with the job supervisor indicated that he was aware of the tag sequence and that access to the four valves required high radiation controls. The job supersisor did not feel that hanging a boundary tag on the associated block valves was required. l l l l
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-14
-On September 22, during release of work order M2-89 05344, a time existed -
during which the pressurizer manway was still installed and locally tagging a boundary valve would have been useful; however, the area was controlled as a high radiation area. Therefore the discretion exercised to not hang local tags c was acceptable per the ACP-QA 2.06A. Conclusion , The inspector found that procedure ACP-QA-2.06A permits the exercise of discretion concerning hanging boundary tags in high radiation areas. The inspector concluded that the discretion exercised by the job supervisor was appropriate. No unsafe conditions were identified. 3.7 Worker Attentiveness to Duty - Unit 2 The NRC resident inspector office inspected a concern that in two separate events licensee workers were reportedly found asleep while on duty. The first incident concerned a plant equipment operator (PEO) working in the Millstone 2 containment on September 16, who allegedly was found asleep three times, and was aroused the last time by the operations supervisor. The second incident reportedly occurred around October 20 and involved a fire watch who was found asleep in the Millstone I cable vault. NRC followup of the events could not substantiate the fire watch concerns, and only partially substantiated the PEO concern as described below. 3.7.1 Elant Equipment Onerator Performine valve Testine - Unit 2 4 The inspector interviewed the Unit 2 operations supervisor, the Unit 2 plant equipment operator, and an operations person. All interviewees - ' agreed upon the ongoing activities at the time; the date, location, and individual involved. The activities involved containment penetration local leak rate testing. The time was between 7:00 8:00 pm on September 16, and the location of the work was the ground elevation inside containment. The Unit 2 operations supervisor observed the individual during setup activities for local leak rate testing on September 16. The supervisor did not observe the individual to be inattentive to duty; only that the individual was sitting down and leaning against some cloth material. The supervisor did not see any need to discipline the individual. However, he did inform the PEO's shift supervisor that the resting position he was in was not appropriate to the situation. The inspector interviewed the plant equipment operator who stated that he was attentive to duty and recognized during activities that he should present a more active position.
i 17 The licensee has declared an Unusual Event on five occasions since January 1989. On three occasions in which E.SF systems were declared inoperable, and shutdown was initiated but not achieved, the ll:ensee declared an unusual event and notified the NRC pursuant to 10 CFR 50.72. On one occasion involving the feedwater coolant injection system, the technical specification limiting condition for operation was not exceeded and shutdown was not required, in ' this case, the licensee declared a " general interest event
- pursuant to its ,
agreement with the state of Connecticut and notified the NRC in accordance with 10 CFR $0.72. The inspector concluded that the licensee is classifying , events involving loss of an ESF function properly, and in accordance with NRC requirements. This item is closed. 4.0 Radioloricil Controls 4,1 Postinc and Control of Radiclocical Areas - All Units During plant tours, posting of contaminated, high airborne radiation, and hlgh radiation areas was reviewed with respect to boundary identification, locking requirements, and appropriate hold points. The ;nspector had no significant observations. l 42 Badiochemistry SamNine Unit _2 On or about September 27,1990, an authorized work order (AWO) was initiated to allow a vendor to pump out and clean a number of oil water separator sludge tanks (sewers). After pumpdown of the number 3 tank, a radiochemistry sample of the removed sludge was taken and some trace arnounts of Cs 138 and Co 60 were identifisd. The contents were pumped to a truck and the truck was decontaminated. The waste is in storage and will be i processed as radioactive material. Plant personnel have initiated a plant incident report to investigate the source of the low level contamination and to ensure adequate controls are in place to prevent unmonitored releases from the oil water separator tanks. I, i The inspector had no further questions regarding this licensee activity. The t inspector concluded that licensee actions were appropriate. L L i-h l M g y % erg-w. - y- ur- uw'-ve --g
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0 i 61 evidence of cracked welds, missing bolts, loose fittings or damaged brackets. The lift pole and ICI plate threads were in good condition, with the exception that one third turn on the lift pole starting thread was damaged. , t with the exception of C 16, no damage was observed on the ICis. Thimble C-16 had a 16 inch longitudinal split along one of the four fluted sections of the detector sheath. The fluted section keeps the ICI centered within the , instrument tube. The ICI detector will remain centered in the tube and its operation will not be affected by the damaged section. In the utslikdy event that the thimble tube separated during the operating cyr.le, it would remain captured in the fuel assembly guide tube and would not become a loose part. If the damaged ICI failed to function, the remaltdng 44 detectors provide adequate margin to tht, minimum number required by TS 3.3.3.2 to support plant operation. Based on the engineering reviews and examinations, the licensee concluded that the ICI plate drop caused no damage that would affect adversely reactor safety or prevent continued reactor operation. Inmmor Reviews and Conclusions The inspector reviewed the videotapes of the ICI and UGS structure,
. interviewed personnel involved with the examinations, and reviewed the engineering evaluations of the consequences of the drop, The inspector noted that the lift tool installation error occurred as a result of a '
combination of inadequacies in the associated procedure, familiarity of the personnel with the job, and supervision of the work. The event constituted a licensee failure to assure tne satisf actory completion of a critical step in the refueling sequence. The inspector noted that the error is one of several , personnel performance issues that have occurred during the refueling outage. ' This NRC concern was addressed to the licensee for action and response in NRC inspection report 50 336/90-18, and will be followed as part of that inspection. Tne inspector concluded that licensee inspections, engineering and reportability evaluations, and conclusions were proper. The licensee's follownp assessment of the event and its causes was extensive and thorough. Engineering support to evaluate the consequences of the event was good. 9.3 1. css of containmentJnitgIity During Fuel Movement Description of Even! On October 2, reactor refueling operations were in progress, with fuel i L L I w w ---g -+W g e m g e n+g v '-e-"o y
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62 g i mosement ongoing in the containment and in the spent fuel pool. During refueling, containment integrity is established to mitigate the potential consequences of a postulated accident involving the dropping of an irradiated fuel bundle. To satisfy containment integrity requirements, the equipment hatch rnust be installed, at least one door of the personnel air lock must be -\ closed, and penetrations either must be secured or capable of automatic isolation. The licensee had established containment integrity to satisfy the ', requirements of technical specification 3.9.4 as a prerequisite for refueling. Plant operators were also prepa ing to drain steam gen ' ator #1 (SG#1). The operators were using Step 5.1.1 of OP 2316A, Main Steam System, to establish a drain vent path using the atmospheric dump valves (ADV). De operator followed step 5.11.6.6 of the procedure to open the SGF1 dump valve. Opening the dump valve also required clearing of a safety tag. The SO#1 ' dump valve was tagged closed on 9/25/90 per clearance M2 2129-90 when the steam generator manway was opened to suppon steam generator maintenance activities. The tagging orc'er stipulated that the atmospheric valve had to be kept closed (along with several other valves) at the direction of the shift supervisor for containment boundary protection. This control was teenforced by a caution in OP 2316A, which stated that the dump valve shou'd not be opened while performing core alterations in order to assure that technical specification 3.9.4 requirements were met. The supervisory control room operator on duty on October 2 was aware that-the secondary manway was open and of the operating procedure caution, but failed to recognize that clearing the tag to open the ADV was prohibited under existing plant conditions and would violate containment integrity, f The dump valve was opened at 6:45 pm on October 2 to suppon th driining evolution. The vent path was opened for about I bour and 5 minutes, when, at 7:50 pm, the duty outage coordinator, a shift supervisor, and a senior reactor operator (SRO), noted the open status of the ADV. The SRO immediately notified the shift personnel that containment integrity requirements were not satisfied. Refueling activities were suspended and, by 8:00 pm, the ADV was closed, reestablishing containment integrity. Fuel handling logs and records (ENO Form 21008-1, page 9 of 73) show that a single fuel bundl: had been moved during the time when containment integrity was compromised. Fuel bundle N-45 was inserted in core location T-7 at 6:42 pm. As the next move in sequence, fuel bundle K 25 was moved from core location T 9 and insened in the nonh upender at 7:10 pm. ' No further fuel movement occurred from then until refueling activitias were halted i. L 1
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63 1 at 7:50 pm, as reactor engineering personnel investigated a problem with a hoist limit switch and processed a temporary procedure change to OP 230312 l to revise a bridge coordinate. . t The licensee initiated plant information report 90109 to document the event and evaluate the incident. The event was reported to the NRC as required by ! 10 CFR 50.73 (a)(2)(i)(B) as licensee event report (LER) 9018 dated November 1,1990. 1 Cause of Event Licensee review attributed the cause of the event to personnel error. De open manways would have established an adequate vent path for the draining activities and obviated the need to open the dump valves. Inspector reviews noted that the status of the steam generator manways was covered during shift turnoser and briefings. Discussion with the operator indicated that he was aware of the procedure and tagging requirements but failed to appreciate the consequence of opening the dump valve. The operator focused on the draining evolution and failed to recognize that opening the ADV was prohibited under , the existing plant conditions and would violate containment integrity. l i. l l l
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E 64 1, ken;tAnig,nund I' valuations Upon discovery of the violation, actions were taken imrnediately to meet the requirements of TS 3.9.4. The licensee's assessment was that there was no g actus) impact on worker or public safety at the time since no radiological source term existed during the 75 minute period in which containment integrity was compromised. , In order to prevent recurrence of the event, the following actions were taken: (i) the caution in Op 2316A on use of the ADVs was moved from step X to Y, to place it closer to the in:nction where the operator tdes the action to open the valves as part of the drain down evolution; and, (ii) operations st pervisors were counseled regarding the need for greater attention to detail during the i performance of extensive maintenance work and changing plant conditions. The inspector reviewed the licensee's responses and determined that they adequately addressed the root cause. The licensee's evaluation of the event was provided in LER 90-18. The inspector reviewed the evaluation with licensee personnel. As no fuel handling accident occurred during the event, there were no a tual technical consequences. The licensec completed an additional assessment of the potential consequences had a fuel drop accident occurred. During the 75 minutes when containment integrity was lost, the actual fuel handling inside: containment took place for 25 minutes and involved the movement of one fuel bundle from the core to an upender. The dump valve is an eight inch diameter, air operated valve (reference drawing 25203-26002). The licensee determined that the valve was manually opened two tums off its teat for the draining evolution, which was calculated to be 1/2 inch of valve travel, and resulted in an opening of 0.087 square feet. Using offsite information system (OFlS) data to review containment pressute from 5:00 om to 9:00 pm on October 2, the licensee noted that containment pressure was positive at about 2.0 inches of water, and, further, was constant i during that period indicating that the open duntp valve had no apparent affect - ' on the containment boundary. Nonetheless, the licensee conservatively assumed, for the purpose of the ar.sessment, that the pontive pressure would have resulted in flow out of the containment duricg a postulated fuel handling l-accident. The calculated flow rate from the containment under the prevailing conditions would be 300 cubic feet per minute (cfm). L l The licensee compared the consequences of the postulated event under die above conditions with the FSAR analysis for a fuel handling accident. De FSAR analysis assumes a fuel decay time of 72 hours, whereas the actual fuel decay time on October 2 was 16 days, thus the potential . source term is reduced
.__ __ . _ . _ . . _. _ .- _ _ _ _ ___ -. _ .___~ _
0 65 e signi6cantly. Further, the FSAR analysis assumes that the containment purge , vahes are open initially and would rernain open for 10 minutes during the event, which would result in a release to the environment at a flow rate of 'i 32,000 cfm. The calculated 300 cfm discharge rate would result in a ' signincantly reduced release rate. The licensee deterinined that the FSAR analysis remains bounding and that an event under the conditions prevailing on October 2 would be much less significant than that analyzed. The inspector - , reviewed the licensee's calcelations, analyses and assumptions. Inim.uct.Roswr and Cmlc [g11coj The' 43. ncN ,%t the personnel error by the operator is one of several
<c p v that have occurred during the refueling outage.
pera . 6 % , This bic o.at : @ L ..ssed to the licensee for action and response in NRC in., :M "v-m k336/9018. Funher a number of issues discussed in this report tu, at suggests a problem with attention to detail in carrying out of
'sperating annities in eccordance with regulatory requirements and licensee procedures.
The failure to maintain containment integrity during fuel movement as required by TS 3.9.4 is an apparent violation of containment integrity technical specifications (50 336/90 22 06). 9A Dase Plate AnchcijolLCorrosion Program to Evaluate Seismic Category I S m on) . Initial NRC review of this issue was documented in Region I inspection report 50 336/90 82, Section 3.4.2, which considered the licent,ee's dispositioning of degraded anchor bohs on theJC" reactor building closed cooling water _ (RBCCW) beat exchanger in 1989. This item was reviewed during this inspection period to evaluate the actions taken since the 1989 outage and in prorgess during the present outage to address the potential st.pport degradations. During interviews with site engineering personnel, the inspector noted that the licensee had previously identified the potential for anchor bolt corrosion and the need to address the concern generically, particularly in light of the experience with the RBCCW heat exchangers. The corrosion mechanism and the location of bolt wastage resuhed in significant loss of material with attendant loss of margin to the bolt design stre.ngth, with few obvious external indications of corrosion or degradation. Indirect evidence of underlying corrosion included cracked grout or rust weepage on or around the support base plates.
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Do:ket No. 50 336 Th rA* "M , ' 7d This refers to your memorandums recently received by out office in which you expressed ! concerns related to the operation of the Millstone facility by your employer, Northeast Nuclear Energy Company, We have initiated acdons to examine your concerns and upon completion of our actions, will inform you of our findings. We must inform you that confidentiality is no.t granted as a routine matter and thetcfore your requests for confidentiality are denied. I assure you that we will attempt to conceal your identit ' while resolving these matters but licensees can and sometimes do surmise the idendty of individuals who provide information to us because of the nature of the information or other factors beyond our control. In such cases, our policy is to neither confirm not deny the licensee's assumption. Should you have any additional questions, or if I ca.n be of any further assistance, please call m collect at (215) 337-5120. *
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<s u Donald Haverkamp Chief Division of Reactor Projects, Section 4B 4
l cc: W. Raymond, SRI Information in this record was deleted Allegation File in accordwce with th Fp cfInformation i t Act, ex9ptipqs eb l ' F0IA S -Hol .,,- ggt &G, /['[- Nh
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go 2 8 690 : Docket No. 50 245; 50 336; 50-423 3. License No. DPR 21; DPR 65; NPF 49 i - EA No. 90 219 Nonhr.ast Nuclear Energy Company ; i ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear -; Engineering and Operations , P.O.Ika 270 Hartford, Connecticut 06141 0270 Gentlemen:
Subject:
NRC Region .I Combined Inspection Nos. 50-245/90 20, 50 336/90 22, and 50 423/90 20 This letter transmits the NRC repon of our routine safety inspection that was conducted by Messrs. D. Dempsey P. Habighorst, and K. Kolaczyk of this office on September 18 - - November 15, October 2 November 15 (and continued December 313 to evaluate further - one of the significant issues described herein), and October 16 November 15,1990. for Millstone Units 1,2, and 3. respectively. At the conclusion of the inspection the findings: , were discussed by the above inspectors with Mr. S. F, Scace and other members of your-staff. . Areas examined during the inspection are described in the NRC Region 1 inspection feport. which is enclosed with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, and document reviews. . E Based on the results of this inspection, two apparent violations were identified at Millstone. Unit 2 and are being considered ' aalated enforcement action in accordance with the -
" General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy),10 CFR Pan 2, Appendix C (f 990). The apparent violations involve the loss of conta9 ment integrity control as a result of personal errors and are discussed in the Unit 2-opera. ions and outage sections of the enclosed report, Specifically, the apparent violations involve the loss of containment integrity due to the inoperability of the containment purge valve isolation system, and in a separate event, the loss of containment integrity via the No.1: -
steam generator atmospheric dump valve. Accordingly, no Notice of Violation is presently? , being issued for these inspection findings. Please be advised that the number and characterization of apparen_t violations described in the enclosed inspection repott may change ' as a result of funher NRC review. , l . JNN090tof T y w r-- Ty*a 1y-*- tQ%'8mp 'a- -g an. W e'ra1r-*h++- +E-W=, '+'1W **1':- T W*9e<*?pt*+"**f8W-*W6m*f'*-- F M1C TW'F-' .F 'T 'pf4 y* W'9vuw'f-t**P'W7 T *- g- W FT"'T'+wt'w't"
E26E Sortheast Nuclear Energy Company An enforcement conferente to discuss these apparent violations at blillstone Unit 2 has been scheduled fer January 15, 1991. The purposes of this conference are to discuss the appaJent violations, their causes and safety significance; to provide you the opportunity to point out any enors in our inspection report; to provide an opportunity for you to present your t proposed conectne actions; and to discuss any other information that will help us determine ' You will be the appropnate enforcement action in accordance with the Enforcement Policy advised by separate conespondence of the results of our deliberations on this matter. No , response regarding these apparent violations is required at this time, The enclosed report addresses y our performance during the recent refueling and rnalntenance outage on hiillstone 2. Overall, we found the control of outage activities to be good, with effective management of planned activities and aggressive followup of problems. The thorough evaluation of unplanned events, the extensive support by corporate engineering and vendors to disposition of these issues, and the effective interface between site and corporate engineering were notable strengths. Your assessment of the personnel performance aspects of these events was requested in our letter to you dated November 5,1990, enclosing Inspection Report 50 336/90-18. Notwithstanding the above conclusion regarding generajly good performance, we noted a number of esents attnbuted to personnel enor, that apparently resulted from procedure quality and adherence wec.knesses. Further, the failure to satisfacionly comp!cte a cdtical step during the replacement of in-core instnaments that resulted in the dropping of the incore instrument support plate was significant (see section 9.2 of the enclosed report). Our assessment was that the lift tool installation enor resulted (tom a combination of inadequacies in procedure detaus, personnel er.perience, and supervision of the work activity. The event demonstrates the need for greater diligence in the review process for tried and proven" procedures to eliminate any ovcr reliance on personnel experience for critical activities. Based on the results of this inspection at hiillstone Unit 1, certain of your activities appeared to be in violaion of NRC requirements, as specified in tbc Notice of Violation enclosed , herewith as Appendix A. We are concerned about the violation because it involved the operation of Millstone 1 with non conservative setpoints on the steam jet air ejector radiation monitor. You are required to respond to this violation and should follow the instructions specified in Appendix A when preparing your response. In your response, you should dc<.ument the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to Appendix A, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory tequirements. in addition, certain of your activities M Millstone Unit 2 appeared to be in deviation from ' your written commitments, as specified in the Notice of Deviation enclosed herewith as Appendix B We are concerned about the deviations because they involved the failures to test 2 i
Nonheast Nuclear Energy Company N reactor protection channels and to operate the loose pans monitor in accordance with Final Safety Analysis Repon committnents. You are requested to respond to these deviations, and should follow the instructions specified in Appendix B in preparing your response. , in a:cordance with 10 CFR 2.790 of the NRC's ' Rules of Practice," a copy of this letter and '. Its enclosures will be placed in the NRC Public Document Room. The responses directed by this letter and the enclosed Notices are not subject to the clearance procedures of the Of0cc of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No. 96.511. Your coop: ration with us is appreciated. Sincerely,
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. . l'e V. Hehl' } rector ,
ision of Reach, r S tojects
Enclosures:
- 1. Appendix A Notice of Violation
- 2. Appendix B, Notice of Deviation
- 3. NRC Region 1 Combined Inspection Repon No. 50 245/90 20; 50 336/90 2h 50-423/90 20 f ec w!enels'.
W. D. Rornberg. Vice President, Nuclear Operadon.s D. O. Nordquist, Director of Quality Services ; R. M. Kacich, Manager, Generation Facilities Licensing S. E. Scace, Station Director, Millstone H. F. Haynes. Nu: lear Unit Director, Millstone Unit 1
- 1. S. Keenan, Nuclear Unit Director, Millstone Unit 2 C. H. Clement, Nuclear Unit Director, Millstone Unit 3 Gerald Garfield, Esquire Public Document Room (PDR) .
Lccal Public Document Room (LPDR) L Nuclear Safety Information Center (NSIC) NRC Senior Resident inspector State of Connecticut 3 l l l
U.S. NUCLE /sR REGULATORY COhihilSSION REGION 1 Report No.: 50 245/90 20; 50 336/90 22; 50-423/90 20 t ! Docket No.: 50-245; 50-336; 50 423 . License No.: DPR 21; DPR 65; NPF 49 , Licensee: Northeast Nuclear Energy Company , P. O. Box 270 Hanford, CT 06141-0270 Facility Name: hiillstone Nuclear Power Station, Units 1,2, & 3 inspection At: Waterford, CT Dates: September 18 November 15,1990 (hfilistone 1) October 2 November 15, and December 3 13,1990 (htillstone 2) October 16 - November 15,1990 (hiillstone 3) Reporting Inspe: tors: D. A. Dempsey, Resident inspector, Unit 1 P. J. Habighorst, Resident inspector, Unit 2 K. S. Kolaczyk, Resident inspector, Unit 3 Inspectors: W. J. Raymond, Senior Resident inspector D. A. Dempsey, Resident inspector, Unit 1 P. J. Habighorst, Resident inspector, Unit 2 K. S. Kelaczyk, Resident inspector, Unit 3 J. S. Stewan, Senior Project Engineer A. y\egel, Reactor Engineer > Approved by: ,_
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/Zdf[90 Donald R. Haverkamp, Chief [ Date Reactor Projects Section 4A Division of Reactor Projects in3,eection Summary: Report 50 245/90 20; Report 50-336/90-12; Report 50-423/90 20 3_ ten.laspected; Routine NRC resident inspection of plant operations, radiological controls, mamtenance, sutveillance, security, outage activities, licenses teif assessment, and periodic reports.
