A09920, Discusses Review of R1-91-A-0219 Re Various Activities at Millstone Unit 2.W/related Matl
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w (2033665 5000 November 27, 1991 Docket No. 50-336 -
A09920 Ret Employee Concerns Mr. Charles V. Behl, Director Division of Reactor Projects U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Dear Mr. Behl:
1 Millstone Nuclear Power Station, Unit No. 2 R1-91-A-0219 Ve have completed our review of identified issues concerning activities at Millstone Station. As requested in your transmittal letter, our responses do not contain any personal privacy, proprietary, or safeguards information. The material contained in these responses may be released to the public and placed in the NRC Public Document Room at your discretion. The NRC transmittal letter and our responses have received controlled and limited distribution on a "need to know" basis during the preparation of these responses.
ISSUE 01:
i Changes to vendor maintenance and surveillance instructions are not evaluated i and needed procedure changes are not being made in a timely manner. In particular, MP 2720RI - NAMCO Limit Switch Maintenance (E0) references NAMCO EA 4 189-90051 (December 1980) whereas the vendor has superseded EA 189-90051 vith EA 189-90060 (February 4, 1991). This new information varns that removal of the l bottom cover of the limit switch vill negate the qualification (EEO). In addition, NU vas notified in May, 1989 that the vendor would no longer support j re-work and spare parts kits for their solenoid valves and limit switches. The concern is that key craft personnel are not svare of these changes. 1 REQUEST:
Please discuss the validity of the above assertions. If any deficiencies are !
identified, please provide us with the corrective actions you have taken to prevent recurrence. Please provide us with an assessment of the significance l with regard to safety of any identified deficiencies.
I
RESPONSE
In its identification of the reference in procedure MP 2720RI, the issue is
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Mr. Charles V. Behl A09920/Page 2 November 27, 1991 valid as stated; however, the concern about key craft personnel is not warranted. Ve vere aware of the information from NAMCO vhich was,provided to us by the vendor. In the latter part of February of this year, a Millstone Unit No. 2 Maintenance Engineer wrote a memorandum to the Millstone Unit No. 2 Maintenance Supervisor (Electrical) indicating that both NAMCO and ASCO vould no longer support the general tear down and replacement of component parts for their limit switches and solenoid valves, respectively. The memo recommended that the maintenance procedure governing work on NAMCO limit switches be revised to delete information concerning installation of the various spare parts kits and to address limit switch inspection, replacement ano installation of top cover gaskets. A revision to the NAMCO limit switch procedure was drafted and is currently undergoing engineering review.
In the interim, key craft personnel vere notified of the new NAMCO policy by the Maintenance Supervisor (Electrical).
All work, from the beginning of 1991 to present, on components with environmentally qualified limit switches was reviewed and no discrepancies affecting qualification of the limit switches were noted; therefore, there is no safety significance associated with this allegation.
ISSUE 02:
MP 2722B, Annual EOF Diesel Generator Load Run, is deficient in that the division of work responsibilities among electricians, mechanics, and contractors have never been evaluated as appropriate. Further, AVO M2-89-09594, for the annual load test, does not reference MP 2722B or the other controlling procedure, EPIP 4303, and, there appears to be the following discrepancies in the drawings associated with the test:
- 1) Electrical circuit breaker positions on electrical panels KLP1 and KLP2 do not agree with drawing 25205-30007. Circuit breaker No. 26 is in question on ELP1; breakers 10 and 12 are in question on ELP2;
- 2) An electrical remote control panel, PN1 is shovn on drawing 25205-30007; but does not appear to exist;
- 3) The schematic portion of drawing 25205-39002, sheet 3 (or 25205-32008) appears incorrect and is confusing;
- 4) A utility plug is located at the bottom of the electrical power distribution system automatic bus transfer device (ABT). This conflicts with drawing 25205-32008;
- 5) The vendor representative and mechanic involved with the annual load test of the EOF emergency diesel generator are not qualified to perform the electrical portions of the test; and,
- 6) The review of procedures made by PORC is inadequate. i
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v Mr. Charles V. Behl A09920/Page 3 November 27, 1991 REQUESTt "
Please discuss the validity of the above assertions. If any deficiencies are l identified, please provide us with the corrective actions you have taken to prevent recurrence. Please provide us with an assessment of the significance l vith regard to safety of any identified deficiencies.
RESPONSE
In stating that the procedure at issue was not referenced by the work order, this assertion is valid. Instead, the work order referenced a Maintenance Form (270IJ-43) which lists the items of inspection to be. completed during performance of preventive maintenance activities on the EOF diesel.
We were made aware of the above assertions after assigning the procedure at- ~
issue to an individual for review and revision. The procedure at issue was ,
written as revision 0 in 1985 for an annual load test of the diesel in the .
Emergency Operations Facility (EOF). The procedure was reviewed in 1987 and minor typographical and formatting changes were identified for future revision.
These minor changes vote not considered substantial enough to warrant a rewrite of the procedure at the time. In the six times that the procedure has been implemented since it was first issued, no concerns vare raised regarding its ;
adequacy.
As the procedure is currently written, the only electrical requirement is the verification that leads are de-energized by use of a voltmeter. The diesel manufacturer representative provides equipment which is attached to the diesel i to conduct the test. Contrary to the assertion, we have evaluated the division of work responsibilities, and since the contractor mentioned in the assertion is trained by the manufacturer of the equipment being tested, we consider this manufacturer's representative qualified to participate in the conduct of the annual test.
I
.The assertion of inadequate Plant Operation Review Committee (PORC) review is-l not supported by any facts describing the nature of the inadequacy; therefore, j ve are addressing PORC reviews in general. The membership of the PORC, consisting of the department heads for the four functional groups of the plant Operations, Engineering, Maintenance and Instument & Control (IEC), helps to ensure that these areas of expertise are brought to bear on the procedures and changes brought before the committee. Thus, a procedure or change originating in one functional area of the plant is exposed to review by the other functional areas in a manner intended to provide a broad review of the possible impact of any procedure or change.
I i
The PORC reviews procedures from a global perspective and relies on subcommittee reviews, by persons more familiar with the tasks, to incorporate information concerning division of work responsibilities and appropriate references.
Procedure reviews by PORC are from the proper perspective and are adequate.
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.4 Mr.-Charles V. Behl A09920/Page 4 November 27, 1991 The assertions of drawing discrepancies are being evaluated by the Millstone Unit No. 2 Engineering department. Evaluation and submission of any necessary drawing changes should be. completed by late December, 1991.
Af ter our preliminary review and evaluation of these issues, we find that these issues did not present any indication of a compromise of nuclear safety. We vill provide an immediate update should our ongoing drawing evaluation indicate otherwise. We appreciate the opportunity to respond and explain the basis of our actions. Please contact my staff if there are further questions on any of these matters.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY 4
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J . F . Opek's U ,
Executive Vice President cc: V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 E. C. Venzinger, Chief Projects Branch No. 4, Division of Reactor Projects E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, U.S. Nuclear Regulatory Commission, Millstone I
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January 9, 1992 EN2-92-009 To: Ralph Bates MP2 Engineering Supervisor Millstone From: Dick Perry M =-- '
MP2 Engineering (extT 6067)
Millstone Subj: EOF Diesel Generator
Reference:
1/6/92 meeting Pete Regan, USNRC
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Item #1: TS1 control power connection designations !
(EA,EB,EC,NA,NB,NC) appear on 25205-39002 sh.3 i but do not appear on 25205-32008. Drawing 25205-32008 is a later version of the schamatic diagram i
shown on 25205-39002 sh.3 .
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Response
TS1isthelabelfortheencluaurhathouses connections made inside and as such is not an actual label for the above connections. These connections are attached to the same lugs used for power connections. In the absence of a physical label for the control power attachment, drawing 25205-32008 correctly represents the as-built condition.
Item #2: The following jumpers appear on the old r! ras ;.: i (25205-39002 sh.3)- but do not appear on ' h ..
2w i drawing (25205-32008): 1
- 1) TBG-9 to TB7-13
- 2) TB6-1 to TB7-1
- 3) TB6-2 to TB7-2 Response: The plant equipment (transfer switch panel) was walked down to resolve differences between the old drawing and the new drawing. It was determined that the above three jumpers do appear on the new drawing 10 and the i
jumpers do appear as(areas 7C throug.1 depicted in drawing),39002 l sh.3 .
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JAN 9 '92 16:57 PAGE.003
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Item #32 Notes 2;3,4 & 5 are listed on the old drawing ~-
but not on the new drawing.
responses The above notes are operating instructions and are incorporated into the applicable operating procedures. Since the wiring diagram section of We drawing still appears on the old drawing the entire drawing should be cleaned up (remove ,
hatched out schematic, remove border between the wiring notes nextand the schematic to the and put theDM2-P-00 wiring diagram). existin$-92 (DCN) will accomplish this.
Iten #4: The load bank mentioned in the operating procedure does not appear on any of the drawings.
response: The load bank is portable. The plant equipment does have a connect / disconnect arrangement to '
accommodate the load bank. The load bank is connected and disconnected directly to the' output breaker in accordance with MP 2722B. Therefore, the load bank is not included or mentioned on the drawings.
The cc: JW Riley file
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January 14, 1992
'IO: Mark Heinonen, l
FROM: Bruce Danielson (Ext. 2830)
SUBJECT:
MP2 Mechanical Training Program f'-Ittee (TPCC) Meeting -
January 13, 1992.
Mark Heinonen training on theand ForBruce diesel.Danielson met to discuss the need for After a review of task / training requirements, it was agreed that training would be required. he scope of training will be determined by a Task Analysis. We task analysis will be assigned to the MP2 Mechanical unit instructor (Dan Spencer).
Bruce Danielson agreed to involve the Electrical Training group to determine if electrical training would also be required. ,
i BRD:ca cc: Brendan Duffy - MP2 Maintenance Manager Bob Rowe - MP2 Maintenance Supervisor ,
i Dan Spencer - MP2 Unit Instructor John Kiss - Supervisor, Technical Training - Electrical Pete Regan - Millstone NRC Office I
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FROM: tephe Scace %
Millstone Station Director Millstone Ext. 4300
SUBJECT:
Cancellation of " Emergency Operations Facility Emergency Diesel Generator", EPIP 4303.
" Emergency Operations Facility Emergency Diesel Generator", EPIP 4303, has been cancelled and replaced with Millstone Unit 2 procedure " Emergency Operations Facility Emergency Diesel Generator". OP 2399A.
SORC Mtg. No. D~) Effective: 1 ll 9}
SES/MJC I
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g MP 92-155 Febmary 10, IM2 TO: 4000 Series Copyholders cca.c -
FROM: tephe E. Scace Millstone Station Director Millstone Ext. 4300
SUBJECT:
Cancellation of " EOF Fire Detection System", EPIP 4306.
" EOF Fire Detection System" EPIP 4306, has been cancelled and replaced with .
Millstone Unit 2 procedure " Emergency Operations Facility Fire Detection System", OP 2399B.
SORC Mtg. No. 4M Effective: 1 il hJ l SES/MJC e
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, MP 92-156 February 10, 1992 TO: 4000 Sen.es Copyholders FROM: tephe E. cace Millstone Station Director Millstone Ext. 4300
SUBJECT:
Cancellation of " Emergency Operations Facility Ventilation System Filter Testing-Annual", EPIP 4605, and associated surveillance form
" Emergency Operations Facility Ventilation System Filter Test", EPIP
. Form 4605-1.
" Emergency Operations Facility Ventilation System Filter Testing-Annual".
EPIP 4605, and its associated surveillance form " Emergency Operations Facility Ventilation System Filter Test", EPIP Form 4605-1, have been cancelled and replaced with Millstone Unit 2 procedure " Emergency Operations Facility Ventilation System Filter Test", SP 2678A, and " Emergency Operations Facility Ventilation System Filter Test", OPS Form 2678A-1. .
SORC Mtg. No. 4M. Effective: 2 l
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- FROM:
ew tephen E. Scace Millstone Station Director Millstone Ext. 4300
SUBJECT:
Cancellation of " Emergency Operations Facility Emergency Diesel Generator Operability Test", EPIP 4606, and its associated surveillance form " Emergency Operations Facility Emergency Diesel Generator Operability Test", EPIP 4606-1. ,
" Emergency Operations Facility Emergency Diesel Generator Operability Test", EPIP 4606, and its associated surveillance form " Emergency Operations Facility Emergency Diesel Generator Operability Test", EPIP 4606-1, have been cancelled and replaced with Millstone Unit 2 procedure " Emergency Operations Facility Emergency Diesel Generator Operability Test", SP 26788, and
" Emergency Operations Facility Emergency Diesel Generator Operability Test",
OPS Form 2678B-1 ,
SORC Mtg. No. 4A ' Effective: -N \t %
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MP_92-158 TO: 4000 Series Copyholders February 10, 1992 C
FROM: Stephen E.
em cace Millstone Station Director Millstone Ext. 4300
SUBJECT:
Cancellation of " EOF Air Lock Operability Test", EPIP 4608, and its associated surveillance form " EOF Air Lock Operability Test", EPIP Form 4608-1. ',
" EOF Air Lock Operability Test", EPIP 4608, and its associated surveillance form " EOF Air Lock Operability Test", EPIP Form 4608-1, have been cancelled and replaced with Millstone Unit 2 -
Lock Operability Test". SP 2678C, procedure " Emergency Operations Facility Air and " Emergency Operations Facility Air Lock Operability Test", OPS Form 2678C-1.
, SORC.Mtg. No. 41 Effective: tl tt Ih :
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yp_92-159 TO: 4000 Series Copyholders February 10, 1992 FROM: Steph n
E. Scace Millstone Station Director Millstone Ext. 4300
SUBJECT:
Cancellation of " EOF Fire Detection System Test", EPIP 4609, and associated surveillance form " EOF Fire Detection Test", EPIP Form 4609-1.
" EOF Fire Detection System Test". EP!P 4609, and its associated surveillance form " EOF Fire Detection Test", EPIP Form 4609-1, have been cancelled and replaced with Millstone Unit 2 procedure " Emergency Operations Facilit Fire Detection System Test", SP 2678D, and " Emergency Operations Facilit Fire Detection System Test", OPS Form 2678D-1.
SORC Mtg. No. D -7 Effective: lbt l91 i SES/MJC
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9
atJar% TION RECEIPT RDORT h @I-91 M "0D Date/ Time Allegation No.
Received: August 9, 1991 0900 L
Name:
Address:
City /St./ Zip:
Phone: . .
7 Confidentiality:
Was it requested? No Alleger's Employer: NNE00 Position /
Title:
i Facility: Millstone Unit 2 Docket No.: 50-336 Allegation Summary:
Licensee management is not responsive to the concerns of employees because of a disagreement over the use of " Job leaders" who are untrained and not paid for the supervisory responsibilities when so assigned.
(This has been a recurring issue and is documented for the record).
1 Number of Concerns:
Employee receiving allegation: J. T. Shediosky Type of regulated activity: Reactor Functional Area (s): Operations The alleger believes that when a technician Detailed Description of Allegation:is assigned as the " Job Leader" for a sp supervisory position for which the person has not been trained.
The alleger has discussed the assignment of supervisory duties in the planning and conduct of work activities with hie department manager and with the Millsto Station Director. The alleger was provided with a deta However, the alleger those for individuals in manual (or technician) positions.
disagrees with these duties at his level.
During discussions with the alleger, he (This issue has recurred frequently.
understood that the " Job Leader" had no authority concerning wages or disc ary actions for those assigned to the job task.
alleger stated his intention to bring this complaint to the State of The Connecticut Department of Labor because the issue dealt primarily with wages.
This allegation is submitted for the record with no action proposed).
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h AUG 14 '91 18:13 HRC MILLSTONE OFFICE P02 ALLEGATION RECEIPT REPORT Date/ Time Receivec: , MT .l4$l Allegatton No. 2 I-9 f-4-0 22 0 (leave blank)
Name: _ _ , _ _ _
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 No Does a confidentiality agreement need to be sent to alleger? Yes No Has a confidentiality agreement been signed? Yes No -
Memo cocumenting why it was granted is attached? Yes No -
p NM ii Position /
Title:
$C Facility: / 8b Docket No.: h *M (Allegation Summary (brief description of concern (s): FW nounders oncewn imdante owdm.
totosmkrentpro6tes (infi.) , and ne min Iorep idgs ?J nstv.mratery atwKu lmMkJ Aovi.
Number of Concerns: i1 Employee Receiving Allegation: EIkd (first two initials and last name)
Reactor Type of Regulated Activity (a) _ Vendor (d) _ Other: Safeguards (e) -
(b)
(c1 ~~ _ Materials (5pecify)
Materials License No. (if applicable):
Functional Area (s): ) Operations e) Emergency Preparedness
__(b) Construction _((f) Onsite Health and Safety
[(c) Safeguards [ Offsite Health and Safety
___(d) Transportation _
other:
. I ( f) /
(NRC Recion I Form 207 g/V/
@ f SAMPLE RECORD OF ALLEGATION PANEL DECISIONS SITE: bi47oNE PANEL ATTENDEES:
ALLEGATION NO.: h Al-A o22o Chairman - Nk I l DATE: UMO (Panel No.62! 3 4 5) Branch Chief - ,
PRIORITY: High @ IAw Section Chief (AOC) - Mlv i SAFETY SIGNIFICANCE: Yes l CONCURRENCE .
No h Sr. Alleration Coord (SAC) E kreegdeg-OI Reoresentative - l l TO CLOSEOUT: DD BC (Other) Awde !
CONFIDENTIALITY GRANTE@D: Yes @
(See Allegation Receipt Report) i IS THERE A HARASSMENT / DISCRIMINATION
- ISSUE:
! IF YES, Yes Q i 1) has the individual been informed of the DOL process and the need to file a complaint within 30 days Yes No
- 2) has the individual filed a complaint I with DOL Yes No l 3) has a letter been sent to the complainant seeking Yes No l any safety concerns l IS A CHILLING EFFECT LETTER WARRANTED: Yes h j IF YES. HAS IT BEEN SENT Yes No I HAS THE LICENSEE RESPONDED TO THE CHILLING '
EFFECT LETTER: Yes No ACTION:
- 1) fe er ko iC fA$l 4.
2) 3)
4) 5)
NOTES:
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b ALLEGATION MANAGEMENT SYSTEM ALLEGATION NUMBER - RI-91-A-0220 RUN DATE: 09/11/91 DOCKET / FACILITY / UNIT: 05000336 / MILLSTONE 2 /2 DOCKET / FACILITY / UNIT: / /
DOCKET / FACILITY / UNIT: / /
DOCKET / FACILITY / UNIT: / /
ACTIVITY TYPES - REACTOR MATERIAL LICENSES -
FUNCTIONAL AREAS -
OPERATIONS DESCRIPTION - 1) PROCEDURE REVISED JUNE 91 WAS INADEQUATE, HAD WRONG UNITS ON DATA SHEETS, CHECKED WRONG PANELS
- 2) PROBLEMS CONTINUE WITH WORK ASSIGNMENTS AND " JOB SUPER-CONCERNS - VISOR" DESIGNATIONS.