ECRkn See Executive Summary i
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13 3.6 Improper Taccine concem Unit 2 On October 19, the inspector reviewed a concern regarding improper equipment isolation controls. Speci0cally, no local tags were hung on motor- c operated valves 2 RC 403 and 2 RC-405 (pressurizer relief isolation) during maintenance work. The valves are considered to be boundary isolation valves, as denned in procedure ACP-QA 2.06A, section 6.1.8 and, thus should have < been tagged locally. No local tags were hung. The inspector noted that ACP-QA 2.06A, section 6.1.8, requires that, in addition to normal equipment tagging, local operators of motor and pneumatically operated valves be tagged when the valve is used as a system isolation boundary point. If the local operator is in a high radiation area, placement of safety tags is left to the discretion of the shift supervisor, senior control operator, or job supervisor. The inspector reviewed the tagouts for valves 2 RC 402 and 2 RC-404, and work orders M2 90 09844, M2 89 05344 and M2 90 09843. The purpose of the review was to determine when the tagouts were accomplished and when work uas initiated on the pressurizer power operated relief valves. The valves were released for maintenance on September 22, and October 13. The tagout review con 0rmed that appropriate remote work control tags had been placed, , but no local tags had been placed on the motor operators for valves 2 RC 403 and 2 RC 405. On September 22,1990, at approximately 6:30 pm, the pressurizer manway was removed. This established a vent path during the time in which maintenance occurred and obviated the need to maintain boundary valve protection. , inspector review and discussions with health physics personnel indicated that access to both the pressurizer block valves and power operated relief valves require high radiation area access controls. Actual radiation levels at the motor operated block valves constitute a radiation area, but a worker had to traverse a hot spot field of about 8.0 rem / hour to gain access to the relief salves. Inspector discussions with the job supervisor indicated that he was aware of the taf sequence and that access to the four valves required high radiation controls. The job supervisor did not feel that hanging a boundary tag on the associated block valves was required. 1 . W D'
M On September 22, during release of work order M2 89 05344, a time existed during which the pressurizer manway was still installed and locally tagging a boundary valve would have been us'.ful; however, the area was controlled as a high radiation area. Therefore the discretion exercised to not hang local tags c was acceptable per the ACP-QA 2.06A. ; Conclusion The inspector found that procedure ACP-QA 2.06A permits the exercise of , discretion concerning hanging boundary tags in high radiation areas. The inspector concluded that the discretion exercised by the job supervisor was appropriate. No unsafe conditions were identified. 3,7 Worker Attentiveness to Duty - Unit 2 The NRC resident inspector office inspected a concem that in two separate events licensee workers were reportedly found asleep while on duty. The first incident concerned a plant equipment operator (PEO) working in the Millstone 2 containment on September 16, who allegedly was found asleep three times, and was aroused the last time by the operations supervisor. The second incident reportedly occurred around October 20 and involved a fire watch who was found asleep in the Millstone 1 cable vault. NRC followup of the events could not substantiate the fire watch concerns, and only partially substantiated the PEO concern as described below. 3.7.1 Plant Ecuinment Oomtor Performine valve Testing - Unit 2 The inspector interviewed the Unit 2 operations supervisor, the Unit 2 plant equipment operator, and an operations person. All interviewees ' agreed upon the ongoing activities at the time; the date, location, and individual involved. The activities involved containment penetration local leak rate testing. The time was between 7:00 - 8:00 pm on September 16, and the location of the work was the ground elevation inside containment. The Unit 2 operations supervisor observed the individual during setup activities for local leak rate testing on September 16. The supervisor did not observe the individual to be inattentive to duty; only that the individual was sitting down and leaning against some cloth material. The supervisor did not see any need to discipline the individual. - However, he did inform the PEO's shift supervisor that the resting position he was in was not appropriate to the situation. The inspector interviewed the plant equipment operator who stated that he was attentive to duty and recognized during activities that he should present - a more active position.
17 l The licensee has declared an Unusual Event on five oc..sions since January 1989. On three occasions in which ESF systems were declared inoperable, and shutdown was initiated but not achieved, the licensee declared an unusual event and notined the NRC pursuant to 10 CFR $0.72. On one occasion involving ( the feedwater coolant injection system, the technical specification limiting condition for operation was not exceeded and shutdown was not required. In ' this case, the licensee declared a *ceneral interest event
- pursuant to its agreemern with the state of Connecticut and notified the NRC in accordance .
with 10 CFR 50.72. The inspector concluded that the licensee is classifying events involving loss of an ESF function properly, and in accordance with NRC requirements. This item is closed. 4.0 Radiological Controls 4.1 Posting and Control of Radiological Areas All Units During plant tours, posting of contaminated, high airbome radiation, and high radiation areas was reviewed with respect to boundary identification, locking requirements, and appropriate hold points. The inspector had no significant observations. 4.2 Radiochemistrv Sampling Unit 2 On or about September 27,1990, an authorized work order (AWO) was ' initiated to allow a vendor to pump out and clean a number of oil water separator sludge tanks (sewers). After pumpdown of the number 3 tank, a radiochemistry sample of the removed sludge was taken and some trace amounts of Cs 138 and Co-60 were identified. The contents were pumped to a, ' truck and the truck was decontaminated. The waste is in storage and will be processed as radioactive material. Plant personnel have initiated a plant incident report to investigate the source of the low level contamination and to ensure adequate controls are in place to prevent unmonitored releases from the oil water separator tanks. The inspector had no further questions regarding this licensee activity. The inspector crincluded that licensee actions were appropriate. l
d 61 i evidence of cracked welds, missing bolts, loose fittings or damaged brackets, t The lift pole and ICI plate threads were in good condition, with the exception that one third turn on the lift pole starting thread was damaged, ( With the exception of C 16. no damage was observed on the ICis. Thimble C- r 16 had a 16 inch longitudinal split along one of the four fluted sections of the f detector sheath. The fluted section keeps the ICI centered within the instrument tube. The ICI detector will remain centered in the tube and its /- operation will not be affected by the damaged section. In the unlikely event ' that the thimble tube separated during the operating cycle, it would remain captured in the fuel assembly guide tube and ivould not become a loose part. If the damaged ICI failed to function, the remaining 44 detectors prcvide : adequate margin to the minimum number required by TS 3.3.3.2 to suppon plant operation. Based on the engineering reviews and examinations, the licensee concluded that - the ICI plate drop caused no damage that would affect adversely reactor safety or prevent continued reactor operation, inspector Reviews and Conclusions The inspector reviewed the videotapes of the ICI and UGS structure, int:rviewed personnel !nvolved with the examinations, and reviewed the engineering evaluations of the consequences of the drop. The inspector noted that the lift tool installation enor occurred as a result of a ' combination of inadequacies in the associated procedure, familiarity of the personnel with the job, and supervision of the work. The event constituted a licensee failure to assure the satisfactory completion of a critical step in the refueling sequence. The inspector noted that the error is one of several personnel performance issues that have occurred during the refueling outage. This NRC concern was addressed to the licensee for action and response in ; NRC inspection report 50-336/9018, and will be followed as part of that inspection. Tne inspector concluded that licensee inspections, engineering and reportability evaluations, and conclusions were proper, ne licensee's followup assessment of the event and its causes was extensive and thorough. Engineering support to evaluate the consequences of the event was good. 9.3 Loss of Containmg itecrity Durine Fuel Movement Description of Event On October 2, reactor refueling operations were in progress, with fuel l i __ _ _ _
62 mosement ongoing in the conttinment and in the spent fuel pool. During refueling, containment integrity is established to mitigate the potential , consequences of a postulated accideat involving the dropping of an irradiated fuel bundle. To satisfy containment integrity requirements, the equipment ( hatch must be installed, at least one door of the personnel air lock must be l' closed, and penetrations either must be secured or capable of httomatic isolation. The licensee had established containment integrity to satisfy the requirements of technical syciftcadon 3.9.4 as a prerequisite for rcfueling. plant operators were also preparing to drain steam generator #1 (SG#1). The operators were using Step 5.1.1 of OP 2316A, Main Steam System, to establish a drain vent path using the atmospheric domp vklves (ADV). The operator followed step 5.11.6.6 of the pn>cedure to open the SG#1 dump valve. Opening the dump valve; also required clearing of a safety tag. The SG#1 dump valve was tagged closed on 9/25/90 per clearance M2 2129 90 when the steam generator manway was opened to support steam generator maintenance actisities. The tigging order stipulated that the atmospheric valve had to be kept closed (along with several other valves) at the direction of the shift supervisor for containment boundary prote: tion. Thi', control was teenforced by a caution in OP 7316A, which stated that the dump valve should not be opened while performing core alterations in order te assure *. hat technical specification .t9.4 requirements were met. The supervisory control room operator on duty on Ocober 2 was aware that tne secondary manway was open and of the operating procedure caution, but failed to recognize that cle.aring the tag to open the ADV was prohibited under existing plant conditions and would violate containment integrity. . The dump va've was opent.d at 6:45 pm on October 2 to support the draining evolution. The vent path was opened for about I hour and 5 minutes, when, at 7:50 pm, the duty outage coordinator, a shift supervisor, and a senior - reactor operator (SRO), noted the open status of the ADV The SRO immt;diately notified the shift personnel that containment integrity requirements were not satisfied. Refueling activities were suspended and, by 8:00 pm, the ADV was closed, reestablisning containment integrity. Fuel handling logs and records (ENG Form 210081, page 9 of 73) show that a single fuel bundle had been moved during the t}rne when containment integrity was compromised. Fuel bundle N-45 was inseded in core location T. 7at6:42 pm. As the next move in sequence, fuel bundle K-25 was moved from core location T 9 and insened in the north upender at 7:10 pm. No further fuel movement occuiTed from then untu refueling activities were halted l 1 r
63
,' at 7:50 pm, as reactor engineering personnel investigated a problem with a hoist limit switch and processed a temporary procedure change to OP 230312 to revise a bridge coordinate, t The licensee initiated plant information report 90 l(B to document the event and evaluate the incident. Se event was reported to the NRC as required by .
10 CFR 50.73 (h)(2)(i)(B) as licensee event report (LER) 9018 dated November 1,1990. CMLcG1tal Licensee review attributed the caut.c of the event to persormel error. The open manways would have established an adequate vent path for the dralning activities and obviated the need to open the dump valves. Inspector reviews noted that the status of the steam generator m;mways was covered during shift turnoser and briefings. Discussion with the opera.or indicated that he was aware ol'.he procedure ud tagging requirements but failed to appreciate the consequence of opening the dump valve. The operator focused on the draining evolution nnd failed to recognize that opening the. ADV was prohibited under the existag plant conditions and would violate containment in:cgrity. l .
" M- M- - _ _ - - - - - _ _ _ _ _ _ _ , , , , , , _ , _ _ , , _ _ _ , _ _ _ _ _ _ _ _ _ _ _
z 64 t Licensee Actions and Evaluations Upon discovery of the violation, actions were taken immediately to meet the , requirements of TS 3.9.4. The licensee's assessment was that there was no , actual impact on worker or public safety at the time since no radiological source term existed during the 75 minute period in which containment integrity , mmpromised. 1 .erder to prevent recunence of the event, the following actions were taken: (i) the caution in OP 2316A on use of the ADVs was moved from step X to Y, to place it closer to the instruction where the operator takes the action to open the valves as part of the drain down evolution; and, (ii) operations supervisors were counseled regarding the need for greater attention to detail during the performance of extensive maintenance work and changing plant conditions. The inspector reviewed the licensee's responses and determined that they adequately addressed the root cause. The licensee's evaluation of the event was provided in LER 9018. The inspector reviewed the evaluation with licensee personnel. As no fuel handling sccident occurred during the event, there were no actual technical consequences. The licensee completed an additional assessment of the potential consequences had a fuel drop accident occurred. During the 75 minutes when containment integrity was lost, the actual fuel handling inside containment took place for 25 minutes and involved the movement of one fuel bundle from the ' core to an upender. . i The dump valve is an eight inch diameter, air operated valve (referei..e drawing 25203 26002). The licensee determined that the valve was manually opened two turns off its seat for the draining evolution, which was calculated ; to be 1/2 inch of valve travel, and resulted in an opening of 0.087 square feet. Using offsite information system (OFIS) data to review containment pressure from 5:00 pm to 9:00 pm on October 2, the licensee noted that containment pressure was positive at about 2.0 inches of water, and, funher, was constant during that period indicating that the open dump valve had no apparent affect on the containment boundary. Nonetheless, the licensee conservatively assumed, for the purpose of the assessment, that the positive pressure would have resulted in flow out of the containment during a postulated fuel handling accident. The calculated flow rate from the containment under the prevailing conditions would be 300 cubic feet per minute (cfm). The licensee compared the consequences of the postulated event under the above conditions with the FSAR analysis for a fuel handling accident. The E FSAR analysis assumes a fuel decay time of 72 hours, whereas the actual fuel decay time on October 2 was 16 days, thus the potential source term is reduced l .
. 65 significantly. Funher, the FSAR analysis assumes that the containment puree vahes are open initially and would remain open for 10 minutes during the event, which would result in a release to the environment at a flow rate of 32,000 cfm. The calculated 300 cfm discharge rate would result in a \
significantly reduced release rate. The licensee determined that the FSAR analysis remains bounding and that an event under the conditions prevailing on October 2 would be much less significant than that analyzed. The inspector reviewed the licensce's calculations, analyses and assumptions. W jnspector Reviews and Conclusions The inspector noted that the personnel error by the operator is one of several personnel performance issues that have occurred during the refueling outage. This NRC concern was addressed to the licensee for action and response in NRC inspection repon 50 336/9018. Funher a number of issues discussed in. this report funher suggests a problem with attention to detail in canying out of operating activities in accordance with regulatory requirennents and licensee procedures. The failure to maintain containment integrity during fuel movement as required by TS 3.9.4 is an apparent violation of containment integrity technical specifications (50 336/90 22 06).
- 9. .! 13rtPlate Anchor Bolt Corrosion Program to Evaluate Seismic Category 1 Sunnorts ,
Initial NRC review of this issue was documented in Region I inspection report 50 336/90 82, Section 3.4.2, which considered the licensee's dispositioning of degraded anchor bolts on the "C" reactor building closed cooling water .
. (RBCCW) heat exchanger in 1989. This item was reviewed during this inspection period to evaluate the actions taken since the 1989 outage and in progress during the present outage to address the potential suppon degradations.
During interviews with site engineering personnel, the inspector noted that the licensee had previously identified the potential for anchor bolt corrosion and the need to address the concern ge..erically, panicularly in light of the l experience with the RBCCW heat exchangers. The corrosion mechanism and the location.of bolt wastage resulted in significant loss of material with attendant loss of marti n to the bolt design strength, with few obvious external l indications of corrosion or degradation. Indirect evidence of underlying i corrosion included cracked grout or rust weepage on or around the support base plates, i l l-l
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EA No. 90 219 Nonheast Nuclear Energy Company A'ITN: hir. E. J. hiroczka o Senior Vice President Nuclear Engineering and Operations P.O. Box 270 ' Hartford, Connecdcut 06141 0270 Gentlemen:
Subject:
NRC ReEi on i Combined Inspection Nos. 50 245/90 20, 50 336/90 22, and 50 423/90 20 This letter transmits the NRC repon of our routine safety inspection that was conducted by hiessrs. D. Dempsey. P. Habighorst, and K. Kolaczyk of this office on September 18 - , November 15, October 2 November 15 (and continued December 313 to evaluate funher one of the significant issues described herein), and October 16 November 15,1990 for hiillstone Units 1,2, and 3, respectively. At the conclusion of the inspection the findings were discussed by the above inspectors with hir. S. E. Scace and other members of your staff. . Areas examined during the inspection are described in the NRC Region 1 inspection report, w hich is enclose,d with this letter. Within these areas, the inspection consisted of observation of activities, interviews with personnel, and document reviews. ; Based on the results of this inspection, two apparent violations were identified at hiillstone Unit 2 and are being considered for escalated enforcement action in accordance with the
General Statement of Palicy and Procedure for NRC Enforcement Actions (Enforcement Poliep,10 CFR Part 2, Appendix C (1990). The apparent violations involve the loss of conta". ment integrity control as a result of personal errors and are discussed in the Unit 2 operations and outage sections of the enclosed repon, Specifically, the apparent violations involve the loss of containment integrity due to the inoperability of the containment purge valve isolation system, and in a separate event, the loss of containment integrity via the No. I steam generator atmospheric dump _ valve. Accordingly, no Notice of Violation is presently being issued for these inspection findings. - Please be advised that the number and characterization of apparent violations described in the enclosed inspection repon may change as a result of funher NRC review.
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- Northeast Nuclear Energy Company An enforcement conference to discuss these appuent violations at Millstone Unit 2 has been - scheduled for Jan, y 15,1991 The purposes of this conference are to discuss the apparent violations, their ca. s and safety significance; to provide you the opportunity to point out any enors in our inspection report; to provide an opportunity for you to present your '
proposed conective actions; and to discuss any other information that will help us determine ' the appropnate enforcement action in accordance with the Enforcement Policy. You will be advised by separate correspondence of the results of our deliberations on this matter. No response regarding these apparent violations is required at this time. The enclosed report addresses your performance during the recent refueling and maintenance outage on Millstone 2. Overall, we found the control of outage activities to be good, with effective management of planned activities and aggressive followup of problems. The thorough evaluation of unplanned events, the extensive support by corporate engineering and vendors to disposition of these issues, and the effective interface between site and corporate enginecting were notable strengths. Your assessment of the persont.el performance aspects of these events was requested in our letter to you dated November 5,1990, enclosing inspection Report 50 336/90 18. Notwithstanding the above conclusion regarding generally good performance, we no.ed a number of esents attnbuted to personnel enor, that apparently resulted from procedure ' quality and adherence weaknesses. Further, the failure to satisfactorily complete a critical step during the replacement of in core instruments that resulted in the dropping of the incore instrument suppon plate was signi6 cant (see section 9.2 of the enclosed repon). Our assessment was that the lift tool installation enor resulted from a combination of inadequacies in procedure details, personnel experience, and supervision of the work activity. The event demonstrates the need for greater diligence in the review process for "tried and proven" procedures to eliminate any over reliance on personnel experience for critical activities. Based on the results of this inspection at Millstone Unit 1, certain of your activities appeared to be in violation of NRC requirements, as specified in the Notice of Violation enclosed
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herewith as Appendix A. We are concemed about the violation because it involved the operation of Millstone 1 with non conservative setpoints on the steam jet air ejector radiation monitor. You are required to respond to this violation and should follow the instmetions specified in Appendix A when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrcnee. Afic reviewing your response to Appendix A, including your proposed corrective actions and the results of future inspections, the NRC will deterrnine whether further NRC enforcement l ' action is necessary to ensure compliance with NRC regulatory requirements, l In addition, certain of your activities at Millstone Unit 2 appeared to be in deviation from l your written commitments, as specified in the Notice of Deviation enclosed herewith as Appendix B. We are concerned about the deviations because they involved the failures to 2 l
. DEc 2 81993 Nonheast Nuclear Energy Company reactor protection channels and to operate the loose parts monitor in accordance with Final Safety Analysis Report commitments. You are requested to respond to these deviations, and should follow the instnictions specified in Appendix B in preparing your response.
\
In accordance with 10 CFR 2.790 of the NRC's
- Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.
The responses directed by this letter and the enclosed Notices are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduc , Act of 1980, Pub. L. No. 96.511. Your cooperation with us is appreciated. Sincerely, i V
. l'es . Hehl' }, rector ,
, j ision of Reach, r4tojects
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. Appendix B, Notice of Deviation
- 3. NRC Region 1 Combined Inspection Report No. 50 245/90-20; 50 336/90 22; 50 423/90 20 .
cc w/enels: W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Services R. M. Kacich, Manager, Generation Facilities Licensing S. E. Scace, Station Director, Millstone H. F. Haynes, Nuclear Unit Director, Millstone Unit ! J. S. Keenan, Nuclear Unit Director, Millstone Unit 2 l ' C. H. Clement, Nucicar Unit Director, Millstone Unit 3 Gerald Garfield, Esquire f Public Document Room (PDR) . L l_ocal Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) NRC Senior Resident inspector l State of Connecticut i i ! 3 l i I 1
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U.S. NUCLEAR REGULATORY COhiMISSION REGION 1 , I Report No.: 50 245/90 20; 50 336/90 22; 50-423/90 20 j Docket No.: 50 245; 50 336; 50 423 t License No.: DPR 21; DPR-65; NPF 49 Licensee: Nonheast Nuclear Energy Company P. O. Box 270 Hanford, CT 06141 0270 Facility Name: hiillstone Nuclear Power Station, Units 1,2, & 3 Inspection At: Waterford, CT Dates: September 18 November 15,1990 (hiillstone 1) October 2 November 15, and December 3 - 13,1990 (hiillstone 2) October 16 November 15,1990 (hfillstone 3) Reporting Inspectors: D. A. Dempsey, Resident inspector, Unit 1 P. J. Habighorst, Resident inspector, Unit 2 K. S. Kolaczyk, Resident Inspector, Unit 3 Inspectors: W. J. Raymond, Senior Resident inspector , D. A. Dempsey, Resident inspector, Unit 1 P. J. Habighorst, Resident Inspector, Unit 2 K. S. Kolaczyk, Resident inspector, Unit 3 J. S. Stewart, Senior Project Engineer .. A. Vegel, Reactor Engineer Approved by: dkh @ /2/2J[90 Date Donald R. Haverkamp, Chief f Reactor Projects Section 4A Division of Reactor Projects
]mpection Summarv: Report 50 245/90 20; Report 50 336/90 22; Report 50-423/90 20 Areat inscected: Routine NRC resident inspection of plant operations, radiological controls, maintenance, surveillance, security, outage activities, licensee self assessment, and periodic reports.
Results: See Executive Summary i
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13 3.6 Imoroner Tagging concem - Unit 2 4 On October 19, the inspector reviewed a concern regarding improper equipment isolation controls. Specifically, no local tags were hung on motor- ( operated valves 2 RC 403 and 2 RC 405 (pressurizer relief isolation) during maintenance work. The valves are considered to be boundary isolation valves, as defined in procedure ACP-QA 2.06A, section 6.1.8 and, thus should have < been tagged locally. No local tags were hung. The inspector noted that ACP-QA 2.06A, section 6.1.8, requires that, in addition to normal equipment tagging, local operators of motor and pneu;natically operated valves be tagged when the valve is used as a system isolation boundary point. If the local operator is in a high radiation area, placement of safety tags is left to the discretion of the shift supervisor, senior control operator, or job supervisor. The inspector reviewed the tagouts for valves 2 RC-402 and 2 RC-404, and work orders M2-90 09844, M2 89-05344 and M2 90 09843. The purpose of the resiew was to determine when the tagouts were accomplished and when work uas initiated on the pressurizer power operated relief valves. The valves were released for maintenance on September 22, and October 13. The tagout review confirmed that appropriate remote work control tags had been placed, but no local tags had been placed on the motor operators for valves 2 RC 403 and 2 RC 405. On September 22,1990, at approximately 6:30 pm, the pressuiner manway . was removed. This established a vent path during the time in which maintenance occurred and obviated the need to maintain boundary valve protection. Inspector review and discussions with health physics personnel indicated that access to both the pressurizer block valves and power operated relief valves require high radiation area access controls. Actual radiation levels at the motor-operated block valves constitute a radiation area, but a worker had to l traverse a hot spot field of about 8.0 rem / hour to gain access to the relief l valves. Inspector discussions with the job supervisor indicated that he was aware of the tag sequence and that access to the four valves required high radiation controls. j j- The job supervisor did not feel that hanging a boundary tag on the associated l - block valves was required. I~ \
.f 1
l' 14 On September 22, during release of work order M2 89 05344, a time existed
' during which the pressurizer manway was still installed and locally tagging a boundary valve would have been useful; however, the area was controlled as a high radiation area. Therefore the discretion exercised to not hang local tags '
was acceptable per the ACP-QA 2.06A. Conclusion , The inspector found that procedure ACP-QA 2.06A permits the exercise of discretion concerning hanging boundary tags in high radiation areas. The inspector concluded that the discretion exercised by the job supervisor was appropriate. No unsafe conditions were identified. 3.7 Worker Attentiveness to Dutv - Unit 2 The NRC resident inspector office inspected a concern that in two separat events licensee workers were reportedly found asleep while on duty. The first incident concerned a plant equipment operator (PEO) working in the Millstane 2 containment on September 16, who allegedly was found asleep three times, and was aroused the last time by the operations supervisor. The second incident reportedly occurred around October 20 and involved a fire watch who was found asleep in the Millstone I cable vault. NRC followup of the events could not substantiate the fire watch concerns, and only partially substantiated the PEO concem as described below. 3.7.1 Plant Equipment Ontrat.or Perform!nc Valve Testine Unit 2 The inspector interviewed the Unit 2 operations supervisor, the Unit 2 plant equipment operator, and an operations person. All interviewees ' agreed upon the ongoing activities at the time; the date, location, and individual involved. The activities involved containment penetration local leak rate testing. The time was between 7:00 8:00 pm oa September 16, and the location of the work was the ground elevation inside containment. The Unit 2 operations supervisor observed the individual during setup activities for local leak rate testing on September 16. The supervisor did not observe the individual to be inattentive to duty; only that the individual was sitting down and leaning against some cloth material. The supervisor did not see any need to discipline the individual. However, he did inform the PEO's shift supervisor that the resting position he was in was not appropriate to the situation. The inspector interviewed the plant equipment operator who stateo that he was attentive to duty and recognized during activities that he should present a more active position.
17 The licensec has declued an Unusual Event on five occasions since January 1989. On three occasions in which ESF systems were declared inoperable, and shutdown was initiated but not achieved, the licensee declared an unusual event and notified the NRC pursuant to 10 CFR 50,72. On one occasion involving s ' the feedwater coolant injection system, the technical specification limiting . condition for operation was not exceeded and shutdown was not required. In this case, the licensee declared a " general interest event" pursuant to its . , agreement with the state of Connecticut and notified the NRC in accordance w h 10 CFR 50.72. The inspector concluded that the licensee is classifying ,, events involving loss of an ESF function properly, and in accordance with NRC requirements. This item is closed. 4.0 Radiological Controls , 4.1 Eosting and Control of Radiological Areas - All Units During plam tours, posting of contaminated, high airborne radiation, and high radiation areas was reviewed with respect to boundary identifica: ion, locking requirements, and appropriate hold points. The inspector had no significant observations. 4.2 BMiochemistry Sampjjpg Unit ? On or about September 27,1990, an authorized work order (AWO) was ' initiated to allow a vender to pump out and clean a number of oil water separator sludge tanks (f. ewers). After pumpdown of the number 3 tank, a radiochemistry sample of the removed sludge was taken and some trace amounts of Cs 138 and Co-60 were identified. The contents were pumped to a7 truck and the truck was decontaminated. The waste is in storage and will be processed as radioactive material. Plant personnel have initiated a plant incident report to investigate the source of the low level contamination and to ensure adequate controls are in place to prevent unmonitored releases from the oil water separator tanks. The inspector had no further questions regarding this licensee activity. The-inspector concluded that licensee actions were oppropriate.
l 1 61 i evidence of crscked welds, missing bolts, loose fittings or damaged brackets. The lift pole and ICI plate threads were in good condition, with the exception that one third turn on the lift pole starting thread was damaged.
)
( With the exception of C 16, no damage wa observed on the ICis. Thimble C- l 16 had a 16 inch longitudinal split along one of the four fluted sections of the detector sheath. The fluted section keeps the ICI centered within the instrument tube. The ICI detector will remain centered in the tube and its , i operation will not be affected by the damaged section. In the unlikely event that the thimble tube separated during the operating cycle, it would remain l captured in the fuel assembly guide tube and would not become a loose part. If the damaged ICI failed to function, the remaining 44 detectors provide i adequate margin to the minimum number required by TS 3.3.3.2 to support plant operation. Based on the engineering reviews and examinations, the licensee concluded that the ICI plate drop caused no damage that would affect adversely reactor safety I or prevent continued reactor operation. Mperor Reviews and Conclusions ! The inspector reviewed the videotapes of the ICI and UGS structure, J interviewed personnel involved with the examinations, and reviewed the engineering evaluations of the consequences of the drop. The inspector noted that the lift tool installation error occurred as a result of a ' combination of inadequacies in the associated procedure, familiarity of the personnel with the job, and supervision of the work. The event constituted a licensee failure to assure the satisfactory completion of a critical step in the refueling sequence. The inspector noted that the error is one of several ; personnei performance issues that have occurred during the refueling outage. This NRC concern was addressed to the licensee for action and response in NRC inspection report 50 336/9018, and will be followed as part of that ! inspection. The inspector concluded that licensee inspections, engineering and reportability evaluations, and conclusions were proper. The licensee's followup assessment of the event and its causes was extensive and thorough. Engineering support to evaluate the consequences of the event was good. 9.3 Less of Calainment Integrity Qurine Fuel Movement Descrit) tion of Event On October 2, reactor refueling operations were in progress, with fuel l
- y- w w- A -,m.g --c- -1
. 62 movement ongoing in the containment and in the spent fuel pool. During - refueling, containment integrity is established to mitigate the potential consequences of a postulated accident involving the dropping of an irradiated fuel bundle. To satisfy containment integrity requirements, the equipment hatch must be installed, at least one door of the personnel air lock must be \ closed, and penetrations either must be secured or capable of automatic isolation. The licensee had established containment integrity to satisfy the ' requirements of technical specification 3.9.4 as a prerequisite for refueling.
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Plant operators were also preparing to dtain steam generator #1 (SG#1). The operators were using Step 5.1.1 of OP 2316A, Main Steam System, to esteblish a drain vent path using the atmospheric dump valves (ADV). The operator followed step 5.11.6.6 of the procedure to open the SG#1 dump valve. Opening the dump valve also required clearing of a safety tag. The SG#1 dump valve was tagged closed on 9/25/90 per clearance M2 2129 90 when the steam generator manway was opened to support steam generator maintenance activities. The tagging order stipulated that the atmospheric valve had to be kept closed (along with several other valves) at the direction of the shift supervisor for containment boundary protection. This control was reenforced by a caution in OP 2316A, which stated that the dump valve should not be opened while performing core alterations in order to assure that technical specification 3.9.4 requirements were met. The supervisory control room operator on duty on October 2 was aware that the secondary manway was open and of the operating procedure caution, but failed to recognize that clearing the tag to open the ADV was prohibited under existing plant conditions and would violate containment integrity. The dump valve was opened at 6:45 pm on October 2 to support the draining evolution. The vent path was opened for about I hour and 5 minutes, when, at 7:50 pm, the duty outage coordinator, a shift supervisor, and a senior reactor operator (SRO), noted the open status of the ADV. The SRO immediately notified the shift personnel that containment integrity reqt'irerents were not satisfied. Refueling activities were suspended and, by 8:00 pm, the ADV was closed, reestablishing containment integrity. Fuel handling logs and records (ENG Form 210081, page 9 of 73) show that a single fuel bundle had been moved during the time when containment ' integrity was compromised. Fue! bundle N-45 was inserted in core location T-7 at 6:42 pm. As the next move in sequence, fuel bundle K 25 was moved from core location T 9 and inserted in the north upender at 7:10 pm. No funhet fuel movement occurred from then until refueling activities were halted
63
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at 7:50_ pm, as reactor engineering personnel investigated'a problern with a - - hoist limit switch and processed a temporary procedure change to OP 230312 to revise a bridge coordinate. t The licensee initiated plant information report 90109 to' document the event and evaluate the incident. The event was reported to the NRC as required by_ l
' i 10 CFR 50.73 (a)(2)(i)(b) as licensee event report (LER) 90-18 dated November 1,1990. -j Dats01 Ext 31 I icensee review attributed the cause of the event to personnel error. The open !
mans ays would have established an adequate vent path for the draining i activ' ties and obviated the need to open the dump valves. Inspector reviews no.ed that the status of the steam generator manways was covered during shift j turnoser and briefings. Discussion with the operator indicated that he was i' aware of the procedure and tagging requirements but failed to appreciate the consequence of opening the dump valve. The operator focused on the draining-evolution end failed to recognize that opening the ADV was prohibited under the existing plant conditions and would violate containment integrity, j i I
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I I i l l l i l L
64 l_icensee Actions and Evaluations Upon discovery of the violation, actions were taken immediately to meet the requirements of TS 3.9.4. The licensee's assessment was that there was no ( actual impact on worker or public safety at the time s. e4 no radiological ' source term existed during the 75 minute period it + 1 containment integrity was compromised. In order to prevent recurrence r;f the event, the following actions were taken: :e (i) the caution in OP 2316A on use of the ADVs was moved from step X to Y, to place it closer to the instruction where the operator takes the action to open the valves as part of the drain down evolution; and, (ii) operations supervisors were counseled regarding the need for greater attention to detail during the performance of extensive maintenance work and changing plant conditions. The inspector reviewed the licensee's responses and determined that they adequately addressed the root cause. The licensee's evaluation of the event was provided in LER 90-18. The inspector reviewed the evaluation with licensee personnel. As no fuel handling accident occurred during the event, there were no actual technical consequences. The licensee completed an additional assessment of the potential t consequences had a fuel drop accident occurred, During the 75 minutes when containment integrity was lost, the actual fuel handling inside containment took place for 25 minutes and invohed the movement of one fuel bundle from the core to an upender. The dump salve is an eight inch diameter, air operated valve (reference drawing 25203 26002). The licensee determined that the valve was manually opened twa tums off its seat for the draining evolution, .which was calculated ; to be !!2 inch of valve travel, and resulted in an opening of 0.087 square feet. Using offsite information system (OFlS) data to review containment pressure from 5:00 pm to 9:00 pm on October 2, the licensee noted that containment pressure was positive at about 2.0 inches ( water, and, further, was constant during that period indicating that the open dump valve had no apparent affect on the containment boundary. Nonetheless, the licensee conservatively ' assumed, for the purpose of the assessment, that the positive pressure would have resulted in flow out of the containment during a postulated fucl handling accident. The calculated flow rate from the containment under the prevailing conditions would be 300 cubic feet per minute (cfm). The licensee compaed , ' consequences of h postulated event under the above conditions with the FSAR analysis for a fuel handling accident. The ' FSAR analysis assumes a fuel decay time of 72 hours, whereas the actual fuel decay time on October 2 was 16 days, thus the potential source term is reduced ':
y 0 ' aJ ' 65. signincantly/ Further, the FSAR andysis assumes that the containnient purge
- c vahes are open initially and would remain open for 10 minutes during'the ~
event, which would result in ~a release to the environment at a flow rate of;
- 32,000 cfm. The calculated 300 cfm~ discharge rate would result in a significantly reduced release rate. The licensee determined that the FSAR : \
analysis remains bounding and that an event under the conditions prevt.iling on _ October 2 would be much less significant than that analyzed. iThe inspector ' reviewed the licensee's calculations, analyses and assumptions. Inspector Reviews and Conelydom The inspector'noted that the personnel error by the operator is one of several: personnel performance issues that have occurred during the refueling outage.
- This NRC concern was addressed to the licensee for action and response in : ,
NRC inspection report'50 336/9018. Further a number of issues discuss'ed in this report further suggests a problem with attention to detail in carrying out' of.- - operating activities in accordance with regulatory requiremetits and licensee L' procedures. j The failure to maintain ~ containment integrity during fuel movement as required '. by TS 3.9.4 is an appaient violation of containment integrity technical specifications (50-336/90-22 06). l 93 . Rase Plate Anchor Bolt Corrosion Program to Evaluate Seb_rqigfategory 1 O Succorts - l. ! Initial NRC review of this issue was documented in Region I inspection report 50-336/90 82, Section 3.4.2, which considered the licensee's dispositioning of_ degraded anchor bolts on the "C reactor building closed cooling water 3 t
.. tRBCCW) heat exchanger in.1989.1 This' item'was reviewed during this '. .
- . inspection period to evaluate the' actions; takers since the 1989 outage and in _ .3
- progress during the present outageao address the potential supportL degradations.. -
During interviews with site engineering personnel; the inspector neted that the
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licensee had previously identified the potential for anchor bolt corrosion and -
'the need to address the concern generically, particularly in light of the _
experience with the RBCCW heat exchangers,1 The corrosion mechanism and ? the location of bolt wastage resulted in significant loss ~of matedal with
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attendant loss of margirito the bolt design strength, with few obvious external ~
' indications of corrosion or degradation. Indirect evidence of underlying corrosion included cracked grout or rust weepage on or around the support base plates, t
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L L J <203 m m L. 4 e December 21, 1990 Docket No. 50-336 39166
- Mr. E. C. Venzinger, Chief Projects Branch No. 4 Divisfor. of Reactor Projects U. S. Nuclear Regulatory Commission Region I 475 Allendale Road i King of Prussia, Pennsylvania 19406 p
Dear Hr. Venzinger:
Millstone Nuclear Pover Station, Unit No. 2 . RI-90-A-0204 and RI-90-A-0200 i 7 l l Ve have completed our review of an allegation concerning activities at. Hillstone Unit No. 2 (RI-90-A-0204). As requested in your transmittal' letter, l i nur response does not contain any' personal privacy, propriecary, or safeguards ! Information. The material contained in this renponse may be released to the - public and placed. in the NRC Public Document room at your, discretion. The NRC l letter and our response have received controlled and limited distribution on a- )
"need to knov a basin during the preparation of-this response. Based upon our request on December 20, 1990, Region I personnel extended the due date-for !
RI-90-A-0200 to January 21, 1991. Additional. time was needed to ressive technical issues in order to provide a complete response. .I L RI-90-A-0204 Issue _1 ( l Background r These same issues vere raised by an instrument speelalist during the head: i cabling vork activity. There is a close correlation'of the questions posed-by the memo and response. This documentation is available for your staff's i l review. I l N hh /h-~ , l OS3Gt ATV 4 a6
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Hr. E. C. Uanzinger, Chief U. S. Nuclear Regulatory Commission 2 A09166/Page 2 December 21, 1990 The following problems exist with IC 24210: g s
- 1. Step 5.5.6.1: calls for visually inspecting the connector assemblies on each cable for signs of degradation or damage. Problems have been identified with Z-1, 24, 87, and.Z-2, #8. The cables vere put together anyway. Please explain. Further, the step calls for visually inspecting the Grafoil gasket at the Litton-Veam connector, but no information is ,
provided about these connectors or the Grafo11 gaskets. Are I&C technicians trained on these items? Does the procedure require upgrades to explain what the technician is looking for in this step? No figure is provided with the procedure to identify the connectors, and it:is impossible to read the etched numbers on the connectors. How are the connectors to be identified on the job?
Response
The activity of reassembling the HJTC connectors vas begun on a Sunday with an upgraded spC:lalist serving as the first line supervisor and no engineering support readily available During the work activity, several questions were raised and documented. The following day, Honday,_these questions were investigated and resolved by more knowledgeable I&C and Engineering personnel. Connector training performed to date has not included the specific knowledge needed to perform this task. Training requirements for this task are being reviewed for addition to the 16C techaical training . program. As written, the procedure makes the assumption the I&C technician using it has the knowledge necessary to determine (l'e gasket-location and - condition. Ar a result of this experience, the procedure owner has been ; instructed to revise the procedure to include the necessary information. The connections are identified by a small stenelled number. They are also staggered and are in the order of #1 being the highest and #8 being the lovest. As lighting is limited and access is difficult, the normal method for connector identification has been its order of elevation.
- 2. The caution on page_14 is impossible to achieve as only about a 45-degree turn is possible. Is this technically satisfactory? Does this indicate procedural non-compliance during past performance? Has there been repeated coranection damage in complying with this step Response.
On the following Honday, additional technical information was obtained from the responsible vendor. Procedure' changes were implemented to allow an alternative method of torquing the connector. This method was used to ensure the proper assembly'of'the connectors. No significant connector damage has occurred.
I B Hr. F.. C. Uenzinger, Chief U. S. Nuclear Regulatory Commission
. AG9166/Page 3 December 21, 1990
- 3. Dust cover caps are shovn in figures in IC-2421C and IL-2419C. These caps t are not being used. Are dust covers needed? Why are they not used if shown in the figure?
Response
The use of dust covers is not delineated by the procedure. Dust covers only appear in the figure. The need for their use is currently being evaluated. The existing figure has been replicated from.a vendor draving that shows the "as supplied" HJTC probe. The figure was not intende/ to dictate the use of dust capa in the field. The figure also has shortcomings in not supplying the necessary detail needed for inspection and assembly of the connectors-
- 4. St.ep 5.6.4 calls for the verification of the RJTC probes per IC-2419C.
Why is this not dor.e prior to the connection of the detector cables?
Response
'this HJTC probe had been previously used. There vas no need to perform the verification of the known good probe. The retest of the overall system response verifies the condition of the probe, as well cs tne entire system. The step has since been removed from the procedure.
- 5. Are any generic problems with procedural non-compliance or laxity with regards to procedural adherence evidenced by these problems? Please explain.