3 3) MANY AIR SYSTEM VALVE DO NOT HAVE IDENTIFICATION TAGS, INCLUDING INSTRUMENT ROOT VALVE. SOME LABELED VALVE ARE MARKED WITH WRONG IDENTITY.
SOURCE - LICENSEE EMPLOYEE CONFIDENT - NO RECEIVED - 910814 BY - PJ HABIGHORST / RI ACTION OFFICE CONTACT - EM KELLY -
(FTS)346-5183 SAFETY SIGNIFICANCE - UNKNOWN BOARD NOTIFICATION - NO STATUS - OPEN SCHED COMPLETION - 911231 DATE CLOSED -
ALLEGATION SUBSTANTIATED - ALLEGER NOTIFIED -
OI ACTION - OI REPORT NUMBER -
REMARKS - ALLEGER PROVIDED COPIES OF INTERNAL MEMORANDA AND PROCEDURES TO SUPPORT HIS CLAIM.__ PANELED 23AUG91.
SUPPORT OFFICE: RPS-4A ACTION PENDING: REFER TO LICENSEE DOCUMENTATION:
ALLEGER LAST CONTACTED: 14AUG91
REFERENCE:
KEYWORD: PROCEDURES, WORK CONTROL ENTERED SYSTEM - 910822 CLOSED SYSTEM - RECORD CHANGED - 910823 i
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. g I 6/-A / -0?7o-oi SAFETY RELATED INSTRUMENTATION VALUE LINE-UP VERIFICATIOW ' '
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ACCEPTANCE CRITERIA MET:
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.s ACCEPTANCE CRITERIA i
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IN ACCORDANCE WITN REFERENCE PROCEDURE ,, // INITIALS
- 1. PREREQUISITES / INITIAL CONDITIONS COMPLETED [M "d) - @ ~~ 6[
- 2. PRECAUTIONS NOTED (( '""'3f)-/iO S - Y,
- 3. COMMENTS: ilF MAINTENANCE RESTORATION. INDICATE BELOW WORK'dR' DER #. ETC.)
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1&C Form 2436A-1 h Rev.2 b
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!I FT 5277B Aux 3:P6 West Penetr;; tion 5277A Aux 38'6 West Penetration I
Y T Color Desired Check By *-
Color Desired Check By T N r T N r Isolation 4 Red Open Isolation 4 Red Open
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Equalizer / Equalizer / 5 N/A Close N/A Close Vents 5
[ if,$[qt Vents /
FT 5278A Aux 38'6 West Penetration FT 5278B Aux 38'6 West Penetration Y ta Check By Color Desired Check By Color Desired T N r T 7 Open isolation 4 Red Open Isolation 4 Red ,
Equalizer / 5 N/A Close 8 Equalizer /
Vents 5 N/A Close g Visht, Vents g / , ili'sht i
Aux -5 West Penetration FT 306 Aux -25'6 Near Charging Pump Room PT 8117 Check By Color Desired Check By Color Desired T e Nu r T Nu bu
[ Isolation 4 Red Open b isolation 3 ' Red Open
- Mk Equalizer / N/A Close g/ Equalizer / 3 N/A Close DN Vents 2
- d Vents . 61 I 1&C Form 2436A-1 Rev. 2 Pace 2 of 5 SEP 17 Y131
7 Aux -25'6 Wast Penetration-' I 311 Aux -25'6 West Penetration FT 312 Va Tma "*
V* Tm Desired Check By . Color Desired Check By g9n r Color ,
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FT 321 Aux -25'6 West Penetration FT 322 Aux -25'6 West Penetration Check By Color' Desired . Check By I' Color Desired T Num
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isolation 4 Red Equalizer / 3 N/A Close # Equalizer /
Vents 3 N/A Close
), g Vents Cg ---
i FT 331 Aux -25'6 West Penetration FT 332 Aux -25'6 West Penetration To a Check By Color: Desired Check By Color Desired T N r T N 7 Isolation 4 Ired Open Isolation 4 Red Open p fu00 l: b Equalizer / 3 N/A Close M Equalizer /
Vents 3' N/A Close Vents Qg I s l & C F cr m ." '6A-1 '
Rev. 2 Pace 3 of 5 SEP 1 7 1991
. _-_ _ _ _ _ _ . _ _ _ . - - - . - ~ _ . - _ _ . - . . _ . _ . - _ _ . . . . . _- _ _ . _ _ _ _ . _ . . _ - . . _ _ _ _ . _ _ __._.__..__J._
' I 341 Aux -25'6 West Penetration .II - FT 342 Aux -:'5'6 West Penetration- '(- ,
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'k T Color Desired ' Check By.
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T Nu 4 Red Open s isolation 4 Red Open ,
n isolation Ih lAAN Equalizer / 3 N/A Close Equalizer /
vents 3 N/A Close Q vents @
FT 6094 Aux -25'6 West Penetration
- FT 6085 Aux -25'6 West Penetration V" T Color '. Desired Check By Y T Color Desired Check By .g. Nu r ,
T Num YY 4 Red , Open YY l Isolation 4 Red Open UN Isolation Y
Equalizer / N/A' Close Equalizer / 3 N/A Close Vents 3
yp Vents @
i FT 6043 Aux -45'6 "A" ESR FT 3023 Aux -45'6 "A" ESR
- Desired Check By Y T Check By r Color .
Color Desired .g.
i T N
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Isolation 4 Red Isolation 4 ' Red Open '
15 Equalizer /
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Vents -
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I&C Form 2436A-1
- Rev. 2 Pace 4 of 5 SEP 17 1991 . .-
F 12 4 Aux -45'6 "B" ESR ; FT 6042 Aux -45'6 "B" ESR *
- T V T Color Desired Check By Color Desired Check By T N r T Nu r Isolation 4 Red Open xs isolation 4 Red Open e~ro ne m -
Equalizer / 3 N/A Close Equalizer / 3 N/A Close h Vents Vents j LT 3001 RWST (OUTSIDE) LT 3002 RWST (OUTSIDE)
' T Color Desired . Check By 8 Color Desired Check By T Nu r T N bu isolation 1 Red Open Isolation 1 Red Open ,
Equalizer / 1 N/A Close Equalizer / N/A Close Vents 1
p / p $$p Vents J / p $5ht ,
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i RWST (OUTSIDE) LT 3004 RWST (OUTSIDE)
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p/94$1 Equalizer / N/A Close
- Equalizer /
Vents 1 N/A Close g}g'b3m Vents 1
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1&C Form 2436A-1 '
Rev. -?
l' ace 5 of 5 SEP 1 7 1991 .
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i . ~ , i CONTROLLED ROUTING ACTIVITY :
1 unrr i
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- To unrr a JSK oa.e<a asv f.ei 2-91-0095 I l L can arv ouacara 8139 Sylvia i vurrevs 1/1/92 D'; errasvs acan aav out oara Brendan Duffy jg. .T,,s ,
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@ A CH date en.nge to 3/3//42 is reauesteo vor ine voiiowing re. son:
amt / . m 2' ) n A _ 1 , L A l m i/,t-J o r v i i O n. ss sn enio: Aciion coce: ove o.ie:
O s.no copies ei tnis vorm .no .it.cnments to:
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. 6 * '
{STATIO.>N
- ~+ r. s a:. s. DIRE,CTOR/
- . . e m. SECRETARY ACTION
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NUCLEAR OPERATIONS ASSIGNMENT (NOA): Q wu.srows uurr s
=ttsto~e unita 20'-
The response to tne subsect NOA ts attacned/=bove. Please close tnis NOA. ;.O .-e vait a This is a partial response to tne subject NOA. Phase close tnis item. 'De -costo~e v~rr seaves.
O A ove o.t. extentai to ior tne sue,eci NO4 is requesieo. U at's'oas =>t< =<av'crs Di WelsToNE sTAttoN O see above en secoon vor reason.
O rn.s is beino torwaro.o ior iniormation any.
Thss converms . tencon between and on The followmg occurr.d:
O The NOA was cancelled O An == tension to -.scet.ineo. O otner:
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. NORTHEAST UTILITIES -
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(0) - 1 January 6, 1992 l HP-2-92-8 j l
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l TO: Stephen E. Scace Director, Millstone Station FROM:
Jt W n S. Keenan irector, Millstone Unit 2 (Ext. 4302) )
SUBJECT:
CR 8139 Plant Labeling Program The labeling of the remaining portions of systems and components in conformance with I ACP 6.22 will be completed by a revitalized labeling program dove-tailed with the PMMS system. A database of component ids and descriptions will be maintained with a systematic process to ensure all major equipment labeling is verified or identified for labeling.
The tracking mechanism for this program will be via quarterly updates to CR 8139 until closure.
BJD/st cc: B. Duffy J. Diamond J. Criscione CR 8139 File l
l 0570 REV,7-46 BP91
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CONTROLLED ROUTING ACTIVITY OCT 3I 199l Opecie mEv.3-es 1
- FOMwAAD TO: INITIAL, RETURN TO:
FRM STATION SECRETARY / g , gg pf f
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N J // STATION SUPERINTENDENT , g[,
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_ Samvices OTNEa //
STATION SECRETARY NQA a
[Q NOA QUE DATE I
cma CR REY. NOA AEV/
SveJECT
$/3 $ CM// DUE?D
/ TEl9/ CATa j
- 9 C n dy h6 9 het l J N < ;
Work hours expended on this CR: As of. / / N Work Hours: O The following copyrtent or attachtrent(s) should close the CR. Please take appropriate action (sign memo, etc.) if any. f r
is requested for the following reason.
,y r A CR date change to " < - -
fn u o Y&r fe S*iYk b /N ./ /~A #1 V
l BELOW THIS LINE FOR STATION SUPERINTENDENT / SECRETARY USE ONLY l STATION SUPERINTENDENT / SECRETARY ACTION:
O Ciose in. ree.rence CR.
O Change CR Action Code to: l Send copies to:
hange date of referenced CR as requested /// %
Change date of referenced CR to:
cv O Rosassign CR io:
OOin.r:
Action Code: Ove Oste:
NUCLEAR OPERATIONS ASSIGNMENT (NOA):
~~
Forward a copy of this sheet (and attachments) to: VICE PRESIDENT. NUCLEAR OPERATIONS _. Millstone Unit 2
'~~ '#
The response to the subject NOA is attached /above. Please close this NO A. _ Millstone Services
~
0 Tnis is a periisi ,espon io ine subject NOA. Please close this item.
,for the subject NOA is requested. See CR section for reason.
A due date extension to This is being forwarded for information only.
and on O Tnis confirm. s toicon bet een The following occurred:
..s obtained. Otner:
The NOA was cancelled. O An e ionsionio Comments:
Ext.
For further information concerning this NO A, presse catt PEV. 3 DATE: 2-1-89 >
.+ .
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Novenber 5,1991 MP-2-91-157
~
TO: S. E. Scace Station Director Mi Istone 3
FROM: . S. eenan Unit 2 Director Millstone (ext. 4302)
SUBJECT:
CR 8139 - Plant Labeling Program The following provides an update for system and component .
labeling CR 8139 as it pertains to Unit 2 Maintenance.
B. MAJOR COMPONENT LABELING A walkdown of major components was performed. During the walkdown 63% of all components identified were marked during previous efforts to label components. Nearly all of the remaining components while not marked proper are easily identifiable from local valve tags.
D. PIPE E VENTILATION DUCT LABELING:
There are no plans for labeling of pipe and vent duct at this time.
E. BUS, MOTOR CONTROL CENTER (MCC), SWITCHYARD COMPONENT, BREAKER E SWITCH LABELING:
Both 4160 volt and 480 volt breaker cubicles have been completed. These labels were installed by Unit 2 Maintenance and tracked by Engineering. No further i labeling is planned.
l l
l 0570 REv.7-45 c ce -
r:
.o.
F, ROOM LABELS. AREA LABELS AND EMERGENCY EOUTPMENT LABELING:
During the plant walkdown, it was noted nearly all of the doors to high-radiation areas and all emergency equipment have been properly identified. No further action is anticipated.
G. LIGHTING. 125VDC, DISTRIBUTION PANELS?
There are no plans to pursue labeling of lighting and distribution panels beyond the existing identification plates.
The labeling program at Unit Two is very well along. However, further improvements can only be accomplished by a time-intensive review of all the above items to fully establish their conformity with ACP 6.22. With the impending Steam Generator Replacement outage, the time needed to do a good job on this project is unavailable.
Based on the recent inputs in this area (i.e., Chairman Selin), a Station philosophy and action plan needs to be developed. Please extend this controlled routing to January 1. 1992, cc: B. Duffy J. Diamond J. Criscione
- CR 8139 File O
p ,
CONTROLLED ROUTING ACTIVITY - SEP l 71991 8
- . v.
'0 - Fomw Amo to- INiflAL RETVRN TO-rag.
wiy ,
STATION SECRETARY g, p / . d d 9 f*
2 UNIT t t c WW1 AAl
- UNIT 3 STATION SUPERINTENDENT [ [/ ,b
- SERVICES STATION SECRETARY d g M/ [gh . p ,
_ OTHEA CR REV. NOA s NOA REV. / NDA QUE oATE /
CM Pnf SUSJEC
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Work Hour.;
O wo,= nour. .xp.nd.d on ini. CR: A. of: / /
O Tn. foiiowin, comm.ni or .ts.cnm. nit.> .novid cio in. CR. ei.. ... . propri.i. .ciion c.ign ,n.mo. .ic.> ,f .ny
! [ 1. r.que.t.d for tn. following re..on.
A&d.t. cn.ng. tovmm r pusw.
i BELOW THIS LINE FOR STATION SUPERINTENDENT / SECRETARY USE ONLY STATION SUPERINTENDENT / SECRETARY ACTION:
Clo.. . r.f.r.nc. CR. Cn.ng. CR Action Code to-Cn.ng. d.t. of r.fer.nc.d CR .. recue.t.d Cn.ng. d.t. of ref er.nced CR to:
g& D S.nd copie. io:
l' O R.....ign CR to:
DOin.r:
Action Cod.: Du. D.te:
l NUCLEAR OPERATIONS AS$lGNMENT (NOA):
- " " " ' )
O Forw.rd . copy of ini. .n i i.nd .it.cnm.ni.i io: viCE eRESioENT. NuCteAR OPERATIONS ._ Milt. tone Unit 2 O Tn. r..pon.. . in. . uni.ci NOA i. .ii.cn.de... . ei.. cio.. ini. NOA. ; [lga;[al, 1
- ""* S'
- 0 Tni. i. . p.rii.i ,... n.. io in. .ue3.ci NO A. ei.... cio ini. iiem.
l for tn. .ubjeci NOA i. requ..t.d. Se. CR ..ction for re..on A du. d.t. .xt.n. ion to l Tni i. bemg forw.rd.d for inform. tion enly.
O Tni, conti,m. . . icon iw n .nd on Tn. following occurr.d:
OTn.NOAw.cnc.ii.d. O An. sten. ion to w.. obt.ined. O Otn.r:
Comm.nt.:
Ext. ,
For furtn.r inform. tion conc.rning tni. NO A, pie... c.ll AEV.2
, DATE: 2-1-8.
J
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l b I NORTMEAST UDLrt1ES 5 l
g l ' :;:~: ='.r.m1Ta;:'?
r.:.*ta::L'O'A., E w me.es weer ewn car o n _
September 30, 1991 MP-2-91-139 TO: Stephen E. Scace Director, Millstone Station
. l J
FROM: ohn S. Keen sin /
listone Unit 2 Director (Millstone Ext. 4302)
SUBJECT:
CR 8139, Plant Labeling Program The following provides the plans for a system and component labeling schemes in accordance with ACP 6.22 and the guidance in INPO 88-009. It also provides a status report and projected schedule of completion for the labeling.
A. VALVE LABELING:
This program has been underway for approximately 3 years. To date, aoproximately 89Fo of the valves have been relabeled.
vnit 2 Operations continues to pursue labeling of the remaining valves with a projected completion date of December 1992.
B. MAJOR COMPONENT LABELING:
Stenciling or stick-on labels is the method that had and will continue to be used.
Approximately :).QS of the major equipment is identified to date. Formal plans for completion of this segment have not been established to date. Unit 2 Maintenance has the lead on completion of this segment of plant labeling.
C. INSTRUMENT AND GAGE LABELING:
This taskwas initiated in 1990 by Unit 2 I&C. To date,17 of 128 systems have been completed. Limited personnel resources have not allowed a consistent effort to complete the task. In addition, for each instrument, the PMMS data base, loop folder and P&ID must be verified and updated as necessary.
Unit 2 I&C will continue to complete this task on a level of effort basis. Expected completion date is December 1995.
D. PIPE AND VENTILATION DUCT LABELING:
This segment of plant labeling has not been formally initiated to date.
Approximately 159o of pipe and ventilation duct work had been previously labeled before issuance of ACP 6.22 and INPO 88-009. Maintenance has the lead on this segment to provide a long term plan of action.
4 E. BUS. MOTOR CONTROL CENTER (MCCL SWITCHYARD COMPONENT.
BREAKER AND SWITCH LABELING:
Limited retabeline of bus, MCC breaker and switches has been initiated but a formal program rieeds to be established. Maintenance has the lead to determine the extent of labeling to date and evaluate the level of completeness.
.c
m
'~
'I- .
. F. ROOM LABEIX AREA LABEIS AND EMERGENCY EOUIPMENT LABELING:
L Room and area labels need to be reviewed for completion as part of the overall alant labeling effort. Most, if not all, are labeled per the criteria in ACP 6.22 and NPO 88-009. Unit 2 Maintenance has the lead to complete this segment. A formal door labeling effort is near completion. This effort displays door attributes and also some room and area information.
G. LIGHTING.125VDC. DISTRIBUTION PANELS:
An interim number of steps were taken to identify loads for low voltage panels this past year. Long term final plans need to be established to fully complete this segment. ,
The update to this CR will provide a definitive plan for completion of Items B, D, E, F and G. Please extend CR 8139 to November 1,1991.
JSK/JLC:cle cc: List 2DH J. E. Diamond CR 8139 File O
e 9
s I
I IO N PUBLIC DISCLOSURE ENCLOSUP.E Issue 91-220-01 It has been noted that many Air System Valves do not have system identification tags, causing identification confusion, and multiple active Red Tag for the same problem. In particular, during performance .! SP-2404AP, Waste. deutralization Tank P.adiation Monitor Functional, it was noted that the Instrument Root Valves to valves 2-CND-34 and 2-CND-37 did not have system identification tags, only duct tape with hand-written identification marked AOV244A and AOV244B.
Issue 91-220-03 Procedure SP-2404AP, which was revised on June 28,1991, was inadequate in that it did not check the right panel for current indications, did not check the sample pump cut-off functions, did not call for verifying that valve AOV 245 went shut in the test position, did not address the associated equipment correctly by proper nomenclature, and the data needed for Note 2, Page 2 should be in epm as the recorder indicates, not in uCi/cc as specified.