Response
There are no generic problems with procedure non-compliance as evidenced by these problems. Note taat this procedure has not been upgraded as part of the station procedure upgrade program. Veaknesses in such procedures like this one have previously been discussed with the NRC staff. This example serves to demonstrate that with the proper attention to detall-by the technicians involved and with the proper response and direction from their supervision that a proper quality activity results. Where l appropriate, the necessary NCRs and procedure changes were implemented, and the noted need for procedure enhancements took place. Issue 2 l Instrument Calibration Reviev ICR 90-113 vas vritten on November 2, 1990. Please provide the resolution documentation for -the ICR. l
Response
A copy of the Inctrument Calibration Review ICR 90-113 is attached. I
. Hr. E. C. Ucnzing2r, Chief U. S. Nuclear Regulatory Commission 3 A09166/Page 4 December 21, 1990 Issue 3 ( !
Recently, an ICR vas generated concerning out-of-specification test voltages found during the' performance of SP-2404C. Please provide the resolution documentation for the ICR. Historically, the reference voltages have been out l of specification e March 9, 1990 AVO H2-90-02559 May 5, 1990 AVO H2-90-02736 May 5, 1990 AVO H2-90-05237 May 5, 1990 AVO H2 90-05480 l August 30, 1990 Surveillance, SP-2401F, and SP-2401B Please discuss the operability of Channel "C" of the RPS with the historic out-of-specification test voltages. Metc, Drawings 25203-39069, sheet 40, and 25203-25193, sheet 6, supposedly identify the reference test voltage applications.
Response
There is no I&C procedure SP-2404C. Therefote, it is not.possible to supply any information concerning An ICR generated while performing this procedure. The AV0s referenced vere written against RPS channel "C" Reactor Protection System Calibration and Indication Panel (RPSCIP). There are no test voltages generated in this instrument drawer. However, the RPSCIP digital volt meter (DVM) is used to monitor the core protection calculator (CPC) logic power supply output voltage (+/-10vde). A review of surveillance and equipment history shows that following replacement of a-different power supply (+5vde) in May 1989 by PDCR HP2-89-028, the RPSCIP DVH indicated a .003vde offset-from the actual +/-10vde power supply output. Investigation during the 1990 refueling shutdown, while performing PDCR HP2-89-068, revealed a difference in ground potential between the RPSCIP and the CPC. This. caused the RPGCIP DVH to indicate incorrectly by .003ved. This is at the limit of the tolerance. applied to the DVH indication (+/ .003vde). A design change notice (DCN
#DH2-P-022-00) was written against PDCR MP2-89-068 and implemented to correct the problem. As can be seen by reviewing the referenced AV0s, the problem was i identified shortly after the +5vde modification and determined to bel limited '
to the DVH indication only. An AVO to investigate the problem was.vritten at !. that time and scheduled for the refuel. Also referenced are two surveillance procedures for August 30, 1990. One surveillance concerne the CPC, SP-2401F, the RPS high pover trip test, which l is performed monthly. It documents the above DVH indication problem and shows L the'high power trip function of-the CPC to be operable at this time. The ! other referenced surveillance SP 2401B, concerns the vide range flux monitor l functional. This proceoure is performed veekly while shut'down. Since Hillstone Unit No. 2 was not shut down during the referenced time, it is l L difficult to determine how this is related to the issue. l l
~ -- . . . -.
Mr. E. C. Venzinger, Chief U. S. Nuclear Regulatory Conmission 4: A09166/Page 5 December 21, 1990 Two drawings are refecenced which " supposedly identify the reference test ' voltage applications". The first drawing, 25203-39069, sheet 40, is an- , electrical schematic of the CPC. It shows the power supplies connection to the circuit but makes no reference to the application of the voltage with respect to absolute values. It has no connection to the vide range flux monitor circuit. The second drawing referenced, 25203-25193, sheet 6, does not exist in the NUSCO drawing system. More specific information vill be required to resolve this issue. Issue 4 A red tag was improperly hung by Operations on the back of C05/UO6. The tag was hung on TDE but should have been hung on TDD. S00(*) knows the details. Please explain the tagout problem and actions taken to both correct the problem and to prevent recorrence. (*) - identity may be obtained from the SRI.
Response
The pins for Veidmuller Block TDD, on the back of C05/C06, vere properly removed and placed in their correct storage location at the time of tagging, under the guidan'ce of an I&C technician. The red tag was properly filled out for Veidmuller-Block.TDD. The tagging discrepancy-was that the red tag was not attached directly to Veidmuller Block TDD but to a cable directly beside TDD vhich was labeled TDF. This tagging discrepancy was noticed by an operator who consulted with an I&C technician. The operator then moved the tag from the cable to the pins for Veidmuller Block TDD. Long-term actions to prevent recurrence is to formally distribute to all Unit 2 Operations personnel a standardized method for tagging Veidmuller Blocks. Issue 5 . 1 A recent annunciator vindov change C04(CRDR) was apparently not reviewed and-completed properly as procedures SP-2401B, 2401F, and 2401J had to be changed during the performance to accommodate the annunciator vindov changes. Please discuss the accuracy of the statement and any actions that you have taken or may take to rectify any identified problems.
Background
As part of the CRDR improvements, several annunciator vindows were relocated; This activity took place over the course of the 1990 refueling. outage. Changes to ILC procedures were appropriate as the procedures contained the detail of the annunciator vindov location. Changes to SP-2401B and'its data sheet.and SP-24GlJ vere processed at the time the procedures vere implemented. No changer were required to SP-2401F. It is necessary . for procedure changes required by a modification to be approved prior to the implementation of the procedure. This was done in this case.
4 Mr. E. C. Vanzinger, Chief-U. S. Nuclear Regulatory Commission-g A09166/Page 6
-December 21, 1990 Response .
Although several changes vere required, no problems requiring corrective actAon exist. All modification-related procedure changes can not be made in advance of the task being assigned. : , Issue 6 ;, On November 5, 1990, upon completion of the CVP portion of SP-240lJ, (the CRDR change was done as a non-intent change on-this date), Channel C T0-7 was left bypassed by technicians after the surveillance was turned in, and the technicians had to be called by the SS and the channel unbypassed by the SCO. Please discuss the procedural compliance aspects of this statement.
Response
This event is an example of procedural noncompliance. SP-240lJ requires restoration of the bypass key. This was not done prior to turning in the surveillance. The appropriate personnel have been counseled with respect to this issue.
.After our review and evaluation, we find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety. We appreciate the opportunity to respond and explain the basis for our actions. Please contact members of my staff if there are any further questions on any of these matters.
Very truly yours, t NORTHEAST NUCLEAR ENERGY COMPANY ( . ~~~~
.r E. J. tifoczka #
Seniod"Vice President ce: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 T'""" ~ +s., * - , , . , ,, ,
s t w i L _ ATTACHMENT 1 m INSTRUMENT CALIBRATION REVIEV FORH 90-113 ISSUE NO. 2 r + w I $ f f-December 1990
( ,s iORM APPROVED j : :.N#e22MADATF. bd"70N MTG.NO. MO -6 6 s O INSTRUMENT CAllBRATION REVIEW FORM Part 1 To be completed by person performing work. t Time Mj.So ICR Number (D-// 2_,,, AWO# MS cio- /~B:275~ Date 11/ E./Clo
-- s, s Instrument or Device Affected ID#_kf-lOOLt Name8Ldg: vine e fhssne- _
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l Cause: (If Facts are Known) ( ) instrument Drift [ ] Unknown y Other Explain; JFT_ fdh.aS $Phn. $bbbt bbhuh,ftfakd1, ()Y/& /_httu,r)$ Would the Instrument or Control System have performed its function as required by Tech Specs? Yes [ ] No % N/A[ ] P Basis: 'T.@_, lid.e.M _ W : tr.clactrument Cr Contrc; System (A! arm. Bjstable Tr;n. 6tc..) fourid in a cou- ;- ia na;;;Una Yes ( ) No (y',) N/A{ ] ,{ Basis: Mcljdp is a PIR recommended? Yesh/] No ( )
- Technician / Specialist 1&C Supervisor Part 2 [o be completed by SS or SCO.
Mode 'E Power 0 "4 Temp. So o *i: Press. 400
- _
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Human Performance Enhancement System (HPES) review required?
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-m.ame The NRC Region I has completed its followup of the concerns that you brought to our our attention by way of the following memoranda; (1), March 1,1990, alleging NRC impropriety in not identifying your findings as problems and an example of a harassing memorandum that you had received; (2), November 4,1990, alleghs procedural violations associated with the overhaul of service water pump P5C; and (3), November 11,1990, alleging procedural violations associated with the temporary repair of a moisture separator reheater manway gasket steam leak.
With regard to issue (1), we referred your assertion of NRC impropriety to the NRC Inspector General. Because the inspector general is an independent office we cannot provide you any coaclusions in this matter. With regard to your receipt of a harassing memorandum, the NRC does not have regulations regarding harassment from co-workers. We remind you to address harassment issues such as this to the Department of Labor for their investigation and followup. With regard to issue (2), we referred your concerns to the licensee for their review and followup. The licensee responded in a letter dated January 4,1991 (attached). Your ' allegation that bearing oil had not been removed prior to moving the pump and that procedures governing the pump overhaul require that the oil be removed is substantiated. We note that appropnate procedure revisions have been implemented to prevent similar occurrences. The NRC intends to complete action on this item in a future inspection report. We consider this matter closed. With regsid to issue (3), we again referred your concems to the liceniec for their followup and resolution. The licensee responded in a letter dated January 21,1991 (attached). The licensee determined that no procedures were violated in the activity described in the allegation. Based on the licensee's response, it appears that your allegation is unsubstantiated. The NRC plans no further action on this matter, fr$/bN0k 1T
1 t We appreciate you informing us of your concems and feel that we have been responsive to those concerns. If you have any additional questions or if I can be of further assistance, please call me collect at (215) 337 5225. t rely; b 8
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( Edward Wenzinger, Chief Reactor Projects Branch 4 ( Attachments: As stated bec: Allegation File RI-90-A-0208, closecut Allegation File RI-90-A-0205, closecut Allegation File RI 88 A-0003, update only.(B22.1, closecut) W. Raymond J. Stewan 1
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4 January 4, 1991 Docket No. 50-336 A09187 r Mr. E. C. Ventinger, Chief Projects Branch No. 4 Division of Reactor Projects U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406 Dear Mr. Venzinger Millstone Nuclear Power Station, Unit No. 2 RI-90-A-0205 We have completed our review of an allegation concerning activities at .-- Millstone Unit 2 (RI-90-A-0205). -As requested in your transmittal letter dated December 6,1990,- our response does not contain any personal privacy, proprietary, or safeguards information. The material contained in this response may be released to the public and placed in the NRC Public Document Room at your discretion. The NRC letter and our response have a basis received controlled and limited distribution on a "need to knov during .the preparation of this response. Issue On November 4, 1990, the motor from P5C vas observed in the parking lot. The oil had not been removed f rom the motor as required by HP-272004, Step The AVO 5.1.2. which is performed prior to disconnecting the motor leads. for the motor removal did not contain NP-2720D4, and this omission is contrary to ACP-2.02C. The AVO did not contain a material accountability log as required by HP-2720C5, Step 4.4.
)
e t 4 Mr. E. C. Venzinger, Chief , U. S. Nuclear Regulatocy Commission A09107/Page 2 January 4, 1991 t Please discuss these atstements and their validity. Are any of the alleged procedural compliance issues valid, and if so, do these occut'rences suggest a generic procedural adherence probles7 i This issue was segregated into four distinct segments, each of which is addressed in turn below.
- 1. Oil had not been removed from the motor prior to removing the motor.
Response
011 should have been removed from the upper and lower bearing housings prior to removing the pump from the screenhouse. Removal of the motor the is the responsibility of the Electrical Job Supervisor however, removal of the oil and moving the motor is actually performed by the mechanics. The procedure governing the pump overhaul does not address removing oil prior to removing the motor. The problem appears to be more a coordination problem rather than a procedure coupilance problem. The procedure governing the pump overhaul vill be revised to ensure better cuordination between mechanics and electricians for the motor reecval task.
- 2. Oil should have been removed from the motor prior to disconnecting the motor leads.
Response
MP-2720D4 is a generic procedure for the overhaul of 4160 volt motors. Unit 2 Maintenance is in the process of preparing specific procedures prepared HP-2720B2 was . for each model of 4160 volt motors. therefore, HP-271082 Ss specifically for the service water pump motors overhaul activities for the procedure that governs the removal and these motors. The steps associated with the removal of _the motor in HP-2720B2 vere resequenced so that the motor leads are disconnected prior to requiring that the oil be removed from the bearing housfags. Therefore, this aspect of the issue is not a concern. l i 3. The AVO for the motor removal did not contain HP-2720D4 which is I contrary to ACP-2.020.
Response
The procedures that are listed in the AVO are n function o# the local l l ' ID of the equipment. The AVO was prepared under the local ID for the pump rather than the -local ID for the motor.states ACP-QA-2.02C Therefore, that anprocedure approved NP-2720B2 was not listed. procedure is required for disassembly, repair, and reassembly of all Category I equipment. AVO M2-90-13312 specified disconnecting motor leads and motor removal; therefore, a procedure vas not specifically required by ACP-0A-2.02C. In reviewing this event, it is evident that
( ~ f.< 4 Mr. E. C. Ventinger, Chief U. S. Nuclear Regulatory Commission A09187/Page 3 January 4, 1991 procedure HP-2702B2 should have been drtofthework package. This procedure provides guidance on the sequencing of the removal of the motor. The procedures listed for the local ID for the pump have been < revised to include the electrical procedures MP-2720B2 and HP-2720CS, and the pump overhaul procedure has been revised to include instruction for resoval of the motor.
- 4. The AVO did not contain a material accountability log as required by MP-272005.
Response
The material accountability precaution was "N/A'd" and initialed by the job supervisor. Material accountability is not required for either the switchgear cubical door or the rear cable compartment cover that encloses the area where workman grounds are installed. Therefore, a material accountability log is not required to be included in the AVO. This clarification vill be added to procedure MP-272005. After our review and evaluation, we find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety. Ve appreciate the opportunity to respond and explain the Please contact members of my staff if there are any basis for our actions. further questions on any of these asetters. Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY ,
/' mA E. Y ) t 6c h.a /
Seni(r Vice President 1, 2, ce: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. and 3
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Docket No. 50-336 A09188 Mr. E. C. Venzinger, Chief Projects Branch No. 4 Division of Reactor Projects U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Venzinger:
Millstone Nuclear Power Station, Unit No. 2 RI-90-A-0206 Ve have completed our review of an allegation concerning activities at Hillstone Unit 2 (RI-90-A-0206). As requested in your transmittal letter dated December 6, 1990, our response does not contain any personal privacy, proprietary, or safeguards information. The material contained in this response may .be released to the public and placed in the NRC Public Document Room at your discretion. The NRC letter and our response have , received controlled and limited distribution on n' "need to knov"- basis ' during the preparation of this response. Issue 1 On November 6, 1990, the I&C Department completed IC-2435F,. which starts the stator cooling system without the main turbine on line. The procedure. sets the flow control valves and records data on pressures and flovs within the system. On November 11, 1990, -a lov stator- flow condition was identified during two-pump operation which in itself is abnormal since one pump is generally in standby vith a lov. pressure. actuation setpoint. Troubleshooting . identified that the flov' control setpoint was not set as required by'IC-2435F. Operations does not. understand the basis for-the lov' flow setpoint of the standby pump. Please discuss the validity of these statements. Are there procedural compliance issues identified in the problems presented? 9/00460/97 - o se.m .... Ng
c. t Mr. E. C. Vensinger, Chief e
- U. S. Nuclear Regulatory Commission A09188/Page 2-January 4, 1991 Background (
The procedure reference is incorrect for stator liquid cooling; IC-2435F is a non-existent. procedure. The correct procedure for this activity is ' IC-2425F. The stator cooling system contains tvo pumps. One is operated , continuously, and the other is considered a reserve. An autostart of the reserve pump is provided by a pressure switch, 63-P60A, which is set to start the reserve pump at n' system pressure that corresponds to the minimum allovable system flov. For additional reliability, an additional switch, 63-P60B, provides a backup autostart signal to the reserve pump. IC-2425F, Stator Cooling System Setup, contains procedural guidance that records system pressures at the required flov settings and uses that information to calculate pressure setpoints for the reserve and backup reserve- pressure switches. This procedure was implemented after completion of maintenance activities on the stator cooling system prior to placing the system in service. During power operation, it van noted that the reserve pump could not be , secured and returned to standby status. Troubleshooting activities noted that the system pressure developed with one pump operating was below the reset value for the reserve autostart pressure switch, 63-P60A. It was also identified that the flow control valve that sets the flovrate to the rectifier cabinet was set at a value of six rather than 32 gpm. This was corrected by AVO H2-90-13938. A setpoint change was then-processed to set the trip value five psi lover than initially calculated via 10-2425F. This , setpoint change vas implemented by AVO H2-90-14323. .These activities resolved the problem of not being able to secure the running reserve stator cooling pump. Response .
- Stator cooling flov was not abnormally lov on Hovember.ll, 1990. On that
- day, Unit 2 16C investigated the problem of not being able.- to secure the reserve stator cooling pump. 'The reset value of the reserve' pump autostart pressure svitch was found to be .above the discharge pressure of the lead-stator cooling pump. The system was found with-the flow control valve for the rectifier cooling set at six rather than 32 gpe. It is apparent that.
the procedural guidance was not followed.
- The procedure vas not properly perforned. The erroneous rectifier cooling flow setting was corrected, and a temporary setpoint was implemented. The procedure vill be revised to -include additional guidance on setpoint determination. - The pump does not have a lov flov setpoint per se. The cystem is designed to start the reserve pump at a system pressure value that is equivalent to the lov system flow condittun.
Mr. E. C. Venzinger, Chief i U. S. Nuclear Regulatory Commission A09188/Page 3 January 4, 1991 Issue 2 On or about November 9, 1990 during recovery from the refueling outage, surveillance procedure SP-2401J, Thermal Margin Lov Pressure (TM/LP) Punction Test, was partially completed in operational mode 3 vith the ' remainder completed in mode 2. Technical Specification Table 4.3-1 requires TM/LP functional tests to be completed in operational modes 1 and 2. Are the above statements valid? If so, do technical specifications and procedures allow this surveillance to be segmented in the two operational modes? Please discuss.
Background
The Control Element Assembly Vithdrawal Prohibit (CVP) is a design feature intended to restrict CEA outvard motion when two or more pretrip conditions occur on the RPS channels of High Power or Thermal Margin / Low Pressure (TM/LP). The TM/LP generated CVP is bypassed below power levels of-10E-4% by contacts operated by the respective channel's vide range nuclear instrumentation. During 1990, procedure changes were made to functional testing procedures to include alarm and interlock testing. The vide range fur.ctional test procedure SP-2401B vas modified to include testing the operation of the relay contacts that serve'to remove the bypass the CVP generated from the . TM/LP pretrips and checked the proper operation of the TM/LP CVP as part of this testing process. This test is a start-up surveillance and was performed on November 4, 1990 prior to entering Mode 2. The CVP circuitry vas therefore tested and operable. The TM/LP functional = test procedure SP-240lJ vas also modified to test the ~ CVP feature. The method used to verify CVP operation only checked the generation of a CVP alarm after the bypass generated from the vide range nuclear instrumentation had been cleared. The procedure did not include the ability to perform the CVP check prior to actual plant conditions reaching greater than 10E-4% power. No violation of the Unit 2 Technical Specifications occurred.
Response
The statements of Issue 2 are not accurate. The portions of SP-2401J that are required to demonstrate the operability of the TH/LP trip were performed prior to entry into mode 2, consistent with the requirements of the Technical Specifications.
Mr. E. C. Venringer, Chief a U. S. Nuclear Regulatory Commission A09188/Page 4-January 4, 1991 During 1990, Unit 2 I&C has taken a conservative approach to the requirement to test alarms. In some cases such as this one, this \ conservative approach to testing alarms has added confusion to which portions of the procedure must be completed to support the operability of , the channel. In addition, this conservative interpretation deviates from 'l the original design objective for the system to be fully testable with the l plant in any operating mode. Unit 2 I&C is-currently planning to reviev ,. the requirement for alarm and interlock testing. This review is intended to make clear the distinction between those activities in a surveillance procedure that are required to support technical specification system operability from those that are not. Additional procedure refinements are expected. Issue 3 During a past technical specification review of alarm functions (i.e., LER 90-01), it was identified that the alarm verification of control element assembly withdrawal prohibit was included in the vide range functional surveillance (SP-2401B), but not in TM/LP, High Power, and " Local Pover. Density Functional" surveillance tests. During recent procedure upgrades, this change was not included in these procedures, though it should have been. Please discuss the validity of the assertion.
Background
The information that was learned during the technical specification review included the fact that the same relay that provides the CVP alarm also has contacts that provide the logic for the input to the-rod control system to
- accomplish the CVP. These contacts are checked during the Vide Range Nuclear Instrument Functional -Test SP-2401B, but not in the TM/LP functional test SP-240lJ and high power functional test SP-2401F. Once the relay has been verified intact and functional through the accomplishment of SP-2401B, there is no need to further verify its function in SP-240lJ and SP-2401F. These latter procedures verify the TM/LP and High Power Channel vill actuate the relay.
Response
Control Element Assembly Vithd aval Prohibit (CVP) alarm is verified 'for TM/LP and High Pover functions in ILC procedures SP-2401J and SP-2401F, respectively. Local Power Density (LPD) has no CVP alarm associated with it. If the issue is related to the alarm, the alarm is being adequately tested with the existing procedures SP-2401B, SP-240lJ, and SP-2401F,
. ~ . - __ ..e. ,
Mr..E. C. Venzinger, Chief i U. S. Nuclear Regulatory Commission A09188/Page 5 January 4, 1991 After our reviev and evaluation, ve find that none of these issues taken ', either singularly or collectively present any indication of a compromise of nue' lear safety. Ve appreciate the opportunity to respond and explain the basis for our actions. Please contact members of my staff if there are any further questions on any of these matters. ' ' Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
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E. J. pfoczka l7 Senior Vice President cct V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 9 9 l
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, , , UNITED STATES f S NUCLEAR REGULATORY COMMISSION I
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Docket No. 50-336 a Northeast Nuclear Energy Company ATTN: Mr. E. J. Mroczka ' Senior Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Mroczka:
The U.S. Nuclear Regulatory Commission recently received an allegation concerning activities at Millstone Nuclear Power Station Unit 2 (RI-90-A-0206). Details of this allegation are enclosed for your review and followup. We request that the results of your review and disposition of this matter be submitted to Region I within 30 dayc of the date of this letter. We request that your response contain no personal privacy, proprietary, or safeguards information so it can be released to the public and placed in the NRC Public Document Room. If necessary, such information shall be contained in a separate attachment which will be withheld from public disclosure. The affidavit required by 10 CFR 2.790.(b) must accompany your response if proprietary information is included. The enclosure to this letter should be controlled and distribution limited to personnel with a "need to know" until your investigation of the allegation has been completed and reviewed by NRC Region I. The enclosure to this letter is considered' Exempt from Public Disclosure in accordance with Title 10, Code of Federal Regulations, Part 2.790 (a). However, a copy of this letter, excluding the enclosure, will be placed in the i NRC Public Document Room. l l The response requested by this letter and the-accompanying enclosure are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. L l e ,e.. f s ,t y D t - i u. y)ogpfn=. ?gs
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i Your cooperation with us is appreciated. We will gladly discuss any questions you may have concerning this information. - Sincerely, R, Nbl f ' [ Edward C. Wenzinger, Chief f"ReactorProjectsBranch4
Enclosure:
Allegation Details (10 CFR 2.790(a) INFORMATION) cc: w/ enclosure W. Raymond, SRI Allegation File (2) cc: w/o enclosure PDR State of Connecticut
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The NRC Region I has completed its followup of the concerns that you brought to our our attention by way of the following memoranda; (1), March 1,1990, alleging NRC impropriety in not identifying your findings as problems and an example of a Mrming memorandum that you had received; (2), Ncrvember 4,1990, alleging procedural violations associated with the overhaul of service water pump P5C; and (3), November 11,1990, alleging procedural violations associated with the temporary repair of a moisture separator reheater manway gasket steam leak. With regard to issue (1), we referred your assertion of NRC impropriety to the NRC Inspector Gereeral. Because the inspector general is an independent office we cannot pmvide. you any conclusions in this matter. With regard to your receipt of a haraasing memorandum, the NRC does not have regulations regarding harassment from co workers. We remind you to address harassment issues such as this to the Department of Labor for their investigation and followup. With regard to issue (2), 'we referred your concems to the licensee for their review and followup. The licensee responded in a letter dated January 4,1991 (attached). Your allegation that beanng oil had not been removed prior to moving the pump and that : procedures goveming the pump overhaul require that the oil be removed is substantiated. We note that appropriate procedure revisions have been implemented to prevent similar occurrences. De NRC intends to c smplete action on this item in a future inspection report. We consider this matter closed. With regard to issue (3), we again referred your concems to the licensee for their followup and resolution. The licensee responded in a letter dated January 21,1991 (attached). The licensee determined that no procedures were violated in the activity described in the allegation. Based on the licensee's response, it appears that your allegation is unsubstantiated. He NRC plans no further action on this matter. Information in this record was deleted in actordance wit the Freedom of Information Act, exem tiorts __ n P E F01A DU e
4 1 , i We appreciate you infonning us of your concerns and feel that we have been responsive to those concems. If you have any additional questions or if I can be of further assistance, please call me collect at (215) 337 5225. v Edward Wenzinger, Chief Reactor Pmjects Branch 4 ( Attachments: As stated
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Allegation File RJ-90-A 0208, closecut Allegation File RI-90 A-0205, closecut Allegation File RI 88-A-0003, update only.(B22.1, closecut)
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January 21, 1991 Docket No. 50-336 A09166 i Hr. E. C. Venzinger, Chief Projects Branch No. 4 l ' Division of Reactor Projects . U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Venzingers ,
f' i
-Millstone Nuclear Power Stat on, Unit No. 2 i
! RI-90-A-0208 ve have completed our review of an Asallegation requested concerning activities letter in your transmittal at ! Hillstone Unit 2 (RI-90-A-0208).our response does not contain any personal privacy, dated November 27, 1990, The material contained'in this. proprietary, or safeguards information. response may be released to the public and placed in_the NRC Public Document' room at-your discretion._ The NRC letter and our response have received controlled and limited distribution on a "need to know" basis during 20,.the 1990, preparation of this' response. Based upon our request on December 21, 1991. Region I personnel extended the.due date for RI-90-A-0208 complete response. ( f$0 o5342/ AEv. 4 88
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e Mr. g. C.' Vensinger, Chief 3: U. S.' Nuclear Regulatory Commission < A09166/Page 2-January 21,'.1991 Issce Maintenance (veld repairs) was performed on *MSRs" on Saturday, November 10, i 1990. The velders objected to _ the exclusion of OSD as it was their opinion that the repairs were governed by the ASME Code. The velders were directed to ,._ proceed. Later OSD determined that ASME Section VIII was applicable. The Maintenance Supervisor ignored QSD, and the repairs continued to completion. < without an NCR or OSD involvement.
Background
the velding issue identified in this allegation concerns seal velding that was performed on the leaking manvay covers of_the moisture separator reheaters (MSRs). The MSRs are ASME Section VIII pressure vessels. The seal velds were applied to stop steam leakage from the gasket joint of the aanvays. The ASME Section VIII Code does not address in-service repairs to pressure vessels as a separate and distinct part of the Code. These typesThe of repairs reason are normally classified _ as temporary repairs by our procedures. ' these types of repairs are classified as temporary repairs is because the repair does not fully meet all Code requirements. Typically, this is due to operational restrictions (i.e., unisolable. vater or steam leakage which is D present while the repair is being made). While not addressed by the Code, these types of repairs are alloved by our procedures and'are consistent with the guidelines established for making ' similar repairs to ASME Class 3 systems that are governed by the Code In some requirements of the ASME Section II Repair Replacement Program. cases, these repairs require the processingInofother _ a nonconformance report cases,' repairs are (NCR) completed prior to the initiation of the repair. ; in accordance with a maintenance procedure'and an NCR is not required. Additionally, the velds in question are not structural-velds. No credit is taken for the structural-integrity of the veld. The veld was made to stop leakage of a gasketed manvay and does not have any adverse effect on the structural integrity'of the pressure vessel. Vith these points clarified, the following responses are provided to the-questions raised by this issue.- i
- 1. What vere the governing procedures for this work and vere they properly prepared and implemented?
Response
The procedures governing the velding performed on the MSR manvays were Haintenance Procedures MP-2701V and_VPS 033. J r
_.___ _ _.~._ _ _ . _ . .__ - . _ _ _ _ _ - _ _ . - - _ .___ .. _ ,_ f s Hr. E. C. Venzinger, Chief U. S. Nuclear Regulatory Commission A09166/Page 3 January 21, 1991 Maintenance frocedure HP-270lv provides instructions lot writing, ( processing, and close out of vorkInorder accordance packages with issued by the Millstone the guldslines Unit 2 Maintenance Departeent. , specified in this procedure, a two-pagePer work thisorder procedure, package nowas veldassembled and processed to authorize this work. inspection plan is required to implement veld repairs of this nature unless speciftrally required by an engineering evaluation or a noncon-formance report (NCR). In this case, the nature of the repair did not require a nonconformance report to be gentrated, and the engineering evaluation did not require the use of a veld inspection plan. The entire repair process was completed in accordance with the identified procedures.
- 2. Vere Code requirements required, included, and implemented?
Response
As previously stated, Section VIII of the ASHE Code does not address veld repairs to in-service vessels separately from construction requirements. There are no specific ASHE Code requirements applicable to this type of repair. Our procedures specified that the identified velds be made by qualified velders, using a qualified veld procedure specifically intended for use in this type of application. No credit is taken for the structural 4grity of the seal velds. The sole purpose of thece velds is to prei < p eakage. .
- 3. Vas OSD consulted and were their comments answeredt i
Response . The guidelines provided by Haintenance Procedure On this HP-2701Vbasis, no OSD do not require an inspection plan for this type of repair. involvement was identitled in order to complete the repair. OSD noted the veld repairs in progress and questioned the need for an NCR since the work vas being performed on an ASHE vessel. These come:nts were addressed by the Maintenance velding engineer. References were made to the guidance provided by Maintenance Procedure HP-2701V and the appli-cability of VPS 033. An NCR vas generated to document and disposition these concerns on Code applicability. Please explain.
- 4. Vere any procedures violated regarding this work item?
Response
Haintenance Procedure M'-2701V. Vork Order Processing, and ACP-0A-2.18, ASME Section XI Repair /Replacean t Program, were the procedures referenced to implement the subject repairs. The requirements and guidance provided
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' U. S. Nuclear Regulatory Commission A09166/Page 4 January 21, 1991 I No procedures were vloisted during the l by these procedures were followed.As a rescit of this issue, we are reviewing ourt ml
' repair process. , nance velding practices to ensure that they fully considered construction ' Code requirements prior to the initiation of work. If considered i necessary, the applicable procedures vill be revised to provide additional This guidance on when an NCR is needed prior to the initiation of work. review is expected to be completed by July 1991. . Aftet our review and evaluation, we find that this issue does not present any indication of a compromise of nuclear safety. Ve appreciate the opportunity to respond and explain the basis for our actions. Please contact menbars of my staff if there are any further questions on this satter. Very truly yourr, NORTHEAST WUCLEAR LNERGY COMPANY k H.# Za/ B. J. TNckk'a ~ ~ + Senior *'Vice President cc: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 f I i J 4 4 i d d
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Deceeber 21, 1990 Docket No. 50-336 A09166 Mr. E. C. Venzinger, Chief Projectr, Branch No. 4 information in this record wt; deleted Division of Peactor Projects U. S. Nuclear Regulatoty Commission 'riswrdance with th Fr t(Dm Cllnlormallon Regton I let, exoptions _ b - I - 475 Allendale Road IDi A- [I I ' d l-King of Prussia, Pennsylvania 19406
Dear Hr. Vensinger:
Hillstone Nuclear Power Station, Unit No. 2 RI-90-A-0204 and RI-90-A-0208 _ Ve have completed our review of an allegation concerning activities at M1113 tone Unit No. 2 (RI-90-A-0204). As requested in your transmittal letter, , our response does not contain any personal privacy, proprietary, or safeguards information. The material contained in this response may be released to the public and placed in the NRC Public Document room at your discretion. The tiRC letter and our <sponse have received controlled and limited distribution on a "need to know" baals during the preparation of this response. Based upon our request on December 20, 1990, Region I personnel extended the due date for RI-90-A-0208 to January 21, 1991. Additfonal time was needed to resolse technical issues in order to provide a complete response. RI-90-A-0204 Issue 1,
Background
These same issues vere raised by an instrument specialist during the head cabling vork activity. There is a close correlation of the questions posed by tha memo and response. This documentation is available for your staff's reviev. 4
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b P Mr. E. C. Venzinger, Chief U, S. ": icer Pegulatory Commission A0916 P ' :p i - Decet 't, i 1990 The folloving problems exist with IC 24210
- 1. Step 5.5.6.1 calls for visually inspecting the connector assemblies on each cable for signs of degradation or damage. Problems have been identified with Z-1, 64, 97, and Z-2, 48. The cables were put together anyway. Please explain. Further, the step calls for visually inspecting the Grafoil gasket at the Litton.Veam r.onnector, but no information is ,
provided about these connectors or the Crafoil gaskets. Are I&C ' technicians trained on these items? Does the procedure require upgrades to expInin what the technician is looking for in'this step? No figure is provided with the procedure to identify the connectors, and it is impossible to read the etched numbers on the connectors. How are the connectors to be identified on the job 7 Response . The activity of reassembling the HJTC connectors was begun on a Sunday with an upgraded specialist serving as the first line supervisor and no engineering support readily available. During the work activity, several luestions vere raised and documented. The following day, Honday, these j questions vere investigated and resolved by more knowledgeable I&C and Engineering personnel. l i l Connector training performed to date has not included the specific knowledge needed to perform this task. Trainitg requirements for this l l task are being reviewed for additinn to the I&C technical training program. As vritten, the procedure makes the assumption the I&C technician using it ' has the knowledge necessary to determine the gasket location and t condition. As a result of this experience, the procedure ovner has been '
- instructed to revise the procedure to include the necessary informat.fon.
The connections are identified by a small stenciled number. They are also staggered and are in the order of #1 being the highest and #8 being the lovest. As lighting is limited and access is difficult, the normal method for connector identification has been its order of elevation.
- 2. The caution on page 14 is impossible to achieve as only about a 45-degree turn to possible. Is this technically satisf actory? Does this indicate ~
procedural non-compliance during past performance? Has there been
- repeated connection damage in complying with this step
Response
On the following Honday, additional technical information was obtained from the responsible vendor. Procedure' changes vere implemented to allov an alternative method of torquing the connector. This method was used to ensure the proper assembly of the connectors. No significant connector damage has occurred. r l l
Hr. E. C. Venzinger, Chief , U. S. Nuclear Regulatory Commission A09166/Fage 3 . Deceaber 21, 1990
- 3. Dust cover caps are shown in figures in IC-2421C and 10-2419C. These caps are not being used. Are dust covers needed? Why are they not used if shown in the figure?
Response
The use of dust covers is not delineated by the procedure. Dust covers only appear in the figure. The need for their use is currently being evaluated. The existing figure has been replicated from a vendor draving that shovs the "as supplied" HJTC orobe. The figure was not intended to dictate the use of dust caps in the field. The figure also has shortcomings in not supplying the necessary decall needed for inspection and assembly of the connectors.
- 4. Step 5.6.4 calls for the verification of the !!JTC probes per IC-2419C.
Vhy is thic not done prior to the connection of the detector cables?
Response
This HJTC probe had been previously used. There was no need to perform the verification of the known good probe. The retest of the overall system response verifies the condition of the probe, as well as the entire system. The step has since been removed from the procedure.
- 5. Are any generic problems with procedural non-compliance or laxity with regards to procedural adherence evidenced by these problems? Please explain.
Response
There are no generic problems with procedure non-compliance as evidenced , by these problems. Wote that this procedure has not been upgraded as part of the station proce.Nre upgrade program. Veaknesses in such procedures like this one have pre viously been discussed with the NRC staf f. This ' exampic serves to demonstrate that with the proper attention to detail by the technicians involsad and with the proper response and direction from . their supervision that a proper quality activity results. Vhere appropriate, the necessary dCRs and procedure changes were implemented, and the noted need for procedure enhancements took place. , Issue 2 Instrument Calibration Review ICR 90-113 vas written on November 2, 1990. Please provide the resolution documentation for the ICR.
Response
A copy of the Instrument Calibration Review ICR 90-113 is attached.
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Mr. E. C. Venzinger, Chief U. S. Iluclear Regulatory Commission A09166/Page 4 December 21, 1990 Issue 3 Recently, an ICR vas generated concerning out-of-specification test voltages found during the performance of SP-2404C. Please provide the resolution documentation for the ICR. Historically, the reference voltages have been out of specifiention: Harch 9, 1990 AVO M2-90-02559 May 5, 1990 AVO M2-90-02736 May 5, 1990 AVO H2-90-05237 May S. 1990 AVO H2-90-03480 Augu9t 30, 1990 Surveillance, SP-2401F, and SP-2401B Please discuas the operability of Channel "C" of the RPS vith the historic out-of-specification test voltages. Note, Dravings 25203-39069, sheet 40, and I 25203-25193 sheet 6, supposedly identify the reference test voltage application >.
Response
There is no I&C prncedure SP-2404C. Therefore, it is not possible to supply any information concerning an ICR generated while performing this procedure. The AV0s referenced vere written against RPS channel "C" Reactor Protection System Calibration and Indication Panel (RPSCIP). There are no test voltages gener ated in this int.trument drawer. However, the RPSCIP digital volt meter (DVM) is used to monitor the core protection calculator (CPC) logic power supply output voltage (+/-10vde). A review of surveillance and equipment history shows that following replacement of a different pover supply (+5vde) in May 1989 by PDCR MP2-89-028, the RPSCIP DVM indicated a .003vde offset from the actual +/-10vde power supply output. In'restigation during the 1990 ref ueling shutdown, while performing PDCR HP2-89-068, revealed a difference in ground potential between the RPSCIP and the CPC. This caused the RPSCIP DVH to indicate incorrectly by .003ved. This is at the limit of the tolerance applied to the DVM indication (+/ .003vde). A design change notice (DCN
#DH2-P-022-00) was vritten against PDCR HP2-89-068 and implemented to correct the problem. As can be seen by reviewing the referenced AU0s, the problem was identified shurtly after the +5vde modification and determined to be limited l to the DVH indication only. An AVO to investigate the problem vas vritten at that time and scheduled for the refuel.
Also referenced are two surveillance procedures for August 30, 1990. One surveillance concerns the CPC, SP-2401F, the RPS high pover trip test, which is pexformed monthly. It documents the above DVM indication problem and shows the high power trip function of the CPC to be operable at this time. The other referenced surveillance SP-2401B, concerns the vide range flux monitor functional. This procedure is performed veekly while shut down. Since Millstone Unit No. 2 vas not shut down during the referenced time. It is difficult to determine hov this is related to the issue.
\
- Mr. E. C. Venzinger, Chief U. S. Nuclear Regulatory Commission A09166/Page 5 December 21, 1990 Two drawings are referenced which " supposedly identify the reference test voltage applications". The first drnving, 25203-39069, sheet 40, is an electrical schematic of the CPC. It shova the power supplies connection to the circuit but makes no reference to the application of the voltage with respect to absolute values. It has no connection to the vide range flux monitor circuit. The second drawing referenced, 25203-25193, sheet 6, does not exist in the NUSCO drawing system. More specific information will be required to resolve this issue.
Issue 4 A red tag was improperly hung by Operations on the back of C05/C06. The tag was hung on TDE but should have been hung on TDD. SC0(*) knows the details. Please explain the tagout problem and actions taken to both correct the problem and to prevent recurrence. (*) - identity may be obtained from the SRI. Respon_Se The pins for Veldmuller Block TOD, on the back of C05/C06, were properly removed and placed in their correct storage location at the time of tagging, under the guidance of an 16C technician. The red tag was properly filled out for Veldmuller Block TDD. The tagging discrepancy was that the red tog was not attached directly to Veidmuller Block TDD but to a cable directly beside TDD which was labeled TDE. This tagging dir,crepancy was noticed by an operator who consulted with on I&C technician. The operator then moved the tag from the table to the pins for Veidmuller Block TDD. Long-term actions to prevent recurrence is to formally distribute to all Unit 2 Operations personnel a standardized method for tagging Veldmuller Blocks. Issue 5 A recent annunciator vindov change C04(CRDR) was apparently not reviewed and completed properly as procedures SP-2401B, 2401F. and 2401J had to be changed during the performance to accommodate the annunciator vindow changes. Please discuss the accuracy of the statement and any actions that you have taken or may take to rectify sny identified problems.
Background
I As part of the CRDR improvements, several annunciator vindows vere relocated. l This activity took place over the course of the 1990 refueling outage. Changes to 160 procedures were appropriate as the procedures contained the l detail of the annunciator vindov location. Changes to SP-24018 and its data sheet and SP-240lJ vere processed at the time the procedures vare implemented. , No changes were required to SP-2401F. It is necessary for p:ocedure changes required by a modification to be approved prior to the implementation of the procedure. This was done in this case.