Request Please provide your review of the above assertion. If the above conditions are valid, notify us of the corrective actions you have taken to prevent recurrence. Also provide us with an assessment of the safety significance of any identified deficiencies, including generic considerations.
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. i STRiBiu PUB IC DISCLOSURE l ENCLOSURE l
l Issue 91-0220-01 ,
It has been noted that many Air System Valves do not have system identification tags, causing identification confusion, and multiple active Red Tag for the same problem. In particular, during performance of SP-2404AP, Waste Neutralization Tank Radiation Monitor Functional, it was noted that the Instrument Root Valves to valves 2-CND-34 and 2-CND-37 did not have system identification tags, only duct tape hand-written identification marked AOV244A and AOV244B.
Issue 91-0220-03 Procedure SP-2404AP, which was revised on June 28,1991, was inadequate in that it did not check the right panel for current indications, did not check the sample pump cut- ,
off functions, did not call for verifying that valve AOV 245 went shut in the test position, did not address the associated equipment correctly by proper nomenclature, and the dats needed for Note 2, Page 2 should be in cpm as the recorder indicates, not in uCi/cc as specified.
Request:
Please provide your review of the above assertion. Regarding Issue 91-0220-03, why was the procedure inadequate and was this procedure a product of the procedure upgrade program? If the above conditions are valid, notify us of the corrective actions you have taken to prevent recurrence. Also provide us with an assessment of the safety significance of any identified deficiencies, including generic considerations.
t I IMITED-DISIRIR60R_EUBL4C1)lSCLOSURE
SAMPLE RECORD OF ALLEGATION PANEL DECISIONS SITE: S <- PANEL ATTEND :
ALLEGATION NO.: A Md- A -ouf Chairman - de.
DATE: 22A491 (Panel No.(D2 3 4 5) Branch Chief- .
PRIORITY: High Medium @ Section Chief (AOC) - fcJly Sr. Allegation Coord (SACNlm4b i SAFETY SIGNIFICANCE: Yes No Unkn
- CONCURRENCE . OI Reoresentative -
l TO CLOSEOUT: DD BC O (Other) lleimm" L tm ro h CONFIDENTIALITY GRANTED: Yes @ SNI%%j'{
Vemp+I t Mfhyst")
A 4-(See Allegation Receipt Report) i IS THERE A HARASSMENT / DISCRIMINATION
- ISSUE:
l IF YES, Yes h .
i 1) has the individual been informed of the DOL
- process and the need to file a complaint within 30 days
- 2) has the individual filed a complaint h No l with DOL Yes No
- 3) has a letter been sent to the complainant seeking Yes No
- any safety concerns IS A CHILLING EFFECT LETTER WARRANTED: Yes No ;
- IF YES, HAS IT BEEN SENT Yes No l HAS THE LICENSEE RESPONDED TO THE CHILLING EFFECT LETTER
- Yes No ACTION:
- 1) T enooer I5%es 3adi de 1) (1 hd<.ek% auf bown
% \eHerh '
- 2) B d u m A4. QAwi leas bem dext 4 w A 6tbe_
3) i 4) 5)
NOTES: fees;c cover d uwler 83- 88 ooli :S dweni / ou er-Ao (Ib3 od.h. 19 89 t)tscust)cn e w $ C.Whl4e. oSo1 m 2 3 /hd.if . v t,. aq r ced ,
uav.c4 de. acdh w as h km 4we ot. h l At. der I A4-1
r ALLEGATION MANAGEMENT SYSTEM ALLEGATION NUMBER - RI-91-A-0225 RUN DATE: 09/11/91 DOCKET / FACILITY / UNIT. 05000245 / MILLSTONE 1 / 1 DOCKET / FACILITY / UNIT: 05000336 / MILLSTONE 2 / 2 DOCKET / FACILITY / UNIT: 05000423 / MILLSTONE 3 / 3 DOCKET / FACILITY / UNIT: / C. N. FLAGG /
' ACTIVITY TYPES - REACTOR MATERIAL LICENSES -
F,UNCTIONAL AREAS -
OPERATIONS SAFEGUARDS ONSITE HEALTH AND SAFETY
' DESCRIPTION - 1) UNLAWFUL DISCHARGE PER 10CFR50.7 (IN 1987) 2) ONGOING DRUG AND ALCOHOL ABUSE AT MILLSTONE SITE 3) UNSAFE RADIO-LOGICAL PRACTICES RESULTING IN OVEREXPOSURES 4) UNSAFE CONCERNS - WORK PRACTICES ON-SITE (OSHA) 5) NU AND CN FLAGG KNOWLEDGE 5 OF AND ACQUIESENCE IN DRUG AND ALCOHOL USE ON-SITE SOURCE - FORMER CONTRACTOR EMPLOYEE CONFIDENT - NO RECEIVED - 910821 BY - DA DEMPSEY / RI ACTION OFFICE CONTACT - EM KELLY -
(FTS)346-5183 SAFETY SIGNIFICANCE - UNKNOWN BOARD NOTIFICATION - NO STATUS - OPEN SCHED COMPLETION - 911231 DATE CLOSED -
ALLEGATION SUBSTANTIATED - ALLEGER NOTIFIED -
OI ACTION - OI REPORT NUMBER -
REMARKS - ISSUES SAME AS THOSE RAISED IN 1988. PANELED 22AUG91.
ISSUES WERE REFERRED TO FBI IN 1988.
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l SUPPORT OFFICE: RPS-4A ACTION PENDING: REFER TO LICENSEE (DRUG AND ALCOHOL ISSUES) 4 DOCUMENTATION: !
ALLEGER LAST CONTACTED: ;
REFERENCE:
RI-88-A-0014 l KEYWORD: FFD, OSHA, DRUGS, ALCOHOL !
ENTERED SYSTEM - 910823 CLOSED SYSTEM - RECORD CHANGED - 910823 ,
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ALLEGATION MANAGEMENT SYSTEM ALLEGATION NUMBER - RI-91-A-0225 RUN DATE: 09/11/91 DOCKET / FACILITY / UNIT: 05000245 / MILLSTONE 1 / 1 DOCKET / FACILITY / UNIT: 05000336 / MILLSTONE 2 / 2 DOCKET / FACILITY / UNIT: 05000423 / MILLSTONE 3 / 3 DOCKET / FACILITY / UNIT: / C. N. FLAGG /
ACTIVITY TYPES - REACTOR MATERIAL LICENSES -
FUNCTIONAL AREAS -
OPERATIONS SAFEGUARDS ONSITE HEALTH AND SAFETY DESCRIPTION - 1) UNLAWFUL DISCHARGE PER 10CFR50.7 (IN 1987) 2) ONGOING DRUG AND ALCOHOL ABUSE AT MILLSTONE SITE 3) UNSAFE RADIO-LOGICAL PRACTICES RESULTING IN OVEREXPOSURES 4) UNSAFE CONCERNS - WORK PRACTICES ON-SITE (OSHA) 5) NU AND CN FLAGG KNOWLEDGE 5 OF AND ACQUIESENCE IN DRUG AND ALCOHOL USE ON-SITE SOURCE - FORMER CONTRACTOR EMPLOYEE CONFIDENT - NO RECEIVED - 910821 BY - DA DEMPSEY / RI ACTION OFFICE CONTACT - EM KELLY -
(PTS)346-5183 SAFETY SIGNIFICANCE - UNKNOWN BOARD NOTIFICATION - NO STATUS - OPEN SCHED COMPLETION - 911231 DATE CLOSED -
ALLEGATION SUBSTANTIATED - ALLEGER NOTIFIED -
OI ACTION - OI REPORT NUMBER -
REMARKS - ISSUES SAME AS THOSE RAISED IN 1988. PANELED 22AUG91. -
ISSUES WERE REFERRED TO FBI IN 1988.
SUPPORT OFFICE: RPS-4A ACTION PENDING: REFER TO LICENSEE (DRUG AND ALCOHOL ISSUES)
DOCUMENTATION:
ALLEGER LAST CONTACTED:
REFERENCE:
RI-88-A-0014 KEYWORD: FFD, OSHA, DRUGS, ALCOHOL ENTERED SYSTEM - 910823 CLOSED SYSTEM - RECORD CHANGED - 910823 l
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- S UNiTi", STATES
- - NUCLEAR REGULATORY COMMISSION
' ogcion i i'r,'b. -[ f E
,k f,,f 47s AttENoALE sc A:
KING of Po'#% A PENNSY LV A NI A 19406-1415
[
50-245 License Nos. DPR-61 Docket Sm.
DPR-65 50-335 50-423 NFP-49 hir. John F. Opeka Executive Vice President. Nuclear Scrtheast Nuclear Energy Company P. O. Box 270 Hartford. Connecticut 06141-0270
Dear 51r. Opeka:
Subject:
Combined Inspection Nos. 50-245/91-19. 50-336/91-23 and 50-423/91-19 On September 23-28.1991 an announced. safety inspection of your September 26.1991.
emergency preparedness exercise and other emergency preparedness activities was conducted by Str. C. Amato and other members of this office. The inspection was conducted at your hlikone Nuclear Power Station, Waterford, Connecticut, and at your offices in Berlin.
Connecticut. In addition the inspection was continued during the period October 7-30.1991 at the NRC R1 office in King of Prussia, Pennsylvania to permit evaluation of documentation obu ned during the on-site portion of the inspection and to review your corrective actions taken in response to the Unusual Event declared as a result of Hurricane Bob. Discussions of our 6ndings were held by hir. Amato with your staff at the conclusion of the on-site portion of the inspection.
The exercise demonstrated the ability of the hiillstone Station and Corporate staffs to take timely and adequate protective measures on behalf of public health and safety. Corporate staff exhibited excellent response to scenario accident conditions. An adequate emergency preparedness program was maintained. and no exercise weaknesses were identined.
Although no violations were identified, an unresolved item is discussed in Section 7.4.4 of the enclosed inspection report.
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DEC 0 e 199, ;
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Northeast Nuclear Energy Comoany
.No repi ;. to t his letter is required. Thank you for your cooperation. I
. Sincerely, A ]
W y l es H. Jo er, 'hief .-
Facilities Radio ogical Safety
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and Safecuards Branc!-
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' l Division of Radiation Safetv .,
t and Safeguards
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Enclosure:
50-245/91-19,50-336/91-23 and 50-c.'.91-19 Combined Inspection Report Nos. f i
cc w/ encl:
W. Romberg, Vice President. Nuclear Operations i D. Nordquist, Director of Quality Services j R. Ka:ich. Manager. Nuclear Licensing :
S. Sca:e. Nuclear Station Director !
H. Haynes, Nuclear Unit Director [
C. Clement. Nuclear Unit Director. ,
G. Garfield Esquire j N. Reynolds. Esquire i K. Abraham, PAO O ['
Public Document Room (PDR)
Local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
^
' SRC Resident Inspector t State of Connecticut i 1
U. S. FEMA-I
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%:meast Nuclear Energy Company 3 DEC 0 t gg, j i
bec w/ encl: ,
' Region 1 Docket Room (w/ concurrences) 3 Management Assistant, DRMA (w/o encli >
J. Williams. PM, NRR l E. Wenzinger, DRP , i E. Kelly,' DRP ;
W. Raymond, SR1, Millstor.e I A. Asars SRI, Haddam Neck ;
R. Arrighi. DRP 'i R. Lobel, EDO
t DRSS SALP Coordinato
J. Joyner. DRSS l
RI-91-A-0219 '!
RI-91-A-0128 t RI-91-A-0228 *l P RI-91-A-0046 ,
RI-91-A-0236 !
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U. S. NUCLEAR REGULATORY CONIAllSSION REGION 1 l 50-245/91-19 l Report Nos.
50-336/91-23 l 50-423/91-19 Doci:et Nos. ~ 50-245 50-336 50-423 License Nos. DPR-61 '
DPR-65 NPF-49 Licensee: Northeast Nuclear Energy Company P. O. Box 270 ;
Hartford, Connecticut 06101-0270 Facility Name: Stillstone Nuclear Power Station Inspection Conducted: September 23-28,1991 and October 7-30.1991 ,
Inspection At: Berlin, Connecticut; Waterford. Connecticut: and Kinc of ;
Prussia, Pennsylvania Inspee: ors: O, M. 8MM .~ //['7/'f7 C. G. Amato. En/ergency Prepar_edness date Specialist. Region I i
D. Dempsey Resident Inspector, Millstone Point Plant ;
P. Habighorst, Resident Inspector, Millstone Point Plant K. Ihnen, Operations Engineer (Examiner) !
J. Jamison, NRC Contractor l K. Kolaczyk, Resident Inspector, Millstone Point Plant W. Raymond. Senior Resident Inspector, Millstone Point Plant )
G. Vissing, NRC Licensing Project Manager )
Appros ed: f M v!2)!$
E. NicCag. Nief. Emergency / date ;
Preparedness Section. Division i of Radiation Safety and Safeguards I l
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Areas inspected: Safety inspection of the licensee's full-participation emergency preparedness exercise, and of the emergency preparedness program. Program areas inspected included Emergency Plan and implementing Procedure changes, training, audits and reviews. maintenance of emergency response facilities, public information, off site activities, and responses to situations requiring emergency classification.
Results: No exercises weaknesses were identified. The exercise demonstrated the licensee's i ability to take timely and adequate protective measures on behalf of public health and safety.
Corporate staff exhibited excellent response to scenario accident conditions. An adequate emergency preparedness program was maintained. No violations were identified. An unresolved item. insolving the extent of training given to operators on the simulator that includes classifying scenarios at the Site Area Emergency and higher levels, is discussed in Section 7.4.4 l
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C-4 24 as no more than 16 consecutive hours on duty, personnel are expected to report for emergency and overtime, and fitness for duty applies, including fatigue. e.g. anyone called after working overtime may claim they are not fit for dut:. due to fatigue.
Nuclear Engineering and Operations Procedure (NEO) NEO-90-G-292. dated 10/25/90, established policies that make supervisors and managers responsible for the response of individuals on the oncall list.
Millstone station maintains two lists one for emergency response and one for station support. Oncall staff may exchange with other staff members either formally or informally. Informally this is accomplished by passing on a pager while still maintaining responsibility. Formally this is accomplished by completing a form and transferring the oncall responsibility. Maintenance of the oncall roster is the lead supervisor's or manager's responsibility. However, this practice deviated from the commonality policy stated in the NEO procedure in that Haddam Neck maintains a single oncall list and permits only formal transfer of oncall responsibility.
The inspector reviewed a draft of "Oncall Procedure", EP-4211, Revision 15. This ,
draft document changes responsibility for maintenance of the oncall list to the SSNEPC. SERO members will be required to forward changes to the oncall schedule to the SSNEPC. The procedure continues to maintain the informal transfer method and assigns responsibilities to seven groups of station staff.
Call-ins are accomplished by the use of a recently installed electronic system. the Emergency Noti:lcation and Response System (EN&RS). The inspector observed several test and exercise uses of this system, which provides both voice and hard copy messages, and concluded it is efficient. Pagers are tested daily, with the signal sent from five towers located throughout the NU service territory.
Based upon the above review this portion of the licensee's program is acceptable.
9.0 Licensee Response to Actual Events The inspectors reviewed reports of licensee response to actual events to determine if the licensee's actions were consistent with their emergency plan and implementing procedures. Since the last inspection,13 Unusual Events (UEs) were declared. Of these, three were declared at Unit 1, two at Unit 2, seven at Unit 3, and one was a Millstone site declaration. The UEs all were correctly classified and notifications were timely. With the exception of the site UE, response actions were as expected. The site UE was declared during Hurricane Bob and was not de-escalated until the next day.
During the Hurricane the following events took place.
- The EOF was powered by the Emergency Diesel Generator (EDG).
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- Meteorological capability was lost.
- The ability to transmit emergency information among ERFs was lost.
As a precaution, the EOF management during the hurricane, decided to isolate the EOF from the Flanders transmission line. The EDG was cut-in and the load was picked up without interruption. An NRC Senior Resident inspector observed this operation. During th'e annual exercise, the licensee tested the EOF EDG. The EDG carried the EOF loads for more than half of the exercise before the loads were shifted manually to the normal transmission source. The inspectors observed the operation of the EDG at this time.
During the Hurricane, an anemometer cup on the 143 foot level of the Meteorological (MET) tower blew off. Peak wind speeds up to 30m/sec were recorded. These speeds are 15 minute averages. so wind gusts were not recorded. The MET tower receives power from the Flanders transmission line. This line subsequently was lost.
and the backup power supply did not maintain MET loads. As a result, MET data were not available. Following the hurricane, a Plant Information Report (PIR) was written to determine why the MET tower backup power supply did not operate.
As a result of previous power failures to the MET tower, a propane powered generator was installed in 1987. This system has failed four times since it wa installed. This system is located next to the MET tower, with the controls in a shed.
The control equipment was in excellent condition. however the generator showed deterioration due to weather. Records show that the maintenance department performs monthly, semi-annual, annual and three year preventative maintenance.
Nine corrective maintenance work orders have been written for this system.
Generally, the testing performed following the maintenance is not indicated.
Consequently, the testing of the automatic transfer equipment can not be determined.
The following was determined regarding maintenance and testing of this system.
- No one group has clear responsibility for this equipment.
- Preventative maintenance is being performed. However the records do not indicate that the equipment, particularly the automatic transfer equipment. is tested.
- There is a weekly automatic run of the equipment. However, no one interviewed by the inspector was sure what this run tests (engine or engine and generator).
- There was a field change made to the equipment (low level coolant trip removed). This change was not reflected on any drawings associated with the equipment.
- There have been no procedures prepared for the equipment, including operating, preventative or corrective maintenance procedures.
Responsibility for maintenance has subsequently been transferred from Unit : to
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UNITED STATES
'p .. *
, NUCLEAR REGULATORY COMMISSION.
l2* !! REGION I -
475 ALLENDALE ROAD l e, [ KING OF PRUSSIA. PENNSYLVANIA 19406 '415
..... j Docket No. 50-336 FEB 24 E :
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Mr. J. Opeka t Executive Vice President - Nuclear l Northeast Nuclear Energy Company ;
i P.O. Box 270 Hartford, Connecticut 06141-0270 ] .
Dear Mr. Opeka:
Subject:
NRC Region I Inspection Report No. 50-336/91-31 Mr. J. T. Shedlosky and others of this office conducted a special safety inspection December 17, l 1991, through February 7,1992, at the Millstone Nuclear Station Unit 2, Waterford, :
Connecticut. The inspection results are documented in the enclosed _ report. They were discussed with Mr. J. S. Keenan and other members of your staff at the conclusion of the inspection. j Areas examined during the inspection are described in the enclosed report. Within these areas, _
j the inspection focused on issues brought to Northeast Utilities by the NRC. Our indeo:ndent i
review evaluated your performance in complying with regulatory requirements importat 9 public and worker health and safety. This review consisted of performance observations of l ongoing activities, inspection of plant equipment, interviews with personnel, and review of records. ;
Our overall assessment was that performance was acceptable. The enclosed inspection report notes a number of issues on which your staff agreed to provide a response to the NRC. ;
NNECO's response to the NRC may be made in communication with the resident inspectors.
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 j this letter are not subjected to the clearance procedures of the Office of Management and Budget J as required by the Paperwork Reduction Act of 1980, Public Law No. %.511.
Your cooperation with us is appreciated.
S' ly, 3 s.
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Edward C. g r, Chief Projects Branch No. 4 V Division of Reactor Projects i
- D 0 3 9 S. 0 2 O b'/
FEB 241992 Northeast Nuclear Energy Company 2
Enclosure:
NRC Region I Inspection Report No. 50-336/91-31 cc w/ enclosure:
. W. D. Romberg, Vice President, Nuclear Operations D. O. Nordquist, Director of Quality Services
R. M. Kacich, Manager, Nuclear Ucensing S. E. Scace, Nuclear Station Director, Millstone J. S. Keenan, Nuclear Unit Director, Millstone Unit 2 Gerald Garfield, Esquire _
Nicholas Reynolds, Esquire K. Abraham, PAO (2)
Public Document Room (PDR) 1.a:al Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC) i NRC Resident Inspector State of Connecticut 1
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i Northeast Nuclear Energy Company - 3 .
bec w/ enclosure: ,
Region I Docket Room (with concurrences) [
Management Assistant, DRMA (w/o encl) l E. Wenzinger, DRP E. Kelly, DRP W. Raymond, SRI, Millstone J. Shediosky, SRI, Haddam Neck R. Ihl, OEDO G. Vissing, PM, NRR ;
R. Arrighi, DRP R. Fuhrmeister, REG I ;
R. Barkley, DRP J. Joyner, DRSS bec w/ Report Cover Sheet and Executive Summary only:
C. Hehl, DRP ,
J. Wiggins, DRP :
I W. Hodges, DRS J. Durr, DRS ,
L. Bettenhausen, DRS l J. Stolz, NRR/PDI-4 .
M. Knapp, DRSS W. Imning, DRSS ,
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1 U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No.: 50-336/91-31 License No.: DRP-65 Licensee: Nonheast Nuclear Energy Company P.O. Box 270 Hartford, CT 06141-0270 Facility: Millstone Nuclear Power Station, Unit 2 i
Iocation: Waterford, Connecticut Inspection Dates: December 17,1991 - February 7,1992 Inspectors: T. R. Fredette, Consultant, AMSEC/SAIC C. M. Meeker, Consultant, COMEX/SAIC P. L. Reagan, Consultant, COMEX/SAIC E. L. Conner, Reactor Licensing / Risk Engineer, i
Technical Support Section, DRP i
Supervisor: J. T. Shedlosky, Senior Allegation Coordinator, Reactor Projects Section No. 4A, ;
Division of Reactor Projects Approved by: [ lo 2!/2!f2 l 4r Eugene Kelly, ddef' / Date Reactor Projects Section 4A i Scope: Special inspection of concerns brought to Nonheast Utilities by the NRC. This report is a continuatice of the special !=5+I-s described in NRC Inspection Report numbers 50-245/91-23 and 50-336/91-27, and 50 336/91-29. It includes plant operations, maintenance and surveillance, and engineering and Whahl suppon issues.
Inspection Results: See Executive Summary i
EXECUTIVE
SUMMARY
MILISIONE UNIT 2 INSPECTION 91-31 Here are several observations relata! to the consistency of interpretation and implementation of quality assurance program requirements, and attention to detail in procedural compliance.
Our overall assessment is that performance is adequate, but we identified weaknesses in several areas listed below:
i PLANT OPERATIONS Management expectations were not always clearly defined and communicated to workers.
Examples included the conduct ofindependent verification activities, qualification requirements for independent verifiers, prerequisites for closure of short form PDCRs, maintenance of instrument calibration data sheets, values for alarm setpoints in procedures, and administration of the work control center.
A strength was NNECO's practice of controlling unnecessary traffic in the control room by effectively using a work control center. i MAINTENANCE AND SURVEIII ANCE i The following were additional examples of weaknesses described in previous inspection reports:
o Biennial review of procedures was not always accomplished in a timely manner, o Supervisors did not always ensure technicians and maintenance personnel had the requisite qualification for assigned tasks.
o Housekeeping was inadequate in some plant areas because unanchored material was stored near safety related components and could detrimentally affect operability of those components during a seismic event.
Opportunities for personnel to improve attention to detail exist, such as the validation of some j valve line up sheets, one instance in which workers did not appropriately transition to a three page work order, and one instance in which changes to a surveillance procedure were not I identified in a timely manner. l ENGINEERING AND TECHNICAL SUPPORT There was inadequate maintenance of some as-built drawings for the Emergency Operations !
Facility Emergency Diesel Generator. Also, in two instances there was inadequate attention to detail in design control. Further, the NRC questioned the adequacy of NNECO's programmatic evaluation of non-seismically qualified instrumentation in seismic category I systems.
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TABLE OF CONTENTS i
I 1.0 WORK CONTROL ...................................... I ;
1.1 Control Room Air Conditioning System . . . . . . . . . . . . . . . . . . . . . . I 1.2 Radioactive Waste Gas Compressor . . . . . . . . . . . . . . . . . . . . . . . . 3 ;
2.0 SERVICE WATER SYSTEM PDCR 2-057-90 . . . . . . . . . . . . . . . . . . . . . . 4 4
3.0 RADIATION MONITOR DRAWINGS . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1 PDCR M2-90-032 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 IAC Procedures for Radiation Monitoring Equipment 6 3.2 .............
4.0 SURVEILLANCE PROCEDURE SP 240lR . . . . . . . . . . . . . . . . . . . . . . . 7 5.0 HIGH RANGE STACK GAS MONITOR . . . . . . . . . . . . . . . . . . . . . . . . . 3 5.1 Status of Procedures for HRSGM I&C Work . . . . . . . . . . . . . . . . . . 9 5.2 Conduct of Surveillance Testing ......................... 10 10 6.0 INSTRUMENT ROOT VALVE LABELS . . . . . . . . . . . . . . . . . . . . . . . .
6.1 Instrument Air Root Valve labels . . . . . . . . . . . . . . . . . . . . . . . . 10 6.2 Instrumentation Valve Line-up . . . . . . . . . . . . . . . . . . . . . . . . . . I1
- 6.3 Qualification of Personnel to do Valve Line-up Work . . . . . . . . . . . . 13 ,
Housekeeping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 6.4 15 7.0 EOF DIESEL GENERATOR ............................... ,
7.1 Status of Procedures for EOF-EDG Work . . . . . . . . . . . . . . . . . . . 16 ,
EOF-EDG 1990 Annual load Test . . . . . . . . . . . . . . . . . . . . . . . . 17 7.2 ,
7.3 As-Built Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 19 7.4 Qualification of Personnel to Perform EOF-EDG Work . . . . . . . . . . .
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8.0 SEISMIC QUALIFICATION OF HPSI SUCTION PRESSURE GAUGES . . . .
8.1 8.2 PI Root Valve Position . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 PDCR 2-89-046 Implementation Status . . . . . . . . . . . . . . . . . . . . . 23
'J 9.0 ADMINISTRATIVE CONTROL OF RADIATION MONITOR MAINTENANCE ...................................... 25 10.0 STATION BA'ITERIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 27 11.0 NNECO RESPONSIVENESS TO EMPLOYEE CONCERNS . . . . . . . . . . . .
12.0 EDG CLEAN WASTE TANK PDCR MP-2-90-035 ................. 28 i
29 13.0 RECORDER CALIBRATION METHODOLOGY . . . . . . . . . . . . . . . . . . .
31 14.0 SAFETY INJECTION TANK PRESSURE SWITCHES . . . . . . . . . . . . . . .
15.0 WORK CONTROL CENTER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 l
16.0 NONCONFORMANCE REPORT 291-272 . . . . . . . . . . . . . . . . . . . . . . . 35 17.0 SUPPLEMENTARY INFORMATION ON PRIOR INSPECTION ISSUES ... 36 17.1 Drawing Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 17.2 Tool and Document Contamination During AWO M2-91-06732 . . . . . . 36 17.3 ASI Curve for LHGR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 17.4 Steam Generator level Calibration Procedure SP 2402D . . . . . . . . . . 38 l
17.5 Rigging Practices for Two Ton Hoist ...................... 39 18.0 M ANAGEMENT MEETINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 l
ATTACHMENTS
- 1. PERSONS CONTACTED
- 2. REFERENCES l
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DETAILS .
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1.0 WORK CONTROL :
i Concems have been identified regarding work control procedures in the I&C Department at j Millstone Unit 2. ,
1.1 Control Room Air Conditioning System
' - One calicern involved the troubleshooting efforts associated with the ventilation heater control !
circuit in the Control Room Air Conditioning (CRAC) system. 'Ihe CRAC heaters could be !
energized when the associated fan was not running.
Assessment I
The inspector interviewed the personnel involved with the troubleshooting effort on the CRAC system and reviewed the documentation used as well as the pertinent governing plant procedures ,
and directives. The following information was obtained: i An Automated Work Request (AWO) was generated (M2-91-06068) to test the operation of the !
flow switches in the CRAC ventilation heater control circuit. Imcal ID X60A was used on the !
AWO. The switches tested satisfactorily, so the problem was assumed to be with the design of i
the installation. The problem with the flow switches was identified around June 17,1991, and a memorandum from a technician to the I&C Manager described the problem and recommended i a change in the sensing points of the switches to allow them to function properly. i AWOs, M2-91-06744/06745, were issued to investigate the flow switch probier , but different ,
pressure switches were used for the Imcal ID on the work requests. 'Ihis action was contrary to the Caution statement in ACP-QA-2.02C, " WORK ORDERS," paragraph 6.2.1, which applies to the PMMS Planner or Authorized Person in Lead Department and states, "Many of the Quality Indicators, Cautions and Other Information which appear on the AWO are automatically filled in by the PMMS computer based on the local ID which is used. For this '
reason, it is important to select a local ID that properly represents the equipment to be maintained. If the ID used is for a procedure, then the nuclear indicators must reflect the QA 1 status of the equipment affected by the procedure." No local ID was assigned in the PMMS ;
system for the flow switches associated with the X60A/B heaters, so the I&C maintenance personnel incorrectly used the local ID of closely related components for the Control Room Air Filter Inlet D/P switches (PDS-8347/8348). As a result, the QA status was incorrectly indicated ;
on the AWOs. 'Ihe job description stated, " Evaluate for new design. Investigate changing ,
sensing points of flow switches on Control Room HVAC heaters to prevent them from :
inadvertently turning on.' 'Ihis was the correct reason for the AWOs, but the incorrect local ID identified the troubleshooting effort on another part of the CRAC system.
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To properly evaluate a change to the sensing location of the flow switches, a jumper device, BJ. ,
91-57, was proposed, properly eygce:d by PORC ~ in accordance with established administrative procedures (ACP QA-2.06B) on October 8, and installed on October 30,1991. i ne purpose of the jumper was to test the effectiveness of shifting the sensing point of the flow ,
switch to the inlet plenum of the associated fan. ,
De test of the flow switchsin the X60A heater control circuit was first performed in the moming 4
of October 30,1991, and the results indicated that a shift in the sensor position did not change l the performance of the switch. De flow switch still did not prevent the heater from being !
turned on when the eew fan was recured. This testing and the subsequent removal of the l flow switch were accomplished under the AWOs that used the Inlet Filter D/P switches (DPS- ,
8347/8348) for the Local ID. At this point, new AWOs were generated to support the subsequent tmubleshooting effort that was expected. De new AWOs, M2-91-11519/20, used J the appropriate X60A/B for the local ID and included directions in the Job Description that "A J three-PAGE AWO will be needed for Repair / Modifications." j At the completion of the initial testing and the removal of the flow switch, an I&C technician ]
told the Control Room watchstanders that the CRAC system may be inoperative due to the -
removal of the flow switch and the Control Room watchstanders initiated an LCO action statement (LCO 3.7.6.1) at 10:55 a.m. that day, in accordance with ACP-QA-2.02C, paragraph 5.10. Since that was the first time any work was performed on the CRAC system, no LCO !
action statements were appropriate or required prior to October 30, 1991. However, the ;
afternoon Control Room watchstanders on October 30,1991, incorrectly determined that the i switch provided no important function; therefore, there was no need for the LCO action statement and they lined out the entry. Although the CRAC duct heaters are not safety related, the ventilation system pressure boundary requires QA controls. System operability evaluations ,
are required prior to work breaching this boundary. !
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Subsequent work progress / planning meetings in the I&C Department identified the errors associated with the incorrect use of a two-page AWO, because the work on the CRAC system l heaters required QA procedures. De error in deleting the LCO action statement was also j identified and corrected at that time. A Plant Incident Report (PIR 91-118) was initiated and -4 a proper three-page AWO (M2-91-11622) was issued to complete the troubleshooting effort I associated with the X60A heater centrol circuit, remove the test equipment, and return the i system to normal. De new AWO also identified the .yywydate LCO action statement j associated with the work. No work or troubleshooting had been performed on the X60B heater )
control circuit at that time, so the required corrective actions were only associated with the I X60A heater control circuit. j l
PIR 91-118 designated the cause of the incident to be personal error on the part of the I&C l i
engineer because he had assumed that the system was out of service. He PIR listed the wrong local ID as only a contributing cause, ne failure of the PMMS Planner /I&C Department l authorized person to heed the caution in paragraph 6.2.1 of ACP-QA-2.02C was not even j mentioned in the PIR. De fact that the Shift Supervisor / Senior Control Operator is responsible 1
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for LCO requirements, in accordance with ACP-QA-2.02C, paragraph 5.10, not an I&C engineer, was also not included in the PIR.
The CRAC ventilation heater flow switch, which is a common commercially used switch and .
sub-component of the duct heaters, is not identified in the PMMS system and, therefore, is not ,
classified by the Material Equipment Parts List (MEPL).
Conclusion j The original ventilation heater troubleshooting AWOs, M2-9106744/06745, were incorrectly _
prepared. As a result, the workers failed to properly transition over to a three-page, QA )
orientated, AWO during the course of the troublMag effort. However, the workers did l properly carry out the work described in the AWO (i.e., investigate changing sensing points of l flow switches on Control Room HVAC heaters to prevent them from inadvertently turning on). !'
Therefore, the supervisory review was at least adequate to make sure that the correct work was accomplished. In addition,140 action statement implementation errors would most likely have been avoided, if the yymydate Imcal ID were used in the AWOs. In fact, the correct equipment was tagged out. Both the LCO implementation errors and the AWO procedural errors were promptly identified by NNECo and appropriately corrected; there was no safety significance to this event.
1.2 Radioactive Waste Gas Compressor A second concern involved the work control procedures used to replace a failed suction pressure .
switch for the radioactive Waste Gas Compressor, "FI A."
Assessment i
The inspector interviewed the I&C technicians involved in the replacement of the suction
, pressure switch for the radioactive Waste Gas Compressor, "FIA," and reviewed the ;
documentation associated with the replacement. The following information was obtained- !
After verifying that the installed pressure switch was defective, the I&C technicians obtained a !
replacement switch based on the information contained on the pressure boundary cover that was i installed on the switch. When the replacement switch arrived, it was noted to be physically i different from the failed switch and would not adjust to the required trip point. The I&C !
technicians initially procured the wrong replacement switch, because they mistakenly used the .
identification data on the pressure boundary cover. 'Ihe I&C technicians went to the associated l Loop Folder and requisitioned the listed part, which was in stock in the warehouse. This switch i was bench calibrated and satisfactorily installed.
i The Loop Folder was updated with additional schematics showing the switch and its appropriate setpoint. 'Ihe I&C technicians verified and reinstalled the pressure boundary covers on the appropriate suction and discharge pressure switches.
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Conclusion When the replacement switch identification error was detected, the technicians referred to available documentation and obtained the correct switch. De technicians made an initial i
procurement mistake, but correctly resolved the problem. De incorrect switch was never installed in this non-safety.related system.
2.0 SERVICE WATER SYSTEM PDCR 2-057-90
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The NRC provided a concern related to a plant design change aquest (PDCR) for installation l of a sample valve and "T" fitting on the service water (SW) strainer differential pressure (D/P) ,
instrument tubing at MP2. For reference, Inspection Reports 50-336/91-20 (section 6.2) and 50- '
336/91-28 (section 5.4) described the NRC's review of NNECO's MP2 SW system biofouling protection program. ,
l A ssessment The SW system is a safety related fluid system that supplies cooling water (sea water) to the I
emergency diesel generators, the reactor building closed cooling water system heat exchangers, and the turbine building closed cooling water system heat exchangers. There are three 50%
capacity SW pumps rated at 12,000 gpm each. At the discharge of each pump, there is a 24" automatic self-cleaning strainer. Each strainer has D/P monitoring instrumentation that initiates ,
strainer backwash and provides a differential pressure alarm. l The purpose of (short form) PDCR 2-057-90 was facilitation of SW system sampling for chlorine content. Implementation and testing of PDCR 2-057-90 was done per work order AWO .
M2-90-15617. To reduce biological fouling, a sodium hypochlorite system chlorinated the SW system. PDCR 2-057-90 involved addition of a tee fitting, a 1/4" Whitey valve, and associated instrument tubing in the 1/4" SW strainer D/P instrument tubing for each strainer. The new sample valves were numbered 2-SW-276A, B, and C. PDCR 2-057-90 included a safety '
evaluation that concluded "this design change is safe and does not constitute an unreviewed safety question as defined in 10CFR50.59." ,
i MP2 Technical Specification 3/4.7.4 required SW operability in modes 1,2, 3, and 4. Prior to work approval, as required by ACP-QA-2.02C, section 6.5, the Senior Control Room !
Operator (SCO) determined installation of PDCR 2-057-90 in accordance with AWO M2 15617 did not affect SW operability. ,
The test plan in PDCR 2-057-90 required in-service leak testing (which was done on December 13,1990) for the completed installation of sampling valves 2-SW-276A and B. Because the "C" SW strainer was unavailable at that time, NNECO deferred in-service leak testing of sampling valve 2-SW-276C. ;
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ACP-QA-2.02C, revision 25, section 6.12, assigned responsibility for overall implementation I
and testing of plant modifications to the Plant Engineer. ACP-QA-2.02C, section 6.12.5, required (in part) that before a system or component may be declared operable and accepted by
- Operations, the Plant Engineer must verify that applicable administrative requirements of ACP-QA-3.10 have been met. .
ACP-QA-3.10, Attachment 1, section 6.1, required (in part) completion of " applicable i
administrative impact items" and evaluation of other required administrative updates such as the examples listed on a PDCR (Form B)." Although not listed as an " administrative impact item"
. or " administrative update," Item 13D in Form B required completion of construction and pre-operational testing.
L NNECO closed PDCR 2 057-90 on January 19, 1991. NNECO completed in service leak d testing of 2-SW-276C on August 13, 1991 and closed AWO M2-90-15617 on September 12,
- 1991.