1
' Mr. B. C. Venzinger, Chief U. S. Nuclear Regulatory Commission ;
A09166/Page 6 ! December 21, 1990
Response
Although several changes vere required, no problems requiring corrective action exist, All modification related procedure changes can not be made in advance of the task being assigned. Issue 6 On November 5,1990, upon completion of the CVP por tion of SP-2401J, (the CRDR ' change van done as a non-intent change on this date), Channel C 70-7 was left bypassed by technicians after the surveillance vas turned in, and the technicians had to be called by the SS and the channel unbypassed by the SCO. Please discuss the procedural compliance aspects of this statement.
Response
This event is an example of procedural noncompliance. SP-240lJ requires restoration of tha bypass key. This sas not done prior to turning in the surveillance. The appropriate personnel have been counseled with respect to this issue. Af ter our reviev and evaluation, ve find that none of these issues taken either singularly or collectively present any indication of a compromise of nuclear safety, Ve appreciate the opportunity to respond and explain the basis for our actions. Please contact members of my staff if there are any , further questions on any of these matters. Very truly yours, . I NORTHEAST NUCLEAR ENERGY COMPANY
/W E. J.~p oczka- 4/
Senio/ Vice-President cc: V. J. Raymond, Senior Resident inspector, Millstone Unit Nos. 1, 2, and 3 e,,g.- w-------w,+ ,,.,-+.y,w- -,,vwr- ( -, y-c,9c++g g- $,i g y g 9 y -eg-,aes w es r- y-s+rp-y y
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_ . . _ _ . _ . . . _ _ . _ . _ _ __. _ _ ~ _ _ _ _ . _ .m _ - _ _ > _ .._..._ _ k ,g ; FORfA APPROVED $6M2]MMATE d-8~9 b MTG.NO. E h,0 - 6 6 ,, INS 'RUMENT CAllBRATION REVIEW FORM Part 1 To be completed by person performing work. Dato _!)/ E /9O_ Time ,L$! % ICR Number _Qf-//] AWO#IMh-9o~]W275' i. i , instrurnent or Device Affected 10# N--Id3 LI _,Namef&2chting e %s,pe., O V Procedure Number:.TgP '2.41[rd
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hvog$.m *fp r ens,* EchxqMad, . o Description of Event or Calibration Results ]dd,$tuldelie#/4E de fop _jos.,, ,_ OM d _, ddL. EO _ 44 e _ L. M kAbtl7o_.__ktv0Sehdk2_W1. LeM ' PGArmu>,_%tm? PTlGoa uw GuuA 1N6 fin.4.nA SI- Prekam, 4 Cause: (11 Facts are Known) ( l Instrument Orlft [ ] Unknown , y Othcr Explain;yaFT {%$Q.f1._&n. F 010tA hk A/ak6_1_f}/jk/ t hijfjjy 1 Would the hstrument or Control System have performed its function as required by Tech Specs? Yes [ } No [>d N/A [ ] 1 Dasis: 3Cn/Jr[wM-
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Modo ~B Power o% Temp. Seo*F Press. 400
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- NOV 1 gy s !
Docket No. 50 336 ! 1his refers to your memorandums recently received by our office in which you expressed concerns related to the operation of the Millstone facility by your employer, Northeast Nuclear Energy Company. We have initiated actions to examine your concerns and upon completion ot'our actions, will inform you of our findings. We must inform you that confidentiality is 001 granted as a routine matter and therefom your requests for confidentiality are denied. I assure you that we will at:empt to conceal your identity , while resolving these matters but licensees can and sometimes do surmise the identity of individuals who provide information to us because of the ratture of the information or other factors beyond our control, in such cases, our policy is to neither confirm nor deny the licensee's assumption. Should you have any additional questions, or if I can be of any further assistance, please call me collect at (215) 337-5120. z waM k Donald Haverkamp, Chief Division of Reactor Projects, Section 4B cc: W. Raymond, SRI InfMasti?n in this ' ccord was de!cted Allegation File in acmante with thf fregom of Infornation s_ Act, er.emptiopkl FO! A $ '-
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'i REGION I Report Nos. $d45/9102 t 50 336/91 03 10 423/91 03 Docket Nos. EQ ?45 10:Us -
30 n3 License Nos. DPR 21 Category: _C DPR 65 C NPF 49 C Licensee
- Em.thyost t Nuclear Enerev Comna.py P. O. Box 270 HartfgidlGDnecticut 06101 facility Name: Millstone Nuclear QIneratine Station. Units 1. 2 and 3 Inspecnon At: Waterford. f onnectitul Inspection Conducted: Eebruatv 18 - 2;Ef!1 Inspectc s: I L hh* A .
!$l'!Ih R. L. Nimitz, CHP, Senior Radiation Specialist Date Approved by; a 4 d$ 3! k/7/
W. J. Pasciak, Chief, Facilities Radiation -Date . 4" s Protection Section . laspeelinn_Summarr NRC Inspection on February 18 22,1991 (NRC Combined Inspection Report Nos. 50 245/9102; 50 336/9103; and 50-423/9103). Areas Revieweda This inspection was a routine unannounced radiological controls l inspection. Areas reviewed were: the licensee's action on previous inspection findings, organization and staffing, training and qualification, external and internal exposure control, radioactive and contaminated material control, ALARA, and process and area radiation monitor calibration and surveillance testing. Results: No vichtlons were identified. The licensee implemented good radiological controls for the Unit 3 outage, Apparent weaknesses were identified in the area of L industrial safety. j L 9j()337063% ~ hlCQ,
.0 13 i i The review was with respect to applicable criteria specified in Technical Specifications. The evaluation of licensee performance in this area was based on review of calibration and surveillance testing documentation and discussions with cognizant personnel.
Within the scope of the resiew, no siolations were identified.
- 10. Instrument Souret Checking During tours of the Unit 3 contamment, the inspector noted that the licensee had established a respirator and instrument issue check station.
PersonnH at the check station performed the following:
. issued respirators to properiy qualliied personnel
. issued digital alarming dosimeters to personnel requiring their use issued survey meters to personnel requesting them performed source checks of radiation measuring instruments The inspector noted that the licensee used several sources to check various instruments at the check station. These were a 20 mdlicurie Cs D7 source used for checking the mtegrating alarming dosimeters, a 2 millicurie St 90 source used .
for checking RO 2siand four Techmtium 99 sources for checking friskers. These sources permitted the licensee to check essentially all ranges of instrurnents issued. The use of the sources was considered a good licensee iniiative to verify operability of radiation monitoring or survey instruments. The inspector's observation indicated personnel did not wear extremity dosimetets wben u:,ing the various sources. _ The inspector noted that the RO 2/RO 2A source ' checking desice (2 milli:urie St 90) produced a dose rate of an indicated 89-Rads /hr at 2 inches from the source and 1.3 Rads /hr at 18 inches. The inspector's review of the RO 2/RO 2A source check device indi:ated the following:
- The source check device was constructed using an_ approved radiation work permit by properly qualified personnel.
. The indisidual who constructed the 2 millicurie checking device wore appropriate personnel monitoring devices for the extremity. .
- The individual cor <tructing the 2 millicurie RO 2 checking device used--
long handled twet a when handling the source. No significant exposure was sustained constructing the device. l h
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- The device was placed in senice in mid January 1991
- As discussed in section 4, personnel using the device received appropriate t training to use the desice.
- A draft instruction was in place for use of the device. The instruction was subsequently approved on February 22,1991.
- The device has never malfunctioned or stuck in the open position causing unnecessary personnel or unmonitored exposure.
- The inspector noted that the licensee had issued extremity dosimetry to one indhidual who was using the source checking devices to evaluate the extremity dose received. The licensee's preliminary review indicated the dose to the extremity would be minimal and would not require issuance of extremity dosimetty.
The inspector noted that personnel could put their hands in an apparent maximum radiation field of 28 mrad /hr when using the RO 2 and RO 2A source checking d te and 40 mR/hr when using the digital alarming dosimeter check source. The licensee indicated an evaluation of expected extremity dose was undenvay and would be documented. Preliminary inspector review indicated expected extremity dose of personnel performing the instrument checks would not require issuance of extremity monitoring devices. This item is unresolved (50 , 245/91 02 01).
- 10. Industrial Safetv During tours of the station, the following observations in the area of industrial safety were made:
- On February 18,1991, at about 4:00 p.m., the inspector observed personnel working in close proximity to the edge of the dry Unit 3 refueling cavity.
The individuals were not using safety belts. Also, the open upender pit area did not have any railing around it or safety barricading to preclude personnel from falling into it. The licensee immediately required personnel to use safety lines The licensee also stationed a guard at the open pit area to preclude personnel from falling into the pit. "he licensee also informed personnel of the need to wear required safety equipment as specified in station procedures. ( Personnel were informed at the morning outage meeting on February 19, 1991.
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hir. Edward hiroczka Senior Vice President , Nuclear Engineering and Operations Northeast Nuclear Energy Company P.O. Box 270 Ilartford CT. 06141 0270 Dear hir. hiroczka; The U.S. Nuclear Regulatory Commission recently received a number of allegations concerning activities at hitlistone 2. Details of these issues are enclosed for your review and followup. We request that the irsults of your review and disposition of these matters be submitted to Region I within 30 days of receipt of this correspondence. We request that your response contain no personal privacy, proprietary, or safeguards information so that it can be released to the public and placed in the NRC Public Document Room. If necessary, such information to be withheld shall be contained in a separate correspondence and the affidavit required by 10CFR 2.790 must accompany your response if proprietary or like information is included. The response requested by this letter and the accompanying enclosures are not subject to the clearance procedures of the Office of hianagement and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. Your cooperation with us is appreciated. Please address any questions that you may have , regarding these issues to hir. Scott Stewart at (215) 337-5232, or hir. Donald IIaverkamp at (215) 337-5120. Sincerely, Edward C. Wenzinger. Chief Reactor Projects Branch 4 Enclosure 1, Allegation RI-90-A-0208, Enclosure 2, Allegation RI 90-A-0204, cc w/ encl: W. Raymond, SRI QV J ' Qf ~, -
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Enclosure 1 RI-90 A4208 l page 1of1 t-Issue: hiaintenance (weld repairs) was performed on "htSRs" on Saturday, November 10. -The - welders objected to the exclusion of "Q.S.D." as it was their opinion that the repairs were govemed by the AShiE Code. The welders were directed to proceed. Later, QSD determined that AShfE Section 8 was applicable. The maintenance supervisor ignored QSD and the repairs continued to completion without an NCR or QSD involvement.
- 1. What were the governing procedures for this work and were they properly prepared and implemented? Were code requirements required, included, and implemented? Was QSD consulted and were their comments answered? Were any procedures violated regarding this work item; Please explain.
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s Issue 1. The following problems exist with IC 2421C: , j
- 1. Step 5.5.6.1 calls for visually inspecting the connector assemblies on each cable for signs of degradation or damage. Problems have been identified with Z-1, #4,#7, and Z-2,#8. The cables were put together anyway. Please explain. Further, the step calls for visually inspecting the Grafoil gasket at the Litton Veam connector, but no information is provided about these connectors or the grafoil gaskets. Are I&C technicians trained on these items? Does the procedure require upgrades to explain what the technician is looking for in this step? No figure is provided with the procedure to identify the connectors and it is impossible to read the etched .
numbers on the connectors. How are the connectors to be identified on the job 7
- 2. The caution on page 14 is impossible to achieve as only about a 45 degree turn is possible, is this technically satisfactory? Does this indicate procedural non-compliance during past performance? Has there been repeated connection damage in complying with this step?
1
- 3. Dust cover caps are shown in Figures in IC-2421C and IC 2419C. These caps are not being used. Are dust covers needed? Why r.re they not used if shown in the Figure?
- 4. Step 5.6.4 calls for the verification of the HJTC probes per IC-2419C. Why is this not done prior to the connection of the detector cables?
- 5. Are any generic problems with procedural non-compliance or laxity with regards to procedural adherence evidenced by these problems 7 Please explain. . .
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page 2 of 3 l t : Issue 2: Instrument Calibration Review ICR 90-113 was written on-112 90. Please , provide he resolution documentation for the ICR.
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Issue 3: Recently, an ICR was generatd conceming out-of specification test voltages found during the performance of SP-2404C. Please provide the resolution documentation for the ICR. - Historically, the reference voltages have been out-of-specification: 3/9/90 AWO M2-90-02559 . 5/5/90 AWO M2 90-02736 : 5/5/90 AWO M2-90-05237 i 5/5/90 AWO M2 90-05480 8/30/90 Surveillance, SP 2401F and SP-2401B Please discuss the operability of Channel "C" of the RPS with the historic out of specification t.est voltages, Notei Drawings 25203 39069 Sheet 40, and 25203-25193 Sheet 6, supposedly identify the reference test voltage applications. t h i A i y- ,,,. 4,-- .ryr.y.,-,e., , m._-.., , . , - p.y., -,y,wa.-_, ,,.y,,. , , - ...er-y. .,, ,, . + -,.._.,,.%- - , m.-_,. _,.. - ,.-..-.,._c,
i Enclosure 2 RI 90-A-0204 page 3 of 3 Issue 4: A red tag was improperly hung by operations on the back of COS/CO6. The tag was , hung on TDE but should have been hung on TDD. SCO( * ) knows the details. Please explain the tagout problem and actions taken to both correct the problem and to prevent , recurrence. ( * ) identity may be obtained from the SRI. Issue 5: A recent annunciator window change C04(CRDR) was apparently not reviewed and completed properly as procedures SP 2401B, 2401F, 240lJ had to be changed during the performance to accommodate the annunciator window changes. Please discuss the accuracy of the statement, and any actions that you have taken or may take to rectify any identified problems. Issue 6: On i1/5/90, on completion of the CWP portion of SP-2401J,(the CRDR change was done as a non intent change on this date), Channel C TO-7 was left bypassed by technicians after the surveillance was turned in and the technicians had to be called by the SS and the channel unbypassed by the SCO. Please discuss the procedural compliance aspects of this statement. , l l 1 l l l-
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. JAN 141991 Docket No. 50-336 File No. Al-90 A-0192, RI-90-A-1BB c' Oear {.
Subject:
Allegations Concerning (1) Tagout for Work on Pressurizer Relief i Valves (2) Oil sampling, and (3) Loss of Containment Integrity at Hillstone Unit 2 l The NRC Region I office has completed its followup in response to the concerns you brought to our attention on October 12 and 19, 1990, alleging that (1) tagout procedures were not followed for work on the pressurizer power operated relief valves 2-RC-402 and 404, (2) potentially contaminated EOG oil was released unmonitored, and (3) containment integrity was not maintained during fuel movement. We found your allegation concerning tagout procedures to be unsubstantiated and have documented our findings in NRC Inspection Report No. 50-336/90-22 section 3.6. Ve concluded that the exercise of discretion concerning hanging of tags on local valve operators was appropriately implemented considering the high radiation area location of the subject isolation valves. This discretion is permitted by ACP-QA-2.06A. We also found your allegation concerning improper sampling of EDG oil to be unsubstantiated and have documented our finding in NRC inspection report 50-336/90-22 section 4.2. We concluded that licensee action to prevent release of potentially contaminated waste oil was adequate. Finally your allegation concerning loss of containment integrity is the subject of ongoing NRC enforcement action as documented in NRC inspection report 50-336/90-22. The final resolution of this item will be documented in future NRC correspondence with the licensee. We appreciate you informing us of your concerns and feel that our actions in this matter have been responsive to those concerns. Should you have any addit'ional questions, or if I can be of further assistance in this matter, please call me collect at (215) 337-5120. Copies of the above noted report sections are attached for your information. Sincerely, Information in this record was dt!tted in accordance withi e f 1 -' of Information y(( [ Act.exem ions Donald R. Haverkamp, lef FOR- 4 I ~ Si ' Reactor Projects Section 4A Division of Reactor Projects Enclosures As Stated bec w/ enc 1: J. Stewart, DRP (2) t M. perkins, ORMA ORC h gazzaoo(b
f . (( 3.l' 't 1991 Docket No. 50-336 File No. RI-90-A-0192, RI-90-A-188 In1ormation in th' pf Dear
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Subject:
Allegations Concerning (1) Tagout for Work on Pressurizer Relief Valves, (2) Oil Sampling, and (3) Loss of Containment Integrity at Hillstone Unit 2 The NRC Region I office has completed its followup in response to the concerns you brought to our attention on October 12 and 19, 1990, alleging that (1) tagout procedures were not followed for work on the pressurizer power operated relief valves 2-RC-402 and 404, (2) potentially contaminated EDG oil was released unmonitored, and (3) containment integrity was not maintained during fuel movement. We found your allegation concerning tagout procedures to be unsubstantiated and have documented our findings in NRC Inspection Report No. 50-336/90-22 section 3.6. We concluded that the exercise of discretion concerning hanging
- of tags on local valve operators was appropriately implemented considering the high radiation area location of the subject isolation valves. This discretion is permitted by ACP-QA-2.06A.
We also found your allegation concerning improper sampling of EDG oil to be unsubstantiated and have documented our finding in HRC inspection report < 50-336/90-22 section 4.2. We concluded that licensee action to prevent release of potentially contaminated waste oil was adequate. Finally your allegation concerning loss of containment integrity is the subject of ongoing NRC enfo' cement action as documented in NRC inspection report 50-336/90-22. The n nal resolution of this item will be documented in future NRC correspondence with the licensee. We appreciate you informing us of your concerns and feel _ that our actions in this matter have been responsive to those concerns. Should you have any additional questions, or if I can be of further assistance in this matter, please call me collect at (215) 337-5120. Copies of the above noted report sections are attached for your information. Sincerely, Original Signed By Donald R. Haverkamp, Chief Reactor Projects Section 4A Division of Reactor Projects Enclosures As Stated 0FFICIAL RECORD COPY ALLEGATION COL RI-90-A-0192 - 0001.0.0 MMMdh[ L
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The Region I office has completed its followup to the concems that you brought to our attention on the dates described. Relevant documentation such as letters from the licens attached to this letter. A synopsis of these concems and our subsequent actions and regulatory conclusions are detailed below. On October 8, and October 11,1990, you provided to us a number of concerns associated with Wide Range Nuclear Instrument operability. You further discussed this issue with me in January,1991. Additionally you provided to us on the same dates, a concern that you were not consulted during a recent procedure change associated with surveillance procedure SP-2417H, and that this omission was contrary to station procedures. We provided these issues to your employer in a letter dated October 26,1990 and they responded in a letter dated December 21,1990 (attached). Additionally, we inspected the issue of wide range nuclear instrument operability and provided you the results of our investigation in a letter dated January 14, 1991. Your assenions that spiking had occurred on the channel "A" of the instrument were true, but inoperability and violations of technical specifications have not been substantiated. With . t regard to SP 2417H, the licensee admitted that you were not consulted for the procedure ! change, but no violation of proecdures occurred and your allegation appears to be unsubstantiated. Furthermore, the procedure change was determined to be adequate. The NRC is satisfied that the licensee addressed your concems, and we plan no further action in these matters. On August 8,1990, you provided us with concems assertig that (1) work associated with the overhaul of the containment radiation monitor was improperly controlled and (2) a bypass jumper tag was improperly controlled during the maintenance of work order M2 90-08033. We provided these concems in a letter to the licensee dated October 2,1990 and the licensee responded in a letter dated December 3,1990 (attached). We note from the licensee response that some problems were identified in the control of work associated with the radiation monitor, but it appears from the licensee's assessment that at least one monitor was operable during the times in question in your assertion. We note that the licensee identified the problems described in your concern and took proper corrective actions. Further the licensee is implementing additional controls to establish better w g gg,g L.i bO/
f t 2 ; coordination of activities between operations and maintenance. Your allegation that there were problems associated with the control of maintenance is substantiated; however, the NRC s considers the problems minor with respect to nuclear safety and notes that appropriate corrective actions have been taken. With regard to issue (2), the pin connector in question was not connected to anything, therefore a bypass tag or a work order would not be < appropriate. Your allegation may be true, but is of no consequence. In any case, we are satisned with the licensee's response and plan no further actions in these matters. On November 9, and November 11,1990, you provided to us (1) a description of events associated with the main turbine stator cooling troubleshooting, (2) a question as to the propriety of completing the thermal margin low pressure surveillance in Modes 3 and 2, and, (3) a question regarding the testing of the control element withdrawal prohibit alarm. We provided these issues to your employer in a letter dated December 6,1990, and received their response in a letter dated January 4,1991 (attached). Your assertions regarding issue (1) were determined to be true, but appear to be of no consequence to nuclear safety. The problems had previously been identined by the licensee and appropriate corrective actions appear to have been taken. With regard to issue (2), it appears no violation of technical specifications occurred, and your allegation appears to be unsubstantiated. With regard to issue (3), the alarm testing was determined to be adequate and your allegation appears to be unsubstantiated. We are satisned that the licensee answered your concerns and we plan no further action in these matters. On September 28,1990, and in several discussions with NRC personnel during the recent refueling outage, you asserted that there were violations of the overtime policy at Millstone Unit 2. We investigated your assertion and discussed the issue with unit management, but could not substantiate your claim. To enable further evaluation, more specinc details are : needed. We note that you have recently provided us a similar concern that overtime restrictions may not have been complied with, and we are inspecting this concern. We will inform you of the results of our inspections when complete. On September 28,1990, and in a November 5,1990, memorandum to our resident inspector, I you provided the NRC with seven concerns involving: (1) the procedural adequacy and implementation of IC 2419C section 5.5.6 which involve the heated junction thermocouple ! inspection; (2) the instrument calibration review that you initiated; (3) the instrument ! calibration review that had been initiated associated with test voltages being out-of-l specification; (4) a red tag that was improperly hung on the Weidmuller Block TDD; (5) an L assertion that an annunciator window change had been improperly handled by the licensee; I (6) an assertion that surveillance procedure SP-240lJ had not been implemented when an instniment and controls technician turned in the paperwork without restoring the bypass key; and, (7) an assertion that you had received harassing mail from a co-worker,
?