Conclusions Rawd on discussions with cognizant NNECO personnel, physical inspection of the equipment >
- in question, and review of relevant documentation, the inspector concluded the following:
because NNECO closed PDCR 2-057-90 prior to completion of the testing specified in AWO M2-90-15617, in this instance there was inadequate attention to detail in compliance with applicable administrative control procedures. Further, requirements for completion of all
- necessary work documents and testing were not clearly promulgated by ACP-QA-3.10 for short i
form PDCRs. NNECO agreed to evaluate opportunities for improvement in ACP-QA-3.10 (that would help ensure completion of work documents and testing prior to PDCR closures), take
- appropriate action as e=y, and respond to the NRC.
Also, based on the prmwiing assessment and review of the PDCR 2-057-90 safety evaluation, the inspector concluded that closure of PDCR 2-057-90 prior to closure of AWO M2-90-15617 l did not materially affect SW system operability.
i 3.0 RADIATION MONITOR DRAWINGS .
The NRC provided concerns to NU related to certain procedures and a plant design change !
(PDC) for radiation monitoring equipment at MP2. NRC disposition of those concerns involved providing the concerns to NNECO for review and resolution, with subsequent NRC evaluation to ensure the adequacy of NNECO's actions. NNECO letter A09559, dated August 9,1991, described NNECO's review of those concerns.
3.1 PDCR M2-90 432 l During implementation of PDCR MP2-90032 in early 1991, NNECO found some discrepancies between drawing 25203-39092, sheet 14C, and the as-built status of equipment. 6 I
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- Inspection Reports 50-245/91-23 and 50-336/91-27 (IR 91-27), section 7.2, identified a similar concern regarding incorporation of design changes in drawings for radiation monitoring I equipment. IR 91-27 concluded, in part, that there were walma== in coordinating vendor information into controlled drawings. Also, refer to the drawing control discussion in the
" Supplementary Information On Prior Inspection Issues" section of this report.
Assessment ;
Plant design change record PDCR MP2-90-032 documented replacement of Magnahelic flow indicating switches with Photohelic switches for FIS-8011, 8123, 8132, 8145, 8262, 8434, and 9095. Those switches provided indication and control of flow to their associated radiation monitor.
With NNECO assistance, the inspector used the Generation Records Information Tracking System (GRITS) to determine if there were outstanding design change requests (DCRs) or design change notices (DCNs) against drawing 25203-39092, sheet 14C. Drawing 25203-39092, sheet 14C, was marked "as built" per DCR M2-P-2-21-77. According to GRITS, NNECO initiated
' DCR M2-P-0089-91 on August 15, 1991, to document the =*"ary drawing changes. Also, GRITS showed there were other DCRs and DCNs initiated in mid-1991 related to radiation monitoring equipment. For example, DCRs M2-P-0081-91, M2-P-0089-91, and M2-S-1031-91, and DCN DM2 P-0021-91 affected drawing 25203-39092, sheet 14E. Because DCR M2-P-0089-91 documented the *unty drawing changes, the inspector had no further questions regarding drawing 25203-39092, sheet 14C.
Conclusions Based on discussions with cognizant NNECO g6onnel and review of relevant documentation, the inspector concluded that NNECO's process for identifying and resolving drawing discrepancies adequately documented the differences between drawing 25203-39092, sheet 14C, and as-built conditions.
3.2 I&C Prweedures for Radiation Monitoring Equipment The inspector reviewed the status of various procedures, related to the components affected by PDCR MP2-90032, used to do I&C work on radiation monitoring equipment.
Background
Maintenance procedure IC 2422B was used to do I&C work on gaseous radiation monitors (RMs), including RM 8134B and RM 8145B. IC 2422D was used to do I&C work on particulate RMs, including RM 8132A, RM 8145A, and RM 8434A. RM 8132B was an MP2 Technical Specification required gaseous monitor that was calibrated using surveillance
A concern was that Photohelic flow switches were not calibrated with their associated RM. MP2 I RMs typically had a particulate monitor (e.g., RM 8132A), gaseous monitor (e.g., RM 8132B),
a Photohelic flow switch, and a sample pump arranged in series. Thus, flow switch calibration i was m=ey during calibration of either the particulate monitor or the gaseous monitor, but ;
not with both monitors. . ,
Assessment The inspector did a general review of Instrumentation and ' Control (I&C) maintenance procedures IC 2422B and IC 2422D, and zveillance procedure SP 2404AF. SP 2404AF included calibration of flow indicating switch (FIS) FIS 8132. IC 2422B included calibration of FISs 8145 and 8434 Biennial reviews of IC 2422B, IC 2422D, and SP 2404AF were overdue. IC 2422B and IC 2422D were in the process of being upgraded to the new procedure format. NNECO prepared ,
a draft revision of SP 2404AF in December 1991 that incorporated a biennial review, and the - ,
procedure was in the PORC review process. ACP-QA-3.02D, section 6.1.1, required a periodic, systematic review of Station Procedures specified by ACP-QA-3.02. ACP-QA-3.02, section 6.2.3, included 2400 series SP or IC procedures. In a quarterly memorandum (MP 918), dated November 1,1991, Document Services identified the last biennial review date for IC 2422B and SP 2404AF as December 1,1989, and July 1,1989, for IC 2422D. Biennial reviews were due within two years from the prior biennial review date.
Conclusions SP 2404AF and IC 2422D adequately described calibration of FISs 8132, 8145, and 8434. The biennial review ofIC 2422B, IC 2422D, and SP 2404AF was not completed in a timely manner as required by ACP-QA-3.02D. NNECO was aware of the need to complete the biennial review of these procedures and had action in progress to complete this activity. Additional examples of overdue biennial reviews were described in Inspection Report 50-336/91-29 and elsewhere in this report.
4.0 SURVEILLANCE PROCEDURE SP 2401R Testing of the control element assembly (CEA) withdrawal prohibit (CWP) function was done i using surveillance procedure (SP) SP 2401R. Formerly, CWP testing was done using SP 2401F and SP 240lJ. 'Ihe NRC provided a concern related to the qualification of technicians assigned to do SP 2401R in mid-1991.
Assessment Previous NRC inspection reports described the results of NRC inspections related to the qualification of NNECO personnel to do maintenance and surveillance activities. Inspection Report numbers 50-245/91-80,50-336/91-80, and 50-423/91-81 (IR 91-80) described the results of an Integrated Performance Assessment Team (IPAT) inspection donc during July 1991. IR 91-80, section 3.2, found fin part) that maintenance and I&C personnel " appeared to be well trained in conducting the observed activities." Inspection Report number 50-336/91-29 (IR 91-
- 29) was a special safety inspection of issues brought to NNECO by the NRC. IR 91-29, section 2.0, described the NRC's general assessment of I&C technician qualification.
SP 2401F and SP 2401J had specific qualification requirements for the I&C technicians assigned to do these surveillance tests. Acting as a two person team, the technicians in this instance had l the requisite qualification to do SP 2401F and SP 2401J, as shown on the Individual Qualification Matrix (IQM).
I The IQM had no specific qualification requirement for SP 2401R. The inspector questioned if specific training and qualification were required for SP 2401R. NNECO agreed to evaluate the need for specific training and qualification for I&C technicians to do SP 2401R, take appropriate action as neceuary, and respond to the NRC.
Conclusions Based on review of applicable procedures and discussion with cognizant NNECO personnel, the inspector concluded NNECO may not have adequately ensured the I&C technicians completed required training and were formally qualified prior to performing SP 240iR, as required by ACP-QA-8.27, section 5.3.1 However, since the I&C technicians were qualified to do SP 2401F and SP 2401J, and testing of the CWP function was formerly done using those SPs, the inspector had no further concerns regarding this issue.
5.0 HIGH RANGE STACK GAS MONITOR The High Range Stack Gas Monitor (HRSGM) is a system for sampling particulate and iodine to measure high range post-accident gaseous releases from the MP2 vent stack. Sample flow is monitored using two geiger-mueller (GM) detectors (RM 8168A and B). GM detectors (RM 8168C, D, and E) also monitor three separate filter assemblies. The HRSGM is designed to alarm if setpoints are exceeded.
Prior NRC inspection of the HRSGM system was documented in NRC Inspection Report 50-336/91-19 (IR 91-19). IR 91-19 concluded the I&C Department implemented a very good program to calibrate the effluent and process radiation monitors. Also, IR 91-19 concluded there was excellent management support to maintain the radiation monitoring system integrity and operability.
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9 NRC Inspection Report 50-245/91-23 and 50-336/91-27 (IR 91-27), section 7.0, described a number of concems regarding incuryvr. tion of design changes into dawings and the accuracy of as-built conditions shown in drawings. As documented in IR 91-27 and Plant Incident Report (PIR) 91-65, there were discrepancies in drawings for RM 8132 and RM 8168. Additionally, IR 91-27, section 8.0, described the Plant Operations Review Committee (PORC) review of surveillance procedure SP 2404AR. IR 91-27 concluded the procedure validation process I appeared to require additional emphasis to detect and correct errors prior to the approval Process.
5.1 Status of Proceduns for HRSGM I&C Work The inspector reviewed the status of various procedures used to do I&C work on the HRSGM.
Insoection Findings The inspector identified and did a general review of two surveillance procedures (SP), SP j 2404AR and SP 2404AS, that applied specifically to the HRSGM. NNECO had previously !
revised SP 2404AR, based on the procedure upgrade program, but had not yet upgraded SP l 2404AS. NNECO stated its intent is to upgrade all such PORC approved procedures by the end l of 1992.
l The inspector reviewed documentation of surveillance 2404AR-1 that was done December 23, 1991. SP 2404AR, section 6.2, is a source check that includes a comparison of RIC-8168, !
Integrated Computer System (ICS) display, and recorder RC-101C readings. Data sheet I&C !
Form 2404AR-1, section 6.2.2, Meter Cross Checks, records "as found" and "as left" values for RIC-8168, ICS display, and RC-101C. Values for RIC-8168, ICS display, and RC-101C are 5.4 E-03,5.2 E-03, and 5.0 E-03, respectively, which satisfies the acceptance criteria.
The biennial review of SP 2404AS was overdue. I&C management initiated AWO M2-90-14507 to do this biennial review. ACP-QA-3.02D, section 6.1.1, requires a periodic, systematic ,
I review of Station Procedures required by ACP-QA-3.02. ACP-QA-3.02, section 6.2.3, includes 2400 series SP or IC procedures. In a quarterly memorandum (MP-91-918), dated November 1, 1991, Document Services identified the last biennial review date for SP 2404AS as December 1, i
1989, and a due date of December 1,1991.
Conclusions l
i The biennial review of SP 2404AS was not completed in a timely manner as required by ACP-QA-3.02D. NNECO agreed to evaluate this matter, take appropriate action as necessary, and respond to the NRC. Additional examples of overdue biennial reviews are described in !
Inspection Report 50-336/91-29 and elsewhere in this report. {
Based on review of SP 2404AR and SP 2404AS, plant walkdown, and discussion with cognizant I&C Department personnel, the inspector concluded SP 2404AR and SP 2404AS were adequate.
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5.2 Conduct of Surveillance Testing De inspector reviewed the conduct of surveillance testing which is required to ensure operability of the HRSGM. .
l Inanectian Findinet MP2 TS 3.3.3.1 requires a minimum of one operable HRSGM chanr.el in modes 1, 2, 3, and
- 4. To demonstrate operability of the HRSGM when in modes 1,2,3, or 4, MP2 TS 4.3.3.1 requires a cnannel check once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a channel calibration once per 18 months, and a channel functional test once per 31 days. De channel check is done as part of SP 2619A. The channel calibration and the channel functional test are done using SP 2404AS and SP 2404AR, l respectively.
Surveillance testing of the HRSGM was done at.yymydate intervals. The surveillance schedule is based on ACP-QA 9.02 and the MP2 TS. ACP-QA-9.02 defines the station surveillance program and ACP-QA-9.02B is the master control list for MP2 surveillance tests. I&C !
Department records for SP 2404AR and SP 2404AS inoicates suivaillane* tests were done at I appropriate intervals during 1990 and 1991.
Conclusions NNECO conducted surveillances SP 2404AR and SP 2404AS at appropriate intervals during 1990 and 1991.
6.0 INSTRUMENT ROOT VALVE LABELS The NRC provided a concern related to the 1=haline of instrument root valves in the instrument o
air system. NP.C disposition of that concern involved providing the concern to NNECO for review and resolution, with subsequent NRC evaluation to ensure the adequacy of NNECO's actions. NNECO letter A09960, dated December 6,1991, described NNECO's review of that Concern.
6.1 Instrument Air Root Valve labels According to Northeast Utilities memorandum MP-2-91-159, dated September 30,1991, valve labeling was 89% complete with a projected completion date of December 1992. Instrument and gauge labeling was complete on 17 of 128 systems, and expectd completion was expected by December 1995. NNECO was tracking the plant labeling progran? status as controlled routing l (CR) item CR 8139.
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Assessment The inspector discussed instrument air valve labelmg with cognizant I&C and Operations personnel and reviewed applicable procedures and documents. NNECO stated its approach was to label instrument air (IA) valves that had assigned identification (ID) numbers. IA supply stop valves to I&C components, such as air wd valve (AOV) positioners, usually did not have ID numbers. Also, IA supply stop valves were normally located adjacent to their as;ociated component, such as an AOV, that had a known ID number. 'Ihus, NNECO stated there was no need to label such IA supply stop valves.
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According to NNECO, the Electromark label data base had approximately 733 IA valve labels.
These were for IA valves that were ctds.sdily operated by Operations personnel. ;
Approximately 30% of that total number of valves actually had labels in place at the end of 1991.
NNECO stated that although IA supply stop valves for I&C components generally did not need labels, labels were provided when warranted in specific cases. For example, the 3 eater drains tank normal level control valve (2-HD-109) was an AOV with two IA supplies to its positioner. [
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'Ihe second IA supply stop valve (2-IA-632) to 2-HD-109 was appropriately labeled as
" redundant air supply to 2-HD-109." Additionally, ACP 6.22 allowed plant personnel to request ,
desired labels.
Physical inspection of AOVs 2-CND-34 and 2-CND-37 showed the following: (1) the AOVs j had adequate labels and (2) IA supply stop valves for these AOVs were clearly associated with a specific positioner because of their close proximity (less than two feet) and an unobstructed view of IA tubing to the positioner. [
Conclusions Based on physical inspection, review of applicable procedures and documents, and discussion with cognizant NNECO personnel, the inspector concluded that, if effectively implemented, j NNECO's labeling program would result in adequate labeling for MP2 systems and components. l Also, the inspector concluded that NNECO had a adequate methodology for labeling IA supply !
stop valves to I&C components. Finally, IA supply stop valves for AOVs 2-CND-34 and 2- l CND-37 did not require labels because they were obviously and uniquely associated with a j specific positioner. ;
6.2 Inst:urnentation Valve IJne-up ,
i To sample the current MP2 I&C Department meadalogy for verifying instrument valve ,
positions, the inspector observed an instrument valve line-up.
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Assessment P
'Ihe inspector did a general review of Instrumentation and Control ma intenance proced ure IC '
f 2436A. 'Ihe inspector also observed a pomon of the conduct of IC 2436A that was done January 8,1992, per AWO M2-91-10712.
ACP-QA-2.12, section 6.4 2.1 required that the MP2 I&C D,usisat ensure proper alignment of certain MP2 instrumentation isolation stop valves. 'Ihe MP2 I&C Department used IC 2436A r and data sheet I&C Form 2436A-1 to document the proper valve line-up of specified safety- [
relat4 instrumentation. *Ihe laWar observed the verification ofinstrument isolation valves, {
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vent hlves, drain valves and equalizer valves for the following level transmitters (LTs) and flow l transmitters (FTs):
i Instrument No. Service LT 3001 Reactor Water Storage Tark (RWST) level :
LT 3002 RWST level RWST level l LT 3003 RWST level l LT 3004 Auxiliary Feed (AFW) Flow - S/G #1 .;
FT 5277A FT 5277B AFW Flow - S/G #1 [
FT 5278A AFW Flow - S/G #2 i
FT 5278B AFW Flow - S/G #2 NNECO was in- the process of implementing a program to improve component labeling ;
throughout Millstone Station, but the above transmitters were not yet adequately labeled. ACP- l 6.22, section 6.2.3.2, required L* ling of instruments and gauges used for either reading a :
measurement or operating the plant. Component identification was done using the technician's i i
knowledge of equipment location, preeitting calibration stickers and pened marked identification numbers, and by tracing instrument sensing lines back to properly labeled root ;
valves.
j Step 6.1 of IC 2436A required verification of valve position per I&C Form 2436A-1, which included " Isolation" and "Equalimr/ Vents." When doing IC 2436A, the technicians actually l verified instrument isolation, equalizatina, and vent and drain valves. The inspector discussed j with I&C supervision the need to clearly state that the position of instrument drain valves was !
verified on I&C Form 2436A-1. NNECO agreed that clarification of I&C Form 2436A-1 l regarding instrument drain valve verification was a warranted enhancement.
IC 2436A, section 6, required L-$-:r '=t verification of valve position per ACP-QA-2,12. l ACP-QA-2.12 referenced ACP-QA-2.20. ACP-QA-2.20, section 6.1.2, required (in part) that !
" Verifier ia&;+-*w must be maintained to ensure the integrity of the indapandant verification by minimizing interactions between individuals." j i
l
., , - .. y - - -
i 13 !
De I&C technicians did IC 2436A in close proximity and maintained oral communication during l verification activities. De lead technician tacated the transmitters and verified valve positions, r while the second verifier (a contract technician) was in the immediate vicinity. De second !
. verifier then verified the valve positions while the lead technician was in the immediate vicinity.
Both technicians individually did their valve position verification in a diligent manner that the
, inspector believed to be consistent with the techniques commonly employed by experienced and ,
competent technicians. . l When questioned by the inspector if "i_Pt verification," as defined in applicable ACPs, !
allowed both verifiers to have significant interaction during IC 2436A, the technicians and I&C l supervision were uncertain ofi%t verification requirements. NNECO agreed to evaluate l the MP2 I&C Department i~ianawat verification methodology, take appropnate action as j a~**=y, and respond back to the NRC. ;
i On January 13, 1992, NNECO verified the correct position of the preceding valves, as !
documented in AWO M2-92 00347. Also, NNECO stated that it initiated a review of Millstone {
Station practices and procedures for independent verification. 'l
[
Conclusions !
During IC 2436A on January 8,1992, there were two verifications of valve position, one by - i cach of two I&C technicians, but the verification was done in a collaborative rather than '!
irht manner. Because NNECO's valve laheling upgrade effett was not yet complete, j the inspector believed there was increased imp.,Fance in doing thorough and stringent !
indapandant verification activities. The Noti.:e of Violation etained in NRC Inspection Report 50-336/91-29 described an example ofinar' equate i%t verification that occurred during surveillance 2404AI-1 on December 4,1991; since the corrective r:tions for that violation have j not yet been completed, the inspector considers this iredent another example of that violation. ;
Based on review of applicable requirements, direct observation of iPt verification j activities, and discussion with cognizant NNECO personnel, the inspector concluded that NNECO management ex=a**6ons for iadapaadant verification activities were not clearly defined and communicated to plant personnel. NNECO agreed to evaluate this matter, take appropriate l action as - y, and respond back to the NRC. I 6.3 Qualification of Personnel to do Valve IJne-up Work ;
t The inspector reviewed the process for identifying the qualification status of personnel doing valve line-up work. l Assessment ;
i ACP-QA-2.12, section 5.3.2, required that I&C supervision " Ensure instrument stop valve checkoffs are performed by qualified personnel at the prescribed frequency." De Nuclear l
l l
i
14 Training Department (NTD) stated that the Combined Administration Course included discussion of independent verincation requirements. Both technicians who did IC 2436A on January 8, 1992, stated they attended the Combined Administration Course. The inspector questioned I&C supervision regarding applicable requirements for qualification to do independent veri 6 cation of safety related instrumentation valve line-ups. Also, the inspector questioned if I&C technician contractors had the requisite qualifications to do IC 2436A. I&C supervision was uncertain of the qualification requirements. NNECO agreed to evaluate this matter, take appropriate action as necessary, and respond back to the NRC.
l 1
Conclusions Based on review of applicable requirements and discussions with NNECO personnel to date, the inspector concluded that, in this instance, NNECO did not adequately ensure personnel assigned to do independent verincation work had the appropriate formal qualincation, as required by
, ACP-QA-2.12. However, no instances were noted in which valve line-up work or independent l verification activities were incorrectly performed. The inspector considers this incident another example of the procedure compliance violation cited in IR 50-336/91-29.
6.4 Housekeeping During inspection of work activities related to AWO M2-91-10712, the inspector observed the following housekeeping issues:
Assessment Unanchored material was located near FTs 5277A and B. Those FTs were in a radioactive materials storage area of the Auxiliary Building west penetration room at the 38'-6" elevation.
Contrary to clearly delineated floor markings, a carton approximately 1/2 cu ft in size and an unknown plastic wrapped metal object approximately 3' by 3' by 6" were stored next to the instrument rack for FTs 5277A and B. NNECO immediately moved the carton to a proper storage location. Additional examples of unanchored material in MP2 included the following:
(1) a welding machine was stored next to containment penetration for a sample line, (2) a six foot ladder was placed against FTs 5278A and B, and (3) tool boxes were stored on a wheeled I cart that was adjacent to FTs 5277A and B. ]
l On January 8,1991, the inspector questioned if storage of unanchored materM Wjacent to safety related components was consistent with the seismic considerations described in ACP-QA-4.01, section 6.4.7, regarding the potential for the unanchored material to detrimen:a!!y affect safety related equipment. NNECO agreed to evaluate this matter, take appropriate action as necessary, and respond to the NRC.
During the next four weeks, the inspector did follow-up inspections to determine if NNECO I
adequately resolved the above described housekeeping issues. The inspector found that NNECO I
15 did respond to specific NRC findings, but effective corrective action was not always taken to maintain conformance with ACP-QA-4.01. .
1 I
Conclusions Based'on review of ACP-QA-4.01 and physical inW of the Auxiliary Building west penetration room at the 38'-6" elevation, the inspector concluded NNECO did not adequately store material in all cases. The inspector found no instance in which there was actual damage to safety related components, but inadequate storage of unanchored material had the potential 1 during a seismic event to detrimentally affect safety related components. Further, NNECO corrective action to date was inadequate to ensure conformance with ACP-QA 4.01 requirements. 'Ihis issue remains an unresolved item pending future review of the adequacy of - ,
corrective actions (50-336/91-3101). {
l EOF DIESEL GENERATOR )
7.0
'Ihe NRC provided a concern related to the Emergency Operations Facility (EOF) Emergency Diesel Generator (EOF-EDG). 'Ihe concern related to the adequacy of the work order (AWO M2-89-09594) for doing the annualload run using maintenance procedure MP 2722B on August 31,1990, and certain EOF-EDG drawings. NRC disposition of that concern involved providing the concern to NNECO for review and resolution, with subsequent NRC evaluation to ensure the adequacy of NNECO's actions. NNECO letter A09920, dated November 27,'1991, j described NNECO's review of that portion of the concern related to AWO M2-89-09594. The balance of the concern regarded assertions of discrepancies in certain EOF-EDG related drawings. NNECO stated that " evaluation and submission of any necessary drawing changes should be completed by late December,1991."
Background
Adequate provisions for emergency facilities and equipment, included "at least one on-site and one off-site communications system; each system shall have a back up power source," according to 10 CFR 50, Appendix E, Paragraph IV.E.9. EOF requirements were defined in NUREG-0737, Supplement 1, section 8.4, and other applicable regulations and licensing commitments.
In section 12.3, the MP2 FSAR stated that the Millstone Nuclear Power Station Emergency Plan (EPlan) addressed the criteria set fonh in NUREG-0654, revision 1, and NUREG-0737, supplement 1. EPlan, section 7.2.5, stated that emergency power was "provided by an auto-start diesel generator that is capable of meeting all EOF power requirements." *Ihe EOF-EDG was designed to provide a backup electrical power supply for the EOF.
The NRC documented its previous inspection of the EOF-EDG in Inspection Report 50-245/91-l 19,50-336/91-23, and 50-423/91-19 (IR 91-23), section 9.0. IR 91-23 concluded, in part, that the EOF-EDG satisfactorily operated in 1991 during hurricane Bob and during the September 1991 EPlan exercise.
L I= _ --_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- .. - - - - - - .._ ~ _- -_ .
=
. 16 On November 1,1991, as documented in Plant Incide st Report (PIR) 2-91-117, the EOF-EDG - ,
failed to load automatically during surveillance EPIP-4606. NNECO cleaned the timer card contacts and ==< fully tested the EOF-EDG. As documented in PIR 2-91-123, on i November 8,1991, the EOF-EDG again failed to load automatically during surveillance EPIP- !
4606. NNECCO prepared AWO M2-91-12093 for repair of the loading problem. Repair work ;
included replacement of the transfer-retransfer module (ONAN part no. 300-1188). l Subsequently, on December 5,1991, NNECO ewee< fully tested the EOF-EDG with assistance ;
from the vendor, GLT Ir.dustries.
1 7.1 Status of Piscedures for EOF-EDG Work ne inspector reviewed the status of various procedures used to do work on the EOF-EDG. j Assessment i
The inspector did a general review of Maintenance Procedure MP 2722B and Emergency Plan '
Implementing Procedures EPIP 4303 and EPIP 4606. MP 2722B was for performance of the annual EOF-EDG load run and EPIP 4303 was for EOF-EDG automatic and manual operation, i
. and EPIP 4606 was a periodic test of EOF-EDG operational readmess. j ne biennial review of MP 2722B was overdue. ACP-QA-3.02D, section 6.1.1, required a -
periodic, systematic review of Station Procedures required by ACP-QA-3.02. ACP-QA-3.02, section 6.2.3, included 2700 series MP procedures and EPIP 4000 and 4600 procedures. In a quarterly memorandum (MP-91-918), dated November 1,1991, Document Services identified l the last biennial review date for MP 2722B as December 6,1989. Completion of the MP 2722B biennial review was due by December 1,1991. De biennial review of EPIP 4606 was done September 15,1991. Document Services, as of January 16,1992, had no record (either revised ;
procedures or Form 364's to document review) that NNECO did the biennial reviews of EPIPs ,
4303,4306,4605,4606,4608, and 4609. NNECO stated that biennial reviews were done but i not yet approved by SORC.
At the end of this in=~*iaa_ period, NNECO converted EPIPs 4303,4306,4605, 4606, 4608, I and 4609 to either operating procedures (ops) or surveillance procedures (SPs). The new l procedures were OP 2399A, OP 2399B, SP 2678A, SP 2678B, SP 2678C, and SP 2678D, respectively. NNECO obtained PORC approval to issue the new ops and SPs, and SORC l approval to cancel the Wing EPIPs.
MP2 FSAR Appendix 12A included Appendix D, listing of Supporting Procedures that Implement the Plan, that listed EPIP 4302. EPIP 4302 was canceled and incorporated with EPIP 4304.
i 17 Conclusions Based on review of applicable procedures and discussion with cognizant NNECO personnel, the ,
inspector concluded the biennial review of MP 2722B and EOF operations related EPIPs was i
- not done as required by ACP-QA-3.02D. NNECO was persuing 'mcorporation of biennial !
reviews for those EPIPs concurrent with the issuance of the new ops and SPs. Additional examples of overdue biennial reviews were described in Inspection report 50-336/91-29 and elsewhere in this report; the inspector considers this findmg to be another example of the ,
procedure compliance violation noted in that report.
Also, the inspector concluded the MP2 FSAR did not contain an accumte listing of EPIPs. .
NNECO agreed to evaluate this matter, take appropriate action as necessary, and respond to the NRC.
7.2 EOF-EDG 1990 AnnualIAnd Test Assessment ,
J NNECO did an annual load test of the EOF-EDG on August 31,1990, using AWO M2 ;
i 09594. ACP-QA-2.02C, section 6.3, required the lead department head or authonzed person to " review all procedures and forms referenced on the Work Order to insure they provide adequate guidance for the work to be performed." Because it was a non-QA work order, there l l
was no requirement for AWO M2-89 09594 to include all work related procedures (e.g., MP 2722B) and forms in the work package. AWO M2-8949594 referenced Maintenance Form 270lJ-43, EOF, and Emergency Security Diesel. Maintenance Form 270lJ-43 was a list of inspection items and a@ce values for operating cycle preventive maintenance.
The annual load test involved opening the EOF-EDG output breaker and installing a load bank.
The safety evaluation for MP 2722B stated (in part) that, "In the event that manning of the EOF is required simultaneously with a loss of normal power while the load run is in progress, the ,
electricians monitoring the run could reconnect the diesel to the EOF within a half an hour."
This work was done by the vendor representative and a NNECO mhanic. The inspector questioned if the safety evaluation was valid because electricians were not present during the load run. NNECO agreed to evaluate this matter, take appropriate action as n~+=y, and respond to the NRC.
Conclusions Based on review of applicable procedures and documentation, the inspector concluded AWO M2-89-09594 contained adequate references. Although it may be convenient for workers to have a reference to MP 2722B in work orders for doing the annual EOF-EDG load test, there was no requirement to do 30.
l 18 7.3 AFBuut Drawings The inspector reviewed drawings 25205-30007,25205-39002 (sheet 3), and 25205-32008 to determine if those drawings adequately depicted the as-built status of certain components. The ,
components were circuit breaker (CB) 26 in electrical lighting panel ELP1, CBs 10 and 12 in ,
ELP2, remote control panel PN1, and a utility plug near the automatic bus transfer (ABT) device. .
Assessment l
Drawing 25205-30007 was marked "non-QA" and stated it was as built per PDCE-2-074-89 and '
DCR-MG-P-105-90. ELP1 slot 26 was labeled " SPACE" and the drawing indicated ELP1 had no CB installed in slot 26. The inspector found a CB !abeled " COMPUTER ROOM HIGH ;
i SPEED PRINTER" installed in ELP1 slot 26. ELP2 slots 10 and 12 were labeled " SPARE" and " TELEPHONE UPS," respectively, and the drawing indicated EPL2 had no CB installed in slot 10. "Ihe inspector found CBs labeled " LIGHTING CONTROL CONTRACTOR" and
" BATTERY CHARGER" installed in ELP2 slots 10 and 12, respectively. NNECO stated that !
it recently did a walkdown of electrical panels ELP1, ELP2, EPP1, and EPP2 to ensure that the -)
actual panel configuration =*had the circuit descriptions in drawing 25205-30007. Based on !
that review, NNECO initiated design change notice (DCN) DM2-P-075-91, dated December 13, 1991, to correct the as-built status of ELP1 slot 26 and ELP2 slot 10 as shown on drawing .
25205-30007. NNECO stated that it intended to revise the label on ELP2 slot 12 to conform !
with drawing nomenclature.
At location G-8, drawing 25205-30007 indicated there' was a " REMOTE CONTROL PANEL l' IN MECH. EQUIP. ROOM." This was not a physically separate panel and there was no requirement to that effect. The remote control panel was located within the ABT cabinet in the mechanical equipment room.
Drawings 25205-32008 and 25205-39002 (sheet 3) indicated they were as-built per DSR-M2-S-
' 371-82. Drawing 25205-39002 (sheet 3) indicated the ABT device schematic was redrawn in drawing 25205-32008. The 4;mtor compared drawing 25205-3200fl with 25205-39002 (sheet 3). Because drawing 25205-32008 depicted some elements in a manner somewhat different than drawing 25205-39002 (sheet 3), the inspector questioned if drawing 25205-32008 matched as-built conditions. NNECO did a walkdown of the ABT panel and confitned the adequacy of technical drawing elements. Drawing 25205-39002 (sheet 3) had references to notes 2, 3, 4, and 5, but drawing 25205-32008 did not reference those notes. The inspector questioned if notes 2, 3,4, and 5 were adequately depicted on the drawings. NNECO stated that it initiated DCN DM2-P-001-92, dated January 9,1992, to clarify these notes and other aspects of drawing 25205-39002 (sheet 3).
The inspector reviewed other assertions of technical concerns regarding drawing 2520542008.
Drawing 25205-32008 correctly depicted a utility plug (PI) that was physically located at the bottom of the ABT cabinet. NNECO stated that switch A9 had incorrect nomenclature ("STOP-l
- l. 19 HC-NORMAL"). The correct switch positions were "STOP-AUTO-RUN." Also, NNECO i stated that it would make appropriate changes, as necessary, in procedures MP-2722B and EPIP-4303 to reflect the correct switch nomenclature.
I Conclusions l Based on physical inspection of equipment, review of applicable drawings, and discussion with cognizant NNECO personnel, the inspector concluded NNECO did not adequately maintain the !
as-built status of drawings 25205-30007, 25205-39002 (sheet 3), and 25205-32008. The '
inspector believed nomenclature differences and drawings that did not match as-built conditions had the potential to cause worker confusion, but the inspector found no evidence that such drawing inadequacies resulted in a significant degradation of EOF-EDG operability or reliability.
Finally, the inspector concluded that, if effectively implemented, DCNs DM2-P 001-92 and DM2-P-075-91 would adequately resolve relevant concerns regarding EOF-EDG drawings. j Inspection Report 50-245/91-23 and 50-336/91-27 (IR91-27), section 7.0, described a number !
I of similar concerns regarding incorporation of design changes into drawings and the accuracy of as-built conditions shown in drawings.
7.4 Qualification of Personnel to Perfonn EOF-EDG Work r
'Ihe inspector reviewed the process for identifying the qualification status of personnel doing l work on the EOF-EDG.
Assessment t The inspector did a general review of the training NNECO provided for Operations and !
Maintenance personnel involved in EOF-EDG work. Operations personnel received on the job :
i training in EOF-EDG operation as part of their training that was documented on Plant Equipment Operator (PEO) Qualification Sheet E5-25. Mechanical and electrical maintenance personnel do not receive specific training on the EOF-EDG. According to the Nuclear Training l Department (NTD), the EOF-EDG is not on the task list for mechanic training. On January 13, ,
1992, NNECO held an MP2 Mechanical Training Program Control Committee (TPCC) meeting l and concluded that EOF-EDG training was required and the scope of training would be determined by task analysis. NNECO stated that the NTD previously did a task analysis of )
electrician work on the EOF-EDG and concluded there was no need for specific EOF-EDG electrician training. I Conclusions Based on review of relevant documentation and discussion with cognizant personnel, the f inspector concluded there were opportunities for improvement in the training and qualification for personnel doing work on the EOF-EDG. 'Ihese were not regulatory requirements, but were training program enhancements that could help ensure the adequacy of EOF-EDG maintenance.
)
)
i l
20
- 8.0 SEISMIC QUALIFICATION OF HPSI SUCTION PRESSURE !
GAUGES .
The NRC provided a concern related to the seismic qualification of suction pressure gauges in !
the High Pressure Safety injection (HPSI) system pump suction piping. The root piping and ,
associated tubing and components are classified as QA, Seismic Category I. !
Background ;
For in service inspection (ISI) purposes, PDCR 2-112-79 installed commercial grade pressure indicating gauges in the suction piping of several safety related pumps. Northeast Utilities !'
memorandum NSE-M-86-59 identified a concern that the installation of those gauges was not o
seismically reviewed and requested an evaluation. ,
t Northeast Utilities memorandum PSE-SA-89 061 described an evaluation of 17 pressure gauges, related to PDCR 2-112-79, that included HPSI system PIs-3046, 3048, and 3050.- That memorandum stated that "the pressure gauges were installed using several na-ry fittings, '
couplings, and valves. No apparent design criteria were used. The present configuration is not ;
in accordance with established plant design criteria. - However, by engineering judgement, a !
postulated DBE seismic event would not result in a structural failure that would compromise the ' :
integrity of the associated piping system." PSE-SA-89-061 recommended that "the pressure .{
gauges be modified aswiding to Figures 14 and 15. The root piping and pressure gauge fittings l
have been evaluated in the modified condition for all applicable load cases. All calculated ;
stresses are within the code allowable limits as defined in ASME III,1974 Edition and are !
documented in reference 2 (NUSCO Calculation PDCR-2-ll2-79-1067 GP, revision 0). ;
r NUSCO reportability evaluation, REF 91-34, dated August 22,1991, determined that "there -
were no pressure boundary concerns for the subject inemlistions. All stresses in the piping have been determined pele and meet the design basis Code allowable limits." "Ihis evaluation was based (in part) on the y..;tice ofisolating the pressure gauges and associated tubing during normal operation. Also, NU memorandum MCE-SA-91-105 stated that NUSCO calculation MP2-LOE-079EM determined there were "no pressure boundary concerns for the subject ;
installations."
NRC Inspection Report 50-336/91-15 (IR 91-15), section 6.6, described a similar concern regarding the seismic qualification of pressure gauges on the service water supply strainers to the emergency diesel generators (EDGs). According to IR 91-15, "NNECO review of existing plant conditions during routine operating activities identified a potential nonconforming condition '
on a safety related EDG. NNECO reviews identified that the 1977 PDCR documentation was discrepant in that the seismicity of the instrument gages on the seismically installed strainers were not specifically addressed." NNECO took adequate action regarding the potential ,
nonconforming condition and there was no impact on EDG operability.
a {
- - - , - - , , - - - , , , - r-,,--. m -- m- - e ., ,--4 ,--,-e .- ,- - ,a -- - -
21 8.1 PI Root Valve Position The inspector reviewed relevant documentation and physically inspected the following PIs.
Drawing status refers to the instrument root valve position depicted on the drawing and as found status refers to the actual position observed by the inspector.