t 3 We referred issues 1 thru 6 to the licensee in a letter dated November 11,1990 and received t their response in a letter dated December 21,1990 (attached). We note that the licensee was aware of the discrepancies that you identified in your assertions and had taken actions to ' correct the deficient conditions when originally identified. In regard to issue (1), the procedure was written for a skilled instrument and controls technician with experience in this type of maintenancel however, you made no assertion that maintenance was improperly or incompletely performed to the extent that operability of essential equipment was affected in any case, your allegations appear to be substantiated but of minor significance with respect to nuclear or personnel safety. With regard to issue (2), the licensee provided us a copy of the instrument calibration review (ICR) and we are providing this copy for your review. Please inform us if you have further questions regarding this matter. With regard to issue (3), the licensee had taken action to address the problem that you described and we have not been informed of any inoperabilities that resulted from the corrective actions. With regard to the drawing concern of issue (3), more information is needed to adequately address your conccrns. Please inform us of any additional details or further questions that you may have in this matter. With regard to issue (4), a minor tagging discrepancy had been previously identified and was promptly corrected. Further, guidance on the tagging of Weidmuller blocks was to have been provided to operations personnel. Your allegation in this case appears to be substantiated, but is of little concern with respect to nuclear safety. With regard to issue (5), no problems were identified, and no corrective actions were warranted. With regard to issue (6), we note that , operations personnel identified. the condition that you asserted and took prompt actions to restore the channel. The technician in this case appears not to have exceeded Unit 2 Technical Specification limitations. However, your allegation regarding implementation of the surveillance procedure was substantiated by the licensee. We are satisfied with the licensee responses to these six issues and plan no further action in these matters. Finally, with regard to issue (7), the NRC cannot take action based on co worker harassment, especially if the alleged harassment is anonymous. If you feel that you are being harassed by your employer, we again remind you to take these issues to the Department of Labor. On September 14,1990, you provided us a three concerns detailing: (1) failure of operators to note that the steam jet air ejector may not be working properly; (2) the failure of your supervisor to respond to one of your questions; and, (3) the failure of another instrument an t controls technician to follow an unspecified department instruction requiring that a trouble report sticker be attached to an instrument after a problem was identified. We provided your concems to the licensee in a letter dated October 4,1990, and they responded in a letter dated December 7,1990 (attached). With regard to issue (1), the licensee identified no inoperability associated with the radiation monitor, but has identified that upgrade of the system is warranted. The licensee plans to replace the monitor in 1991. Your concerns therefore have some validity. With regard to
/
3. Issue (2), communication between you and your supervisor appears to have been either t misunderstood or incomplete. We could not determine the validity of your complaint. With regard to issue (3), the licensee determined that the orange sticker was properly pla operator information and that work was properly controlled by the applicable work ' documents. Your concerns in this case appear to be unsubstantiated. We are satisfied wit the licensee response to the concerns as presented, and the NRC plans no further ac regard to these matters. We appreciate you informing us of your concents and feel we have been responsive concems. Should you have any additional questions or if I can be of further assistance, please call me collect at (215) 337-3225. ! Sincerely; M >
,f j Edward C. Wenzinger, Chief ,
Reactor Projects Branch 4 Attachments: As stated j l bec:wlo enclosures M. Moore DRMA (6) q{ q) g ggo.9,gAg,y
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I mann oca,% F0lAh. - I( f\-!c The Region I office has completed its followup to the concerns that you brought to our attention on the dates described. Relevant documentation such as letters from the licensee are l attached to this letter. A synopsis of these concerns and our subsequent actions and regulatory conclusions are detaDed below. On October 8, and October 11,1990, you provided to us a number of concerns associated with Wide Range Nuclear Instrument operability. You further discussed this issue with me in January,1991. Additionally you provided to us on the same dates, a concern that you were not consuhed during a recent procedure change associated with surveillance pmcedure SP-2417H, and that this omission was contrary to station procedures. We provided these issues to your employer in a letter dated October 26,1990 and they responded in a letter dated December 21,1990 (attached). Additionally, we inspected the issue of wide range nuclear instrument operability and provided you the results of our investigation in a letter dated January 14, 1991. Your assertions that spiking had occurred on the channel 'A" of the instrument were true, but inoperability and violations of technical specifications have not been substantiated. With regard to SP-2417H, the licensee admitted that you were not consulted for the procedure : change, but no violation of procedures occurred and your allegation appears to be unsubstantiated. Furthermore, the procedure change was determined to be adequate. The NRC is satisfied that the licensee addressed your concems, and we plan no further action in these matters. On August 8.1990, you provided us with concerns asserting that (1) work associated with the overhaul of the containment radiation monitor was improperly controlled and (2) a bypass-jumper tag was improperly controlled during the maintenance of work order M2 90-l 08033. - We provided these concerns in a letter to the licensee dated October 2,1990 and the l licensee responded in a letter dated December 3,1990 (attached). I We note from the licensee response that some problems were identified in the control of work associated with the radiation monitor, but it appears from the licensee's assessment that at least one monitor was operable during the times in question in your assertion. We note that the licensee identified the problems described in your concern and took proper l I corrective actions. Further the licensee is implementing additional controls to establish better hlNSHCQh , l
I 2 coordination of activities between operations and maintenance. Your allegation that there were problems associated with the control of maintenance is substantiated; however, the NRC ' considers the problems minor with respect to nuclear safety and notes that appropriate corrective actions have been taken. With regard to issue (2), the pin connector in question was not connected to anything, therefore a bypass tag or a work order would not be , appropriate. Your allegation may be true, but is of no consequer.ce. In any case, we are satisfied with the licensee's response and plan no further actions in these matters. , On November 9, and November 11,1990, you provided to us (1) a description of events associated with the main turbine stator cooling troubleshooting, (2) a question as to the propriety of completing the thermal margin low pressure surveillance in Modes 3 and 2, and, (3) a question regarding the testing of the control element withdrawal prohibit alarm. We provided these issues to your employer in a letter dated December 6,1990, and received their response in a letter dated January 4,1991 (attached). Your assertions regarding issue (1) were determined to be true, but appear to be of no consequence to nuclear safety. The problems had previously been identified by the licensee and appropriate corrective actions appear to have been taken. With regard to issue (2), it appears no violation of technical specifications occurred, and your allegation appears to be unsubstantiated. With regard to issue (3), the alarm testing was determined to be adequate and your allegation appears to be unsubstantiated. We are satisfied that the licensee answered your concerns and we plan no further action in these matters. On September 28,1990, and in several discussions with NRC personnel during the recent refueling outage, you asserted that there were violations of the overtime policy at Millstone Unit 2. We investigated your assertion and discussed the issue with unit management, but could not substantiate your claim. To enable further evaluation, more specific details are ' needed. We note that you have recently provided ut a similar concern that overtime restrictions may not have been complied with, and we are inspecting this concern. We will inform you of the results of our inspections when complete. On September 28,1990, and in a November 5,1990, memorandum to our resident inspector, you provided the NRC with seven concerns involving: (1) the procedural adequacy and implementation of IC 2419C section 5.5.6 which involve the heated junction thermocouple inspection; (2) the instrument calibration review that you initiated; (3) the instrument calibration review that had been initiated associated with test voltages being out-of-specification; (4) a red tag that was improperly hung on the Weidmuller Block TDD; (5) an assertion that an annunciator window change had been improperly handled by the licensee; (6) an assertion that surveillance procedure SP-240lJ had not been implemented when an instrument and controls technician turned in the paperwork without restoring the bypass key; and, (7) an assertion that you had received harassing mail from a co worker. 1
3 i We referred issues 1 thru 6 to the licensee in a letter dated November 11,1990 and received their response in a letter dated December 21,1990 (attached). We note that the licensee was
\
aware of the discrepancie; that you identified in your assertions and had taken actions to correct the deficient conditions when originally identified. , In regard to issue (1), the procedure was written for a skilled instrument and controls , technician with experience in this type of maintenance; however, you made no assertion that maintenance was improperly or incompletely performed to the extent that operability of essential equipment was affected. In any case, your allegations appear to be substantiated but of minor significance with respect ' nuclear or personnel safety. With regard to issue (2), the licensee provided us a copy of the instrument calibration review (ICR) and we are providing this copy for your review. Please inform us if you have further questions regarding this matter. With regard to issue (3), the licensee had taken action to address the problem that you described and we have not been informed of any inoperabilities that resulted from the corrective actions. With regard to the drawing concern of issue (3), more information is needed to adequately address your concerns. Please inform us of any additional details or further questions that you may have in this matter, With regard to issue (4), a minor tagging discrepancy had been previously icentified and was promptly corrected, Further, guidance on the tagging of Weidmuller blocks was to have been provided to operations personnel. Your allegation in this case appears to be substantiated, but is of little concern with respect to nuclear safety. With regard to issue (5), no problems were identified, and no corrective actions were warranted. With regard to issue (6), we note that operations personnel identified the condition that you asserted and took prompt actions to restore the channel. The technician in this case appears not to have exceeded Unit 2 Technical Specification limitatior.s. However, your allegation regarding implementation of the surveillance procedure was substantiated by the licensee. We are satisfied with the , licensee responses to these six issues and plan no further action in these matters. Finally, with regard to issue (7), the NRC cannot take action based on cosvorker harassment. especially if the alleged harassment is anonymous. If you feel that you au being harassed by your emplcyer, we again remind you to take these issues to the Department of Labor. On September 14,1990, you provided us a three concerns detailing: (1) failure of operators to note that the steam jet air ejector may not be working properly; (2) the failure of your supervisor to respond to one of your questions; and, (3) the failure of another instrr.eut and controls technician to fol!cy an unspecified department instmetion requiring that a trouble report sticker be attached to an instmment after a problem was identified. We provided your concerns to the licensee in a letter dated October 4,190, and they responded in a letter dated December 7,1990 (attached). With regard to issue (1), the licensee identified no inope rability associated with the mdiation monitor, but has identified that upgrade of the system is warranted. The licensee plans to replace the monitor in 1991. Your concerns therdorc have some validity. With regard to
7;. ,. 4 S issue (2), communication between you and your supervisor appears to have been either misunderstood or incomplete. We could not determine the validity of your complaint. With >- regard to iss;:. (3), the licensee determined that the orange sticker was properly placed f _ operator informatwa and that work was properly controlled by the applicable work < documents. Your concerns in this case appear to be unsubstantiated. We are satisfied with the licensee response to the concerns as presented, and the NRC plans no further action wi regard to these matters. We appreciate you informing us of your concerns and feel we have been responsive to th concems. Should you have any additional questions or if 1 can be of further assistance, . please call me collect at (215) 337-3225. Sincerely; n I ' A>W > Edward C. Wenzinger, Chief I rA
' Reactor Projects Branch 4- V Attachments: As stated bec:w/o enclosures e gi., go. A . y f c_(.3 4 M. Moore DRMA (6) g g _ q, .jg, ,g,9 ), gg 4 -
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F01A- kl- d6h The Region I office has completed its followup to the concems that you brought to our attention on the dates described. Relevant documentation such as letters from the licensee a attached to this letter. A synopsis of these concerns and our subsequent actions and regulatory conclusions are detailed below. On October 8, and October 11,1990, you provided to us a number of concems associated with Wide Range Nuclear Instrument operability. You further discussed this issue with me in January,1991. Additionally you provided to us on the same dates, a concern that you were not consulted during a recent procedure change associated with surveillance procedure SP-2417H, and that this omission was contrary to station procedures. We provided these issues to your employer in a letter dated October 26,1990 and they responded in a letter da'ed December 21,1990 (attached). Additionally, we inspected the issue of wide range nuclear instrument operability and provided you the results of our investigation in a letter dated January 14, 1991. Your assertions that spiking had occurred on the channel ".A" of the instrument were true, but - inoperability and violations of technical specifications have not been substantiated. With < regard to SP-2417H, the licensee admitted that you were not consulted for the procedure change, but no violation of procedures occursed and your allegation appears to be unsubstantiated. Furthermore, the procedure change was determined to be adequate. The NRC is satisfied that the licensee addressed your concerns, and we plan no further action in these matters. On August 8,1990, you provided us with concerns asserting that (1) work associated with the evethaul of the containment radiation monitor was improperly controlled and (2) a bypass-jumper tag was improperly controlled during the maintenance of work order M2 08033. We provided these concerns in a letter to the licensee dated October 2,1990 and the licensee responded in a letter dated December 3,1990 (attached). We note from the licensee response that some problems were identified in the control of work' associated with the radiation monitor, but it appears from the licensee's assessment that at least one monitor was operable during the times in question in your assertion. We note that the licensee identified the pob! cms described in year concern and took proper currective actions. Fmther the licensee is implementing additional controls to establish better k%2.M w _ .. s . ..._ . . . , , _ . . , , , , , , , , , , , , , , , , ,, MQD
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( e coordination of activities between operations and maintenance. Your allegation that'there were problems associated with the control of maintenance is substantiated; however, the NRC considers the problems minor with sespect to nuclear safety and notes that appropriate -'. corrective actions have been taken. With regard to issue (2), the pin connector in question was not connected to anything, therefore a bypass tag or a work order would r.ot be appropriate. Your allegation may be true, but is of no consequence. In any case, we are satisfied with the licensee's response and plan no further actions in these matters. On November 9, and November 11, 1990, you provided to us (1) a description of events associated with the main turbine stator cooling troubleshooting, (2) a question as to the propriety of completing the thermal margin low pressure surveillance in Modes 3 and 2, and, (3) a question regarding the testing of the control element withdrawal prohibit alarm. We provided these issues to your employer in a letter dated December 6,1990, and received their response in a letter dated January 4,1991 (attached). Your assertions regarding issue (1) were determined to be true, but appear to be of no consequence to nuclear safety. The problems had previously been identified by the licensec and appropriate corrective actions appear to have been taken. With regard to issue (2), it appears no violation of technical specifications occurred, and your allegation appears to be unsubstantiated. With regard to issue (3), the alarm testing was determined to be adequate and your allegation appears to be unsubstantiated. We are satisfied that the licensee answered i your concerns and we plan r.o further action in these matters. t On September 28,1990, and in sevemi discussions with NRC personnel during the recent refueling outage, you asserted that there were violations of the overtime policy at Millstone Unit 2. We investigated your assertion and discussed the issue with unit management, but l could not substantiate your claim. To enable further evaluation, more specific details are , needed. We note that you have recently provided us a similar concern that overtime restrictions may not have been complied with, and we are inspecting this concern. We wi'll l inform you of the results of our inspections when complete. l On September 28,1990, and in a November 5,1990, memorandum to our resident inspector, l you provided the NRC with seven concerns involving: (1) the procedural adegaacy and implementation of IC 2419C section 5.5.6 which involve the heated junction thermocouple inspection; (2) the instrument calibration review that you initiated; (3) the instrument ' calibration review that had been initiated associated with test voltages being out-of-specification; (4) a red tag that was impropey hung on the Weidmuller Block TDD; (5) an assertion that an annunciator window change had been improperly handled by the licensee; (6) an assertion that surveillance procedure SP-2401J had not been implemented when an instrument and controls technician turned in the paperwork without restoring the bypass key; l and, (7) an assertion that you had received harassing mail from a co-wocer. l I
m., ,
~
G g _3 , We referred issues I thru 6 to the hcensee m a letter dated November Il,1990 and received c their response in a letter dated December 21,1990 (attached). We note that the licensee was i 6 aware of the discrepancies that you identified in your assertions and had taken actions to correct the deficient conditions when originally identified. , In regard to issue (1), the procedure was written for a skilled instruinent and controls . ,. technician with experience in this type of maintenance; however, you made no assertion that maintenance was improperly o_r incompletely performed to the extent that operability of-essential equipment was affected. In any case, your allegations appear to be substantiated but-of minor significance with respect to nucIcar or personnel safety. With regard to issue (2), the licensee provided us a copy of the instrument calibration review (ICR) and we are providing this copy for your review. Please inform us if you have further questions - regarding this matter. With regard to issue (3), the licensee had taken action to address the problem that you described and we have not been informed of any inoperabilities that resulted from the corrective actions. With regard to the drawing concern of issue (3), more . 4 information is needed to adequately address your concerns. Please inform us of any - additional details or furtner questions that you may have in this matter. With regard to issue (4), a minor tagging discrepancy had been previously identified and was promptly corrected.- Further, guidance on the tagging of Weidmuller blocks was to have been provided to operations personnel. Your allegation in this case appears to be substantiated, but is of little - concern with respect to nuclear safety. With regard to issue (5), no problems were identified, and no corrective actions were warranted. With regard to issue (6), we note that - , operations personnel identified the condition that you asserted and took prompt actions to restore the channel. The technician in this case appears not to have exceeded. Unit 2 Technical Specification limitations. ' However, your allegation regarding implementation of the surveillance procedure was substantiated by the licensee. We are satisfied with the - ' licensee responses to these six issues and plan no further action in these matters. ' Finally, with regard to issue (7), the NRC cannot take action based on co-worker harassment,
~
especially;if the alleged harassment is anonymous. If you feel that you are being harassed by ' . your employer, we again remind you to take these issues to the Department of Labor. l On September 14,1990, you_ pavided us a three concerns' detailing: (1) failure of operators to note that the steam jet air ejector may not be working properly; (2) the failure of your supervisor to respond to one of your questions; and,E(3) the failure of another instrument and controls technician to follow an unspecified department instmetion requiring that a trouble report sticker be attached to an instrument after a problem was identified. We provided your; concerns to the licensee in a letter dated October 4,1990, and they responded in a letter dated December 7,1990 (attached). With regard to issue'(1), the licensee identified no inoperability associated with the radiation monitor, but has icentified that upgrade of the system is warranted. The licensee plans to replace the monitor in 1991'. 'Your concerns therefore have some validity. With regard to e g-m' *ur W+_vm.21~*mt-i P- ** re r -- e 4 -- - e--M T Wv- -
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l.'t L 4 4 issue (2), communication between you and your supervisor appears to have been either t misunderstood or incomplete. We could not determine the validity of your complaint. With '" regard to issue (3), the licensee determined that the orange sticker was properly placed for operator information and that work was properly controlled by the applicable work documents. Your concerns in this case appear to be unsubstantiated. We are satisfied with the licensee responw to the concems as presented, and the NRC phuts no further action with regard to these matters. We appreciate you informing us of your concerns and feel we have been responsive to those concerns. Should you have any additional questions or if I can be of further assistance, please call me collect at (215) 337-5225. Sincerely; f - L ) o ieu n ' Edward C. Wenzinger, Chief 1 Reactor Projects Branch 4 Attachments: As stated bec:w/o enclosures ' q g .g .g . y y c{.3..s. M. Moore DRMA (6) . g ), q, _tt ,g,9 ug, gg,,4 . J. Stewart (h) t U~%~k~ '(* \# h = W, Raymond A 1 A - ilt, c fe.,s ced R I A -I&o clov*d p t . 9o -A - n4 elesced 1.
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'Ihe Region I office has completed its followup to the concerns that you brought to our attention on the dates described. Relevant documentation such as letters from the licensee are attached to this letter. A synopsis of these concerns and our subsequent actions and regulatory conclusions are detailed below.