Pressure Gauges In Pump Suction Lines
- Root Valve Position Gauge ID Root Associated Pump Drawing 25203- As Valve 2- Found P&ID No. Status Status PI-3046 SI-090 P-41 A HPSI 26015/sh 2 Closed Closed PI-3048 SI488 P-41B HPSI 26015/sh2 Closed Closed PI-3050 SI-086 P-41C HPSI 26015/sh 2 Closed Closed PI-5403 CN-%B P-9A Aux. Feed 26005/sh 3 Closed Closed PI-5405 CN-97B P-9B Aux. Feed 26005/sh 3 Closed Closed PI-5401 CN-95B P-4 Aux. Feed 26005/sh 3 Closed Closed PI-6743 RB-111 A P-11A RBCCW 26022/sh 1 Closed Closed .
PI-6745 RC-111C P-11B RBCCW 26022/sh 1 Closed Closed PI-6747 RB-111E P-11C RBCCW 26022/sh 1 Closed Closed PI-3051 SI-093 P-42A LPSI 26015/sh 1 Open Open PI-3053 SI-091 P-42B LPSI 26015/sh 1 Open Open l
PI-3055 CS-032 P-43A Cont.Sp. 26015/sh 1 Closed Closed PI-3057 CS-030 P-43B Cont.Sp. 26015/sh 1 Closed Closed PI-7436 RW-126A P-13A SFP Cool 26023/sh 2 Closed Closed PI-7662 RW-126B P-13B SFP Cool 26023/sh 2 Closed Open PI-8859 CHW-7 P-122A CHW 26027/sh 2 Closed Open PI-8863 CHW-36 P-122B CHW 26027/sh 2 Open Open
a i
22 f i
Assessment ;
ne inspector found the following PI root valves to ayyivyriately be in the closed position: SI- i 090, SI-088, SI 086, CN-96B, CN-97B, CN-95B, RB-111 A, RB-111C, RB-111E, CS-032, and !
CS-030. De inspector had no further questions regarding the position of those valves.
i Low pressure safety injecdon (LPSI) valves 2-SI-093 and 2-SI-091 were open because they were t also the root stop valves for FT-3051 and PT-3053, respectively. This was consistent with the ;
applicable drawing, prior engineering evaluation, and valve line up sheets (OPS Forms 2604L-2 !
and 2604M-2). De function of FTs 3051 and 3053 was to provide a low pressure alarm when in RCS reduced inventory operations. When questioned by the inspector, NNECO stated that it did not do a r pwr.liility evaluation, per NEO 2.25, for the modification that installed a " tee" l in the instrument tubing thht <=W FTs and PIs downstream of 2-SI-093 and 2-SI-091. l NEO 2.25 promulgated NNECO's instructions for 10 CFR 50.72,10 CFR 50.73, and 10 CFR I 50.9 operability and reportability determinations. NNECO stated that it would do an NEO 2.25 l evaluation of the above described 1990 plant modification that was done per PDCR 2-016-90. l 1
Testing of the spent fuel pool cooling (SFP) pumps was in progress during the inspection, nerefore, the inspector was not certain if valves 2-RW-126A and 2-RW-126B were in their normal position or a test position. According to the valve line up sheet (OPS Form 2305-1),
valves 2-RW-126A and 2-RW-126B were normally open.
l Regarding the chilled water system (CHW), the inspector found valves 2-CHW-7 and 2-CHW-36 I to be in the open position. According to the valve line up sheet (OPS Form 2330C-1), valve l 2-CHW-7 was normally closed and valve 2-CHW-36 was normally open. OPS Form 2330C-1 '
l and the valve label showed normally closed valve 2-CHW-37 to be ." Chill Water Pump (P122B)
PI-8863 Isolation." Rus, the inspector found that valve 2-CHW-7 was not in its expected position and that there were inaccuracies in OPS Form 2330C-1 regarding valves 2-CHW-36 and 2-CHW-37. Also, the inspector found that drawing 25203-26027, sheet 2, showed PP-8858 installed downstream of normally open valve 2-CHW-6, but 2-CHW-6 was normally closed and PP-8858 was not installed. De inspector discussed the CHW valve line up with the on duty SS and SCO who prompdy corrected the position of valve 2-CHW-36 and agreed to have cognizant NNECO personnel review OPS Form 2330C-1.
Because there were inaccuracies in OPS Form 2330C-1, the inspector questioned if the valve line-up sheets contained accurate information for the other valves listed in the preceding table.
NNECO promptly reviewed the affected OPS Forms and found that the OPS Forms agreed with the applicable drawing, except for the SFP system. Subsequently, NNECO initiated a change to OPS Form 2305-1 to indicate a normally closed position for 2-RW-126A and 2-RW-126B.
ne inspector questioned if a normally open position for valves 2-SI-093,2-SI-091,2-RW-126A, 2-RW-126B, and 2-CHW-36 was consistent with design assumptions regarding the seismic qualification of the Pls associated with these instrument root valves. NNECO stated that its NEO 2.25 evaluation appropriately assumed an open position for these valves. The inspector
j
.~ _ _ _ _ .
s
. I 23 ;
questioned why 2-CHW-36 was not a normally closed valve. NNECO agreed to evaluate if 2- ]
CHW-36 and 2-CHW-7 should both be in the normally closed position, and take appropriate _j action as necessary.
L
'Ihe inspector reviewed a representative sample of surveillance procedures (SPs) for pump ;
operability determination. 'Ihose SPs typically involved use of the Pls listed in the precedmg :
table for their associated pump. *the SPs did not specifically indicate if the instrument root l' valve for each PI was to be opened during pump operation and then closed during restoration to its normal position. The inspector discussed with cognizant Operations personnel the need !
. to clanfy instrument root valve position in such SPs. NNECO agreed that specifying instrument root valves in such instances may be desirable to ensure operators correctly position the valves.
Conclusions mwt on discussion with cognizant NNECO personnel, physical inspection, and review of .
[
relevant documentation, the mspector concluded that NNECO adequately maintained, in the- {
proper position, the instrument root valves delineated in the preceding table. 'Ihere was one i instance (2-CHW-7) in which a valve was found in an incorrect position, but the inspector believed that may have been due to an ongoing ISI test of the associated pump.
mwi on the inaccuracies noted for OPS Form 2330C-1 and OPS Form 2305-1, the inspector r concluded that personnel attention to detail in validation of these valve line up sheets may have been inadequate. NNECO agreed to evaluate the adequacy of OPS Forms 2330C-1 and OPS Form 2305-1, take appropriate action as m*=ry, and respond to the NRC. ;
mwt on review of some representative procedures and discussion with cognizant Operations personnel, the inspector concluded that an opportunity for improvement was listing in SPs the l instrument root valves that had to be opened or closed. 'Ihis was not a regulatory requirement, ;
but was an enh==awat NNECO agreed could be helpful. NNECO stated that it would evaluate the need to specify in SPs the instrument root valves for PIs used in pump operability ;
determinations, take swivysiate action as a-===ry, and respond to the NRC. l l
8.2 PDCR 2-89-046 Implanantation Status NNECO used a separate work order for each PI configuration that was to be modified due to l PDCR 2-89-046. 'Ihe inspector reviewed the implementation status of work orders associated !
with PDCR 2-89-046. ;
Assessment Work associated with HPSI pump PIs 3046, 3048, and 3050 was being done per AWOs M2 ;
08578, M2-91-08579, and M2-91-08580, ictively. Modification of the instrument tubing 4- for PI-3046 and PI-3050 was done as outlined by PSE-SA-89 061. NNECO issued DCN DM2-S-511-91 for the unique configuration required for PI-3048. Remaining work for all three r
I 24 AWOs involved procurement and installation of replacement gauges with a higher range.
Work associated with AFW pump Pls 5403,5405, and 5401 was done per AWOs M2-91-08569, ;
M2-91-08571, and M2-9108572, respectively. Work associated with RBCCW pump PIs 6743, ;
6745, and 6747 was not done; however, PIs 6744, 6746, and 6748 were donc per AWOs M2- i 91-08577, M2-91-08574, and M2-91-08573, respectively. The inspector questioned why the i' modified Pls were not the PIs listed in the PDCR. NNECO stated that PIs 6744,6746, and 6748 were the PIs actually used to do ISI testing of the RBCCW pumps; therefore, NNECO modified these Pls rather than PIs 6743, 6745, and 6747. NNECO stated it would prepare a j DCN that would define this change for PDCR 2-89-046.
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Because the gauges and tubing associated with PIs 6743,6745, and 6747 did not appear to be a standard design, as outlined by PSE-SA-89-061, the inspector questioned if NNECO evaluated the adequacy of these gauges. A similar example was instrumentation installed in the CHW system at valve 2-CHW-37. NNECO stated it did not evaluate these gauges, but would do so. !
Work associated with LPSI pump PIs 3051 and 3053 was done per AWOs M2-91-08581 and !
M2-91-08582, respectively. Work associated with Containment Spray pump PIs 3055, and 3057 l
was being done per AWOs M2-9108583 and M2-91-08584, respectively. Modification of the ;
instrument tubing for PI-3046 and PI-3050 was done as outlined by PSE-SA-89-061. Remaining ;
work for both AWOs involved procurement and ineallation of replacement gauges with a higher ;
range. Work associated with SFP pump PIs 7436, and 7662 was done per AWOs M2-91-08585, ;
and M2-91-08586, respectively. Also, work associated with CHW pump PIs 8859 and 8863 was :
donc per AWOs M2-91-08587 and M2-91-08588, respectively. j i
i The inspector questioned: (1) were there instances at MP2, other than those already described, that involved installation of non-seismically qualified instrumentation in a seismic category I ;
system, and (2) if so, when based on an appropriate evaluation, were the root valves maintained in a closed position (if required). NNECO stated that although no such evaluation was done on a programmatic basis, NNECO was not aware of any other simdar installations that deviated from the original root piping installation guidelines. Original guidelines included specifications , ,g 7604-MS-64 and 7604-MS-66. NNECO agreed to assess the need for programmatic evaluation ,
of non seismically qualified instrumentation installed in seismic category I systems, take appropnate action as m==y, and respond to the NRC. ;
I Conclusions -
Based on review of applicable documentation, physical inspection, and discussion with cognizant NNECO personnel, the inspector concluded that, if effectively implemented, completion of ,
outstanding DCNs and AWOs will result in satisfactory completion of the design change defined l in PDCR 2-89-046. Regarding the work associated with the RBCCW PIs, the inspector >
concluded this was an example in inadequate attention to detail in design control. The work i documents resulted in modification of PIs not listed in the design change document. NNECO agreed to issue a DCN to document the actual design implementation. r
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Conclusions.regarding the adequacy, on programmatic basis, of non-seismically qualified i instrumentation installed in scismic category I systems were pending NRC evaluation of I
NNECO's assessment of this matter.
9.0 ADMINISTRATIVE CONTROL OF RADIATION MONITOR MAINTENANCE >
[
. Concerns had been expressed regarding administration deficiencies surrounding activities for i radiation monitors. Specific issues include contradictions between procedures and vendor ;
manuals, setpoint control, and inadequate radiological work practices.
l Assessment One concem involved discrepancies between =- ;-m criteria specified by a vendor technical j manual and that specified by surveillance procedure SP 2404AG, " Waste Gas Process Radiation :
Monitor (RM 9095) Functional Test," revision 1. Specifically, the vendor manual had stated ;
that correct operation of the Upscale Check system was verified by obtaining a " count level
(
equal to the check source level." However, the procedure specified aw,cc criteria for the i Upscale Check as " count levelindicator increase." Because of this discrepancy, the validity of l the surveillance, and therefore the operability of the monitor was in doubt. l l
'Ihe licensee had r*W that the vendor manual contained generic recommendations for i Upscale Check tests, and that these recommendations were not applicable and were superseded l by the PORC-approved procedure. The Northeast Utilities Service Company (NUSCO) :
Radiological Analysis Branch had confirmed the adequacy of the Waste Gas Monitor functional l test in a memorandum (NE-91-RA-338, dated May 28,1991) and concurred that the procedure l took pr miarm over the vendor manual. -
The original concem and licensee response referenced Section 6.2 of a draft Revision 2 to the !
surveillance procedure as pmviding the intended acceptance criteria (" Upscale Check >
Background"). To date, this revision has yet to be approved. 'Ihe inspector considered that l NNECO should have referenced an approved procedure in response to this concern. !
The inspector also reviewed a copy of SP 2404AF, Revision 1, Change 5, which includes corrections to identified problems and extensive procedure step rewrites that were incorporated 3 as part of this latest upgrade. 'Ihe inpr determined, based on review, that the original l procedure " deficiencies" would not have prevented a knowledgeable I&C technician from )
completing the calibration in a satisfactory manner. The Change Routing Sheet used to process j and implement the I&C procedure change is not required by Millstone administrative procedures, !
but rather is a tool dcvc.+4 by the MP2 I&C department to initiate, track, and document !
actions taken by personnel in the procedure upgrade process. l l
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A concern was identified related to the disposition of setpoint control forms for radiation
- monitor surveillance. Specifically, promdure OP 2383C, " Radiation Monitor Setpoint Control,"
requires that Alarm Setpoint Control form OP 2383C-1 be forwarded to the Engineering j Department for review following an equipment setpoint change. However, during the ;
performance of SP 2404AV, "RBCCW Radiation Monitor RM 6038 Calibration," it was noted !
that Setpoint Control forms for two setpoint changes conducted May 11,1991, and July 8,1991, i had not been forwarded to the hp==ing Department. ,
NNECO responded that the swamury forms were on file, having been reviewed in September [
_ 1991ryne inspector obtained copies of the forms in question and conducted a review of the forms and procedure OP 2383C. De procedure specifies no time frame restrictions for routing Setpoint Control forms to the Engineering Dwhi,ent, and the inspector concluded that the ,
procedure was being followed. Unit 2 Engineering does maintain a file of all radiation monitor setpoint control forms. ,
i Conclusion f The inspector determined that NNECO took appropriate action in response to the above issues. j Correct actions were taken to resolve vendor manual and procedure differences for RM 9095. :
Additionally, the Change Routing Sheet was adequate as initially filled out to implement the [
r->=ry procedure change of SP 2404AF, and the I&C supervisor was exercising supervisory ;
discretion in his assignment of the procedure change action. The issues reflected minor ;
administrative problems in the conduct of routine maintenance, procedures, and record keeping. ;
These issues did not affect nuclear safety, and the corrective actions taken indicate that these l concerns should be closed. !
t 10.0 STATION BA'ITERIES l l
The NRC provided a concern about the safety-related Station Battery and the non-safety-related !
Turbine and Computer Battery procedures. The NRC disposition of this concern involved an ,
initial NRC inspection of the safety aspects of the concern and then the concern was provided to the licensee for review and resolution. After the licensee response was received, a subsequent l NRC inspection was conducted to evaluate the adequacy of the licensee's corrective actions. !
NNECO letter A10024, dated January 8,1992, described the licensee's evaluation of the concern. De results of the subsequent NRC inspection are as follows: l Assessment i The inspector reviewed the battery procedures in question along with the technical manual and ;
other applicable technical documentation associated with the installed batteries. The inspector t also interviewed the engineer responsible for battery procedures and reviewed the ongoing actions at Millstone to improve the battery procedures. ;
I A change to " Battery Pilot Cell Survei11=* " SP 2736A, did provide retorque values for the l
1 1
1
d 27 Station Batteries, 201 A and 201B. However, the vendor technical manual and the Institute of ,
Electrical and Electronics Engineers (IEEE) Standard 450-1980 requirement to perform periodic .;
connection retorque checks and the IEEE requirement to observe the battery for inter-cell !
connectson heating are not contained in present Station Battery or Turbine and Computer Battery procedures. Periodic terminal resistance checks are presently performed during Battery Service [
Tests, which are conducted every 15 to 18 months and use Individual Cell Voltage (ICV) (
measurements. NNECO is'm the process of revising the applicable battery procedures to include j the connection retorque check frequency and a periodic inter-cell electrical resistana .
measurement method, acceptance values, and test frequency. NNECO does not intend to !
institute electrical connection har temperature measurements during battery performance l discharge tests. NNECO technically justified this action and obtained the vendor's concurrence with this decision. j During the inspection, opportunities to improve the Battery Pilot Cell Surveillance procedure, )
SP 2736A, Computer and Turbine Battery Inspections procedure, MP 2720F1, and Battery Terminal Inspection and r'la-aing procedure, MP 2720F2, were noted. These are not j necessarily regulatory requirements, but constitute enhamments that would be helpful. The
- i following are examples of such improvement opportu'tities: l l.
- Incorriste the Caution statement of the vendor technical manual, VTM2-127-001A, i paragraph 4.3, that requires di-Mag the battery from the load and charger [
equipment when performing the connection checks; j
- Coordinate the battery procedure revisions so that the common notes, cautions, and !
actions are worded in standardized formats in all the appropnate procedures; and i o Since the Computer battery is not made by the same vendor as the Turbine battery, a j thorough review of both Technical Manuals should be made to insure that procedure !
guidance properly reflects the requirements of both batteries. If significant differences l are noted, it may be more appropriate to produce separate procedures for each battery and not retain the present common procedure. Since a Computer Battery technical I manual was not available on site, the inspector was unable to perform such a review. l Conclusion i The inspector concluded that the Millstone Unit 2 storage battery procedures were adequate for l routine operations, but that the applicable surveillance and maintenance procedures have not !
incorporated the periodic connection tightness checks contained in applicable technical documentation. NNECO is in the process of correcting these discrepancies. .
11.0 NNECO RESPONSIVENESS TO EMPLOYEE CONCERNS j The NRC received approximately 26 concerns regardmg the lack of responsiveness by NNECO to employee concerns, particularly from technicians. Specifically, it was serted that j l
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h
. 4 28 technicians provided feed back and suggested improvements, but did not receive timely responses from their managers. In response to this panicular category of concerns, the system that was j established by the Unit 2 I&C Manager to track such employee concerns was inspected to . ;
evaluate the validity of these assertions. NNECO's overall program for responding to and ;
resolving employee concerns will be addressed in a broader, more generic manner. !
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i L Assessment The records for 1990 and 1991 of the Unit 2 I&C Department Manager's employee concerns l tracking system, titled 'Worklist/ Memo," were reviewed in an attempt to determine the l effectiveness of the system and evaluate the responsiveness of the I&C Manager to employee l concerns.
The system is maintained in a computer data base with the I&C Manager's secretary entering !
the data. There were a total of 114 items documented in 1990 and 62 items in 1991. Thirty- j eight percent (38%) of the 1990 items and 24% of the 1991 items were logged as closed, which on the surface appeared to be quite low. However, when the lists were reviewed more thoroughly, many of the items that were listed as open were effectively resolved, but still carried as open items by the I&C Manager awaiting the completion of some administrative or follow-up action. 'Ihe system was used by the I&C Manager as a way to track actions and not as a feedback system to the individuals submitting the concerns. A monthly printout of the open and closed items is made and interested individuals in the depanment can check this printout to insure that their concerns have been acted upon. A feedback response to the individual submitting a concem might have eliminated some of the assertions, but would also increase the administrative burden. For such a small department, the monthly printout would appear to be adequate.
Conclusion The I&C Department has a system to track employee concerns (and has expended a large amount of effort to respond to them), but the individual must take some action to' determine the status of their concerns.