On October 8, and October 11,1990, you provided to us a num'::r of concems associated with Wide Range Nuclear Instrument operability. You further discussed this issue with me in January,1991. Additionally you provided to us on the same dates, a concern that you were l not consulted during a recent procedure change associated with surveillance procedum SP-2417H, and that this omission was contrary to station procedures. We prmided these issuct l to your employer in a letter dated October 26,1990 and they responded in a letter dated l December 21,1990 (attached). Additionally, we inspected the issue of wide range nuclear instrument operability and provided you the results of our investigation in a letter dated January 14, 1991. l Your 2ssertions that spiking had occurred on the channel " A" of the instrument were true, but inopeihbility and violations of technical specitications tave not been substantiated. With - regard o SP-2417H, the licensee admitted that you were not consulted for the procedure but no violation of procedures occurred and your allegation appears to be changh;sntiated. unsubs. Furthermore, the procedure change NRC iQ.atisfied that the licensee addressed your concems, and we plan no further action in these mgers. On Auguk 8,1990, you provided us with concerns asserting that (1) work associated with the overhaul of the containment radiation monitor was improperly controlled and (2) a bypass-jumper tag was improperly controlled during the maintenance of work order M2 08033. We provided these concerns in a letter to the licensee dated October 2,1990 and the licensee responded in a letter dated December 3,1990 (attached). 1 We note from the licensee response that some problems were identified in the control of work l associated with the radiation monitor, but it appears from the licensee's assessment that at least one monitor was operable during the times in question in your assertion. l We note that the licensee identified the problems described in your concern and took proper l corrective actions. Further the licemer. is implementir g additional controls to establish better 1 i
*** 4 j W(D
2 coordination of activities between operations and maintenance. Your allegation that there were problems associated with the control of maintenance is substantiated; however, the NRC c considers the problems minor with respect to nuclear safety and notes that appropriate j corrective actions have been taken. With regard to issue (2), the pin connector in question was not connected to anydting, therefore a bypass tag or a work order would not be , appropriate. Your allegation may be true, but is of no consequence. In any case, we are satisfied with the licensee's response and plan no further actions in these matters. , On November 9, and November 11, 1990, you provided to us (1) a description of events associated with the main turbine stator cooling troubleshooting, (2) a question as to the propriety of completing the thermal margin low pressure surveillance in Modes 3 and 2, and, ~ (3) a question regarding the testing of the control element withdrawal prohibit alarm. We provided these issues to your employer in a letter dated December 6,1990, and received their response in a letter dated January 4,1991 (attached). Your assertions regarding issue (1) were determined to be true, but appear to be of no consequence to nuclear safety. The problems had previously been identified by the licensee and appropriate corrective actions appear to have been taken. With regard to issue (2), it appears no viciation of technical specifications occurred, and your allegation appears to be unsubstantiated. With regard to issue (3), the alarm testing was determined to be adequate and your allegation appears to be unsubstantiated. We are satisfied that the licensee answered your concerns and we plan no further action in these matters. l On September 28,1990, and in several discussions with NRC personnel during the recent refueling outage, you asserted that there were violations of the overtime policy at Millstone Unit 2. We investigated your assenion and discussed the issue with unit management, but l could not substantiate your claim. To enable further evaluation, more spo:ific details are , needed. We note that you have recently provided us a similar concern that overtime restrictions may not have been complied with, and we are inspecting this concem. We will
- inform you of the results of our inspections when complete.
f i On September 28,1990, and in a November 5,1990, memomndum to our resident inspector, you provided the NRC with seven concerns involving. (1) the procedural adequacy and implementation of IC 2419C section 5.5.6 which involve the heated junction thermocouple inspection; (2) the instrument calibration review that you initiated; (3) the instrument calibration review that had been initiated associated with test voltages being cat of-specification; (4) a red tag that was improperly hung on the Weidmuller P, lock TDD; (5) an assertion that an annunciator window change had been improperly handled by the licensee; (6) an assertion that surveillance procedure SP-2401J had not been implemented when an instrument and controls technician turned in the paperwork without restoring the bypass key; and, (7) an assertion that you had received harassing mail from a co-worker. -
B-3 4 We referred issues I thru 6 to the licensee in a letter dated November 11,1990 and received t their response in a letter dated December 21,1990 (attached). We note that the licensee was aware of the discrepancies that you identified in your assertions and had taken actions to correct the deficient conditions when originally identified. , In regard to issue (1), the procedure was written for a skilled instmment and controls technician with experience in this type of maintenance; however, you made no assertion that maintenance was improperly or incompletely performed to the extent that operability of essential equipment was affected. In any case, your allegations appear to be substantiat-d but of minor significance with respect to nuclear or personnel safety. With regard to issue (2), the licensee provided us a copy of the instrument calibration review (ICR) and we are providing this copy for your review. Please inform us if you have further questions regarding this matter. With regard to issue (3), the licensee had taken action to address the problem that you described and we have not been informed of any inoperabilities that resulted from the corrective actions. With regard to the drawing concern of issue (3), more information is needed to adequately address your concerns. Please inform us of any additional details or further questions that you may have in this matter. With regard to issue (4), a minor tagging discrepancy had been previously ideatified and was promptly corrected. Further, guidance on the tagging of Weidmuller blocks was to have been provided to operations personnel. Your allegation in this case appears to be substantiated, but is of little concern with respect to nuclear safety. With regard to issue (5), no problems were identified, and no corrective actions were warranted. With regard to issue (6), we note that operations personnel identified the condition that you asserted and took prompt actions to restore the channel. The technician in this case appears not to have exceeded Unit 2 Technical Specification limitations. However, your allegation regarding implementation of the surveillance procedure was substantiated by the licensee. We are satisfied with the ; licensee responses to these six issues and plan no further action in these matters. Finally, with regard to issue (7), the NRC cannot take action based on co worker harassment, especially if the alleged harassment is anonymous. If you feel that you are being harassed by your employer, we again remind you to take these issues to the Department of Labor. On September 14,1990, you provided us a three concems detailing: (1) failure of operators to note that the steam jet air ejector may not be working properly; (2) the failure of your supervisor to respond to one of your questions; and, (3) the failure of another instrument and controls technician to follow an unspecified department instruction requiring that a trouble report sticker be attached to an instrument after a problem was identified. We provided your concerns to the licensee in a letter dated October 4,1990, and they responded in a letter dated December 7,1990 (attached). With regard to issue (1), the licensee identified no inoperability associated with th,e radiation monitor, but has identified that upgrade of the system is warranted. The licensec plans to replace the monitor in 1991. Your concerns therefore inve some validity. With regard to
i 4 issue (2), communication between you and your supervisor appears to have been either misunderstood or incomplete. We could not determine the validity of your complaint. With ( , regard to issue (3), the licensee determined that the orange sticker was properly pla operator infortnation and that work was properly controlled by the applicable work ' documents. Your concerns in this case appear to be unsubstantiated. We are satisfied with the licensee response to the concerns as presented, and the NRC plans no further action,,w regard to these matters. We appreciate you informing us of your concems and feel we have been responsive to concerns. Should you have any additional questions or if I can be of further assistance, please call me collect at (215) 337-5225. Sincerely;
.h>
Edward C. Wenzinger, Chief Reactor Projects Branch 4 Attachments: As stated bec:w/o enclown ( pi go .4. y y c,(.3.a 4 g g _ q, _,9 7,9 ,g, g,,a M. Moore DRMA (6) J. Stewart % g , , g , 4 , t og c g, ge,a W. Raymond it t A - t % c. f.> s eed ' R I- % ~ A -lho chx=d gg 90 -A - n4 c.lesced
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,e 8**** KING OF PRURSM PENNSYLVANIA 19404 ffR 10 %91 t Information ir, this record was de!eted in accordance with pe reqdyn of Information Act, nem hans W-lL D TOlfe._ dh 7 The Region I office has completed its followup to the concerns that you brought to our attention on the dates described. Relevant documentation such as letters from the licensee are attached to this letter. A synopsis of these concerns and our subsequent actions and regulatory conclusions are detailed below, On October 8, and October 11, 1990, you provided to us a number of concerns associated with Wide Range Nucleu Instrument operability. You further discussed this issue with me in January,1991. Additionally you provided to us on the same dates, a concern that you were not consu!ted during a recent procedure change associated with surveillance procedure SP-2417H, and that this omission was contrary to station procedures. We provided these issues to your employer in a letter dated Ociober 26,1990 and they responded in a letter dated December 21,1990 (attached). Additionally, we inspected the issue of wide range nuclear instrument operability and provided you the results of our investigation in a letter dated January 14, 1991.
Your assenions that spiking had occurred on the channel " A" of the instrument were true, but inoperability and violations of technical specifications have not been substantiated. With regard to SP 2417H, the licensee admitted that you were not consulted for the procedure l change, but no violation of procedures occurred and your allegation appears to be unsubstantiated. Furthermore, the procedure change was determined to De adequate. The ; NRC is satisfied that the licensee addressed your concerns, and we plan no further action in : ( these matters. i On August 3,1990, you provided us with concerns asserting that (1) work associated with the overhaul of the containment radiation monitor was imprcperly controlled and (2) a bypass-jumper tag was improperly controlled during the maintenance of work order M2 90-l 08033. We provided these concerns in a letter to the licensee dated October 2,1990 and the licensee responded in a letter dated December 3,1990 (attached). We note from the licensee response that some problems were identified in the control of work associated with the radiation monitor, but it appears frorn the licensee's assessment that at least one monitor was operable during the times in question in your assertion.
/
We note that the licensee identified the prob! cms described in your concern and took proper corrective actions. Further the licensee is implementing additional controls to establish better
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). 2 coordination of activities between operations and maintenance. Your allegation that there were problems associated with the control of maintenance is substantiated; hnwever, the NRC , considers the problems minor with respect to nuclear safety and notes that appropriate corrective actions have been taken. With regard to issue (2), the pin connector in question was not connected e anythmg, therefore a bypus tag or a work order would not be appropriate. Your allegation may be true, but is of no consequence. In any case, we are , satisfied with the licensee's response and plan no further act ons i in these matters. On November 9, and November 11,1990, you provided to m (1) a description of events associated with the main turbine stator cooling troubleshooting, (2) a question as to the ! propriety of completing the thermal margin low pressure surveillance in Modes 3 and 2, and, (3) a question regarding the testing of the control element withdrawal prohibit alarm. We i provided these issues to your employer in a letter dated December 6,1990, and received their (- response m a letter dated January 4,1991 (attached). L ! Your assertions regarding issue (1) were determined to be true, but appear to be of no - consequer.cc to nuclear safety. The problems had previously been identified by the licensee and appropriate corrective actions appear to have been takers. With regard to issue (2), it appears no violation of technical specifications occurred, and your allegation' appears to be unsubstantiated, With regard to issue (3), the alarm testing was detennined to be adequate and your allegation appears to be unsubstantiated. We are satisfied that the licensee answered , your concerns and we plan no further action in these matters. On September 28,1990, and in several discussions with NRC personnel during the recent refueling outage, you asserted that there were violations of the overtime policy at Millstone Unit 2. We investigated your assertion and discussed the issue with unit management, but could not substantiate your claim. To enable further evaluation, more specific details are needed. We note that you have recently provided us a similar concern that overtime restrictions may not have been complied with, and we are inspecting this concern. We will-inform you of the results of our inspections when complete. On September 28,1990, and in a November 5,1990, memorandum to our resident inspector,
. you provided the NRC with seven concerns involving: (1) the procedural adequacy and j implementation of IC 2419C section 5.5.6 which im olve the heated junction thermocouple
" inspection; (2) the instrument calibration review that you initiated; (3) the instrument calibration review that had been initiated associated with test voltages being out-of-specification; (4) a red tag that was improperly hung on the Weidmuller Block TOD; (5) an
. assertion that an annunciator window change had been improperly handled by the licensee; l
(6) an assertion that surveillance procedure SP-240lJ had not been implemented when an L L instrument and controls technician turned in the paperwork without restoring the bypass key; and, (7) an assertion that you had received harassing mail from a co worker. I
, _. - . , , . _ . _ . - - . _ , _ , _ _ _. . ~ , , , _ _
f 7 3 , We refer x1 issues 1 thru 6 to the licensee in a letter dated November 11,1990 and received t their response in a letter dated December 21,1990 (attached). We note that the licensee was aware of the discrepancies that you identined in your assertions and had taken actions to - correct the deficient conditions when originally identified. In regard to issue (1), the procedure was written for a skilled instrument and controls technician with experience in this type of maintenance; however, you made no assertion that maintenance was improperly or incompletely performed to the extent that operability of - c sential equipment was affected. In any case, your allegations appear to be substantiated but of minor significance with respect to nuclear or personnel safety. With regard to issue (2), ' the licensee provided us a copy of the instrument calibration review (ICR) and we are providing this copy for your review. Please inform us if you have further questions regarding this matter. With regard to issue (3), the licensee had taken action to address the problem that you described and we have not been informed of any inoperabilities that resulted from the corrective actions. With regard to the drawing concern of issue (3), more information is needed to adequately address your concerns. Please inform us of any additional details or further questions that you may have in this matter. With regard to issue (4), a minor tagging discrepancy had been previously identified and was promptly corrected. Further, guidance on the tagging of Weidmuller blocks was to have been provided to operations personnel. Your allegation in this case appears to be substantiated, but is of little ' concern with respect to nuclear safety. With regard to issue (5), no problems were identifled, and no corrective actions were warranted. With regard to issue (6), we note that operations personnel identified the condition that you asserted and took prompt actions to restore the channel. The technician in this case appears not to have exceeded Unit 2 - Technical Specification limitations. However, your allegation regarding implementation of the surveillance procedure was substantiated by the licensee. We are satisded with the licensee responses to these six issues and plan no further action in these matters. Finally, with regard to issue (7), the NRC cannot take action based on co-worker harassmem, especially if the alleged harassment is anonymous. If you feel that you are being harassed by-your employer, we again remind you to take these issues to the Department of Labor. On September 14, 1990, you provided us a three concerns detailing: tl) failure of operators to note that the steam jet air ejector may not be working properly; (2) the failure of your - supervisor to respond to one of your questions; and, (3) the failure of another instrument and- ' controls technician to follow an unspecified department instruction requiring that a trouble report sticker be attached to an instrument after a problem was identified. We provided your-concerns to the licensee in a letter dated October 4,1990, and they responded in a letter dated December 7,1990 (attached). t With regard to issue (1), the licensee identified no inoperability associated with the radiation monitor, but has identified that upgrade of the system is warranted. The licensee plans to replace the monitor in 1991. Your concerns therefore have some validity. With regard to f
, , , - -- ~ . ~ . . n. .,
. 4 issue (2), communication between you and your supervisor appears to have With been either <.
misunderstood or incomplete. We could not determine the validity of your complaint. regard to issue (3), the hcensee determined that the orange sticker was properly plac operator information and that work was properly controlled by the applicable work documents. Your concerns in this case appear to be unsubstantiated. We are satisfied with the licensee response tc the concerns as presented, and the NRC plans no further action wi regard to these matters. We appreciate you informing us of your concerns and feel we have been responsive to concerns. Should you have any additional questions or if I can be of further assistance, picase call me collect at (215) 337-5225. Sincerely;
/f3/7 -
.lt) h Edward C. Wenzinger, Chief 44^ 'h Reactor Projects Branch 4 h Attachments: As stated j- bec:w/o enclosures e q g 9e .4. ,y y c( 3,a M. Moore DRMA (6) g 3_ q,,k ,7,9 cg, g,y -
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Cear l-This letter respds to ycur staterent to the NRC at East Lyme, 0:nnecticut, en June 13, 1989, ard to your June 27, 1989 talepN:na call to the NRC senior resident - r at Ki11 stone. A copy of the transcript of ya.tc statsraant and the 16, 1987 NRC letter to yo2 (signed by n,y sutcrdinata, D:e C. M:Cabe, Jr.) are enclosed for refererce. The enclosed transcript h!Ls been reviewed by no and othex NR' managers an a regietal NRC Allegation Panel. CXAr ruview noted that ycur general ccnoezn abaut. ron-nuclear activities not beirg contro11sd by the rsrlear tacput po:codure remains. He also noted that the transcript of year Jurm 13," 1989 statecent alleges that the Octcber 16, 1987 letter frara the NRC statad: . . .if they (tagging precedures) vare not beirg foliced in a rmscImr capacity, that would carry over into the nuclear capacity." He differ en this point. 02r 0:teber 16, 1987 letter (Drlosure 2) did not rake axh a st.StM. Our regulatory position is that non ruelear activities must rot adversely
- i. ct ruclear activities, but nocier rade precadures need rot be used to a eve that. If r-g, we can n ruelear grade controls cNer Islated rtn-nuclear areas. In the case ycu identified, MFC rollcaip found to unsafe cerriiticn and no carryover into the ruclear area. ,
We do agree with you that orming similar activities under differant precedures can potentia 11 have an adverse isapact en follo<irg procedures, incitriing safety-related procedures. W do rot agree that rwassarily ha; para , unless ruelear grade contmls are generally led to ren-roclear activities. Lesser controls are acceptable in nest rm-r ear cases. If a carrywar h ruclear safety is specifica11 irri1catal for an activity which is rot ccvered by the licensee's ruclear ity assurarce , an evaluaticn is m de, apprcpriata contals are estaolished, and ated precedares are upgraded. As our October 16, 1987 Jett*r to ycu stated, we have continoed,to acnitor activities involving ruclear safety, includirg electrical tag; fin. Overall, we have found the licensee's control of these activities to be acceptable. Our dim _m= lens with the licensee have previded information that, after the ron-nuclear electrical tageut occurrenca you decribed, the licerme decidal to irpose additicani ccritaol over ncn-mclear tagcuts. Aftar considerirg a
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_ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _____.__m_________________um. _ - _ _ - _
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2 JUL 111983 ren-nuclear tagout procature, the licem.ee cptai to use the rarlear tagcut t precadure throughcut the site. This '".'rtheir choice ard not a requirer:ent upcsed by'the HRC. He'acknculedge ycur June 27, 1989 tal request for further follow-up of f W previous allegations. We also a leige receipt of your taleghcre e' state::ent that HRC folicw-up of ycur irtut in 1988 was inaffective in i j ' _preventirn what you fplicW oenerally at statni to be Rillstone the present Staticn, as haqtai prcblem of HMures in allegaticra by not beirg -
; Many different PGc f
~bspectors have fourd ard contirue to fird basically scurd proca3ures ard
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MilYstone ru: lear p:n er plant perfor1:ance have fourd a proper safety ( perspec+.ive ard appnyriate controls over activities affecting safety. le see ro tracable linkage betvoen your allegations ard these of the other allegen you rentioned. Nrther, no significant safety fradequacy has been identified N. m hTc follcu-up of your allegatices. Our eva.1mtion o-f the transcript of your stateent ard of cur telephcne rc:ord of your June 27, 1989 call concluderi that they ccntain insufficient identification of safety specifics to warrant adiiticral NRC follem. ' you to have specific details of any uncozTucted safety fra&quacies er of failm centrol ruelear safety in accordance with the established prgram, we vot apprtciate your sutnittirg the:s to this office in writirg, in detal.1, vi* .nas30 days, for our censideration. Otherv2se, se plan to cicsa ycur allegatici unsutstantiated, with ro further NRC folloN. I bcpa that the above infumstico resolves your corcems to your satisfac- on. Thank ycu for this cpportunity to address your cxrcerra. l t s - 2 -
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W WA Divisicn of r ject ( I DT1cGureS: l (1) Transcript of June 13,19 (2) crpy of october 16, 1987
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- ie 'S NUCLEAR REGULATORY COMMISSION k REGION I k * * * [f #5 ALLENDALE ACAD KING OF PAusslA. PENNSYLVANfA 194o4 b=$ l i 1QCQ Docket No. 50-336 File No. RI-90-0042 Information in this record was deleted
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FOIA- LC
Dear {,
,,, .m. 1 This letter is in response to your letter dated October 27,1989. and in acknowledgement of your telephone conversation with William Raymcad, Senior Resident Inspector, Millstone Nuclear Power Station, on June 5,1990. L in your letter of October 27, 1989, you requested NRC action with respect to a ! proposed settlement agreement and the potential violation of your rights under i 10 CFR 50. We addressed your concerns about the proposed settlement agreement in our letter dated June 14, 1990. Concerning your complaint that your rights I were violated under 10 CFR Part 50 and 29 CFR Part 24 based on findings in ' Inspection Reports 50-336/88-13 and 50-336/89-13, we found your concern un-substantiated. First, in your letter dated October 27, 3989, you did not specify inspection Report 50-336/89-13, but based on our understanding of your ! concerns we surmised that you were referring to that report vice Inspection Report 50-336/88-13 which you reference later. We reviewed the above stated l inspection reports and found no violations of your rights with respect to 10 CFR Part 50 and 29 CFR Part 24. As we stated in our previous correspondence with you in the encloscd letters dated July 12, 1989, and September 26, 1989, we have evtluated your technical allegations as being unsubstantiated and do . not intend any further dedicated inspection followup specific to those con-cerns. But, as we stated before, the NRC is continuing to monitor licensee performance in the areas of work control and tagouts of plant systems important to nuclear safety through routine resident inspector coverage. Concerning your telephone conversation with William Raymond on June 5,1990, you expressed the concern that Connecticut Yankee Atomic Power Company does not operate their reactor properly; specifically you stated "they don't follow pro-l cedures-over there." Your concern does not express sufficient detail to-war-rant additional NRC followup. If you have specific details of procedural non-compilance we would appreciate your submitting them to this of fice in writing, in detail, as soon as possible for our consideration. Thank you for your concern. S erely,A - L
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ward C. n in er, Ch e Projects Branch 4 Division of Reactor Projec h&hYY
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Enclosures:
- 1. Copy of July 12, E.C. Wenzinger to MS1989, letteg f rom
- 2. Copy of Sep 26, 1989 etter from Samuel J. Collins to bcc:
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