12.0 EDG CLEAN WASTE TANK PDCR MP-2-90 035 The NRC provided a concern that a modification to install float switches in the Emergency Diesel Generator (EDG) Clean Waste Tank at Unit 2, per PDCR MP-2-90435, failed to provide correct as-built drawings and that a blue colored wire was substituted for the yellow colored wire specified in the PDCR, due to non-availability of the yellow colored wire.
Assessment The drawings that were alleged to be inaccurate, 25203-31165 (Sheet 22), 25203-31175 (Sheet
< j
- 29 l
- 11), and 25203-32018 (Sheet 10), were obtained from Nuclear Records. The drawings were 1 called up on the Generation Records Information and Tracking System (GRTIS) and each drawing in GRITS reflected exactly the same revision as the drawings obtained from Nuclear Records, but all three drawings also indicated an open DCR, titled "M2, P0059-91 (PDCR)," ,
and listed the Engineering Supervisor as the contact person for the change. De inspector prmwed to the engineering office and obtained copies of the three drawings in question. Each drawing contained the modifications associated with the float switch installation. Dese revised i drawings correctly indicated the use of blue colored wire versus the originally specified yellow I colored wire. !
i Conclusion l
The updated drawings correctly indicating the modifications associated with the installation of j the EDG Clean Waste Tank float switch installation were properly identified in the GRITS and l would have been available to maintenance personnel who used the GRITS to verify drawing l accuracy prior to initiating work.
4 l
13.0 RECORDER CALIBRATION METHODOLOGY ;
The NRC provided a concern regarding the method for recorder calibration. The concern was j that, when calibrating recorders, NNECO did not use a calibrated voltmeter to measure the j output of a calibrated voltage source. A related concern was that supervisory review of some !
completed instrument calibration data sheets may have been inadequate because the listed test :
instruments did not include in all cases both the calibrated voltmeter and the calibrated voltage !
source. Specific examples cited were boric acid flow control recorder FR-120Y and process i radiation monitor multipoint recorder RJR-9373. NRC disposition of this concern involved 1 providing the concern to NNECO for review and resolution, with subsequent NRC evaluation !
to ensure the adequacy of NNECO's actions. NNECO letter A09961, dated December 19, l 1991, described NNECO's review of this concern.
BackFround NRC Inspection Report 50-336/91-20 (IR 91-20), section 5.3, described a previous inspection of boric acid flow control system corrective maintenance. IR 91-20 concluded (in part) that NNECO efforts to identify, troubleshoot, and repair boric acid system equipment deficiencies were appropriate.
NNECO used PORC .wid IC procedures and SPs and their associated data sheets as the ~
vehicle to document calibration data and test equipment for safety related I&C components, as described in the station surveillance program (reference ACP-QA-9.02). For non-safety related I&C components, either the applicable AWO or an IC procedure document calibration data and the test equipment used to do I&C maintenance and surveillance.
f 30 Examples of non-safety related maintenance and surveillance procedures, that were not required to be done by MP2 Technical Specifications (TS), included IC 2427A and IC 2428B. IC 2427A .
described the steps to ensure all recorders were periodically calibrated, cleaned, and functionally tested. The data sheets (e.g., I&C Form 2427A-2) associated with IC 2427A gave a monthly list of recorders to be calibrated and required recording of test equipment in each appropriate instrument loop folder. IC 2428B was the procedure for calibration of the boric acid and primary makeup water (PMW) to volume control tank (VCT) systems. IC Form 2428B-1 listed the test equipment used in doing IC 2427B, as "DVM," " Pressure Gauge," and "Transmation" (model 1040).
NNECO also documented certain I&C information on instrument calibration data sheets in I&C Department working files. I&C Instruction 3.02 established standard forms, including Form -
3.02-1 A, for use by I&C Department personnel. As described in I&C Instruction 1.10,1&C Department Form 3.02-1 A recorded calibration data in instrument loop folders for all instruments that were not incorporated with a PORC approved I&C procedure. Form 3.02-1 A was a general purpose form to record instrument calibration data, including spaces to record a supervisor's signature and test equipment numbers. ,
Station Form (SF) 1018 identified the QA records retention and turnover schedule for the MP2 I&C Department. I&C Department Form 3.02-1A was not a QA record, but was included in I&C Department working files as a general reference by I&C Dwdment personnel.
Assessment The inspector questioned if NNECO used as test equipment any calibrated voltage supplies that were load sensitive. NNECO stated that Transmation model 1040 was the only such device used as plant test equipment at MP2. In Unit 2 I&C Technical Bulletin TIB 89-5, NNECO described the need to monitor, with a QA digital voltmeter, the output of Transmation model 1040 when used as a voltage source. 'Ihis was because the digital display in the model 1040 did not always represent the actual voltage output since some circuitry was load sensitive at low impedance loads. If the Transmation model 1040 were used as a current source, for example when calibrating flow recorders, there was no need to use a QA digital voltmeter.
An example of this was boric acid flow controller FR-210Y trouble shooting that was done on September 20,1991, per. AWO M2-91-08650 using IC 2428B-1. AWO M2-91-8650 listed the -
test equipment as digital multimeter (DMM - QA #926) and Transmation (QA #1018). _ Also, the instrument calibration data sheet in the I&C Department instrument loop folder for FR-210Y contained a list of the'same test equipment. Another example was the calibration of FR-210Y done on February 20,1991, per AWO M2-90-02080 using IC 2427A-2. AWO M2-90-02080 stated the test equipment was " equipment used by instrument folders." The FR-210Y instrument calibration data rheet listed the Transmation test equipment, but did not list the DMM. NNECO stated a DMM was used as =*=y to complete IC 2427A; however, NNECO inadvertently omitted documentation of DMM use on the instrument calibration data sheet. Separate files in Nuclear Records and the Metrology 12b contained detailed records regarding the usage and
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calibration of test equipment, including DMMs and Transmation model 1040s, as required by l
ACP-QA-10.04.
The inspector found no I&C Department procedure or other applicable NNECO procedure that i promulgated specific administrative instructions for completion of Form 3.02-1 A. Because Form :
3.02-1 A was not used as a QA record, there was no regulatory requirernent to provide such an l I&C Department proceduie.
Conclusions ,
1 Based on review of applicable documentation and discussion with cognizant NNECO personnel, the inspector concluded that NNECO had adequate guidance in TIB 89-5 describing the use of a calibrated voltmeter for measuring the output of Transmation model 1040 test equipment.
Further, NNECO adequately identified and maintained QA records for I&C maintenance and surveillance work on non-safety related recorders.
There was evidence that NNECO may not have always listed on instrument calibration data sheets DMMs used with Transmation model 1040's, and the supervisory review of such data sheets did not always identify such discrepancies. NNECO promptly corrected specific examples noted during the iaW and initiated action to sample the adequacy of additional instrument calibration data sheets. The inspector concluded that management expectations for maintenance of instrument calibration data sheets in I&C Department working files were not clearly defmed I in I&C Department Instructions. Because the instrument calibration data sheets were not QA records, and because flow recorders such as FR-210Y and RJR-9373 were not safety related, there was no regulatory requirement to document test equipment usage on instrument calibration data sheets. The inspector had no further concerns regarding this matter.
14.0 SAFETY INJECTION TANK PRFMURE SWITCHES
- Ihe NRC provided a concern regarding the adequacy of MP2 safety injection tank (SIT) pressure switches (PSs).
Background
MP2 Technical Specifications (TS), section 3.5.1.d, has a limiting condition for operation that requires SIT cover pressure of between 200 and 250 psig when in modes 1 or 2, and when in mode 3 if pressurizer pressure is equal to or greater than 1750 psia. NNECO ensuits SIT cover pressure is between 200 and 250 psig at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using OPS Form 2619A-1.
High pressure and low pressure switches are set to alarm in the control room prior to exceeding the MP2 TS allowable range. Operating procedure OP 2306 is used to make adjustments in SIT cover pressure. SITS are initially pressurized to approximately 215 psig prior to plant startup.
A desirable operating practice was to have no control room annunciators illuminated during steady state full power operation. Control room panel C-01 annunciator windows C-10, C-11,
32 C-12, and C-13 were for SIT high pressure alarms. NNECO indicated in AWO M2 91-04556, dated September 11,1991, that previous SIT high pressure alarm trouble reports (TRs) related i to alarm " lock in" were too frequent, and may have been caused by a wide deadband on the PSs associated with these alarms. Operations believed an additional factor was the pressure increase caused by the SITS normally warming from ambient temperature to approximately 110*F during l plant startup. NNECO initiated AWO M2-9104556 to establish, if possible, a reliable method j of adjustment for SIT high pressure alarm pressure switches (PSs) 313,323,333, and 343. i These PSs were Custom Component Switches, Inc. (CCS) model 604GR3-353S. j NNECO determined in evaluation MP2-CD-674 that only the pressure boundary of PSs 313, !
323, 333, and 343 was safety related. Failure of these pressure switches during a seismic event - )
could vent the associated SIT and prevent the affected SIT from performing its safety related i function of supplying borated water to the reactor coolant system. The PS alarm function was -i net safety related. SIT pressure indication was available on Control Room instrumentation and j on the MP2 process computer, j NNECO initiated PDCR 2-090-91 to replace the SIT CCS model 604GR3-353S pressure l switches with CCS model 5NN-K5-U9-CIA-PCPB pressure switches. The new switches had l a nominal 5 psig deadband that NNECO believed would avoid " lock in" alarms during plant ,
operation.
Assessment l i
According to documentation associated with AWO M2-91-04556, CCS model 604GR3-353S :
pressure switches had an adjustment range of 60 to 225 psig (increasing) and 45 to 210 psig I (decreasing). Because the SIT high pressure Pss setpoint and reset values were in excess of the ;
switch nameplate rating, the inspector questioned if the CCS model 604GR3-353S PSs were. !
adequate for their intended purpose. NNECO compared switch nameplate data with procurement i documents and found differences in the switch adjustment range. For example, the switch .j nameplate listed an increasing adjustment range of 60 to 225 psig, but drawing 25203-29115, i sheet 37, indicated the adjustment range on increasing pressure was 50 to 250 psig. NNECO l stated that the nameplate adjustment range information was incorrect. Because these switches i will be replaced per PDCR 2-090-91 and the switches were not readily accessible during plant operation, NNECO stated there was no need to correct the evitch nameplate. j CCS vendor manual VTM2-167-002A was a maintenance and parts replacement manual for standard commercial switch models 604 and 605GC. According to the equipment list for i VTM2-167 002A, it applied to PSs 313, 323, 333, and 343.
The inspector questioned if CCS model 604GR3-353S pressure switches were adequate for their current application. NNECO stated that these devices were adequate because they were appropriately mounted on seismically qualified supports and Nonconformance Report 291-225
- concluded these devices were acceptable for use-as-is. Also, the pressure retaining parts of these .
pressure switches had a proof pressure rating of 4500 psig, but the normal operating pressure !
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- l 33 l was less than 250 psig.
The control room annunciator response book (CRAB) and OP 2306 used a value of 250 psig for the SIT high pressure alarm, but the actual setpoint was 245 psig (increasing). ACP-QA-3.02A, !
secten 6.8.3, stated that "if actions are required based on receipt of an annunciated alarm, then l list the setpoint of the alarm for case of verification." Also, ACP-QA-3.02A, section 6.8.4, !
stated to " provide an acceptable range instead of a point value, when applicable." Calibration i data indicated PS reset occurred in the range of 227 to 236 psig. When questioned by the }
inspector, NNECO stated that in some cases procedures also used MP2 TS limits rather than ;
actual setpoints for values described in the CRAB, cps, and SPs. l t
Conclusions l Based on review of applicable documentation and discussion with cognizant NNECO personnel, j the inspector concluded the CCS model 604GR3-353S PSs were adequate for epplication as SIT l high pressure alarm switches. Also, the inspector concluded that, unier, workers thoroughly (
researched relevant information, inaccurate nameplate data could case worker confusion -
regarding the adequacy of CCS model 604GR3-353S P'3s. l Finally, the inspector concluded that NNECO may not have adequately described in all cases ;
alarm setpoints in the CRAB, Ops, and SPs, as required by ACP-QA-3.02A. NNECO agreed j to evaluate this matter, take appropnate action as nemry, and respond back to the NRC.
i 15.0. WORK CONTROL CENTER The NRC providrd various concerns related to the Work Control Center (WCC). The concems ,
were generally on the subjects of safety tags, work orders, and administration of the WCC.
'Ihirty two (32) concerns were associated with tagging and 41 were associated with other WCC ,
issues. Previous NRC actions in response to these concerns included inspection of specific j issues that may have had some potential safety significance and referral to NNECO with subsequent NRC evaluation.
Background
The NRC documented in inspection reports its review of previous employee concerns and other issues related to the WCC. For example, InW Report 50-336/91-04 identified (open) '
unresolved item (UNR) 50-336/91-04-02 as post maintenance control of safety related- i equipment. Future NRC inspection of this item will include verification of licensee corrective i actions to strengthen control of post maintenance activities. Another example was (open) UNR 50-336/91-28-01 in Inspection Report 50-336/91-28 (IR 91-28). Future NRC inspection will ;
assess the adequacy of the equipment tag-out restoration process.
3
7 s l 34 IR 91-28 addressed (open) UNR 50-336/91-04-02. As described in IR 91-28, section 4.2.1, !
PMMS field entry controls in the WCC appeamd to be informal and there was no specific procedural guidance to prescribe WCC activities. l Assessment The WCC at MP2 is located in an office adjacent to the Shift Supervisor office at the main !
control room. Using seven operating crews in a six shift rotation, NNECO established a j schedule that has one crew serve as the WCC staff for seven consecutive weeks. The WCC-primarily operates during the day shift on weekdays, but also operates at other times (e.g.,
during outages) as --ry. Thus, the WCC has a highly qualified staff available to support on duty operating crews during peak work periods. !
A major WCC function is to eliminate na-=ey distraction of the on duty operating crews. ;
During peak periods, the WCC effectively serves as the primary point of contact between work ;
crews and plant operators. Personnel could discuss work activities such as safety tag clearance ,
and job authorization in an area away from the main control boards. Also, because of the reduced administrative burden allowed by the WCC, Shift Supervisors (SSs) and Supervising 1 Control Operators (SCOs) have more time to focus on important operational activities.
WCC personnel have the qualification and authority to accomplish their responsibilities for processing work orders and station tags, as defined in ACP-QA-2.02C and ACP-QA-2.06A, respectively. The Operations Work Coordinator (OWC) is a qualified SCO with an active Senior Reactor Operator (SRO) license. To ensure the on duty SS and SCO are aware of WCC actions and changing conditions, there was close coordination between the OWC and the on duty
- SCO as well as other WCC and operations personnel.
NNECO r-iw the need for administrative instructions for WCC activities. As an interim measure, January 15,1992, the WCC SS issued a memorandum to all department heads that ,
defined WCC expectations and proposed standard guidance for WCC activities.
To help reduce the potential for tagging errors and facilitate the tagging process, NNECO is developing the Millstone Automated Tagging System (MATS). This was a computer based :
information system with various capabilities that includes identifying standard clearance tag lists :
and printing information for safety tags. !
WCC personnel conduct their activities in a thorough and diligent manner. Communication among WCC personnel is generally informal, but adequate. WCC coordination with work group supervisors and support forjob leaders appear to be highly responsive. WCC liaison with on duty operations personnel is very effective. 'Ihe inspector found no instance in which WCC personnel failed to adequately execute their responsibilities, as defined in applicable ACPs.
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35 Conclusions Based on discussion with cognizant NNECO operations, maintenance and I&C personnel, observation of WCC activities, and review of relevant documentation, the inspector concluded that MP2 WCC helps reduce SS and SCO administrative burdens during peak work periods and supported work group needs. This is considered a management strength. Further, if effectively implemented, NNECO effgrts to standardize WCC activities through written instructions and to implement innovative programs such as MAPS would result in an excellent enhancement of MP2 work control activities.
16.0 NONCONFORMANCE REPORT 291-272 l
The NRC provided a concern related to a 10 yde reference power supply in the reactor protection s;rstem (RPS) core protection calculator (CPC). The concern was that this power supply may have been modified, without appropriate design controls, by drilling an access hole in the plastic case that covered a circuit board. NRC disposition of that concern involved providing the concern to NNECO for review and resolution, with subsequent NRC evaluation to ensure the adequacy of NNECO's actions. NNECO letter AO9962, dated December 19, ,
1991, described NNECO's review of this concern.
Assessment ;
NNECO inspected all similar CPC power supply modules, as described in A09962. Since the access hole for the affected power supply module did not appear on the manufacturer's product drawing, NNECO suspected this modification was made after original installation. Accordingly, NNECO initiated Nonconformance Report (NCR) 291-272 as required by ACP-QA-1.20 and ACP-QA-5.01. Because t!.e hole did not degrade the function of the plastic case, which was circuit board support, NNECO determined that the affected power supply modules were acceptable for use-as-is.
The inspector questioned if this was an isolated instance or if there were other similar examples of modifications made without adequate design change controls. NNECO stated it was not #
aware of any similar modification of other power supply modules that did not have the requisite ;
documentation. Because this was an apparently i;,alated incident with no significant impact on l
plant safety, there was no requirement to initiate either a Plant Incident Report (PIR) per ACP-QA-10.01 or a Corrective Action Request (CAR) per ACP-QA-10.10. l1 Conclusions Based on review of applicable documentation and discussion with cognizant NNECO personnel, the inspector concluded that NNECO adequately resolved this matter in NCR 291-272. ,
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36 17.0 SUPPLEMENTARY INFORMATION ON PRIOR INSPECTION ISSUES The following are either clarifications or additional documentation regarding issues the NRC described in previous insp~+iaa .W.
17.1 Drawing ContmP 1 l
Employees found it convenient to use drawings located within vendor technical manuals, but '
expressed concerns that drawings within vendor technical manuals were not up to-date. Vendor drawings typically depicted standard equipment designs and were not necessarily the exact configuration installed at MP2. The licensee was committed to maintenance of vendor drawings and incorycwating them, as necessary, with the NUSCO controlled drawing system. Licensee I&C Department management stated it was considering an enhancement of current NNECO practices such that vendor manuals would include appropriate references to NUSCO drawing numbers.
Inspection Report (IR) 50-336/91-27, section 7.1, documented the findings of a review made of the drawing control system and its use during activities at the station. The report incorrectly stated that there were no administrative control procedure requirements to verify the latest drawing information prior to the use of a drawing for quality work. This was an error. ,
Administrative Cor. trol Procedure ACP-QA-3.03, Document Control, Revision 33 in section 6.2, Design Document Control, requires that persons using drawings for quality work activities are responsible for verifying they have the latest revision of a drawing by referencing the Drawing Status File within the Generation Records Information Tracking System.
This information does not change the report conclusion that some personnel were not using the drawing control system when required, but it does correct the findmg of a deficiency within the ,
system of administrative controls. ,
17.2 Tool and Document Cantamla=# tan During AWO M2-91-06732 i
'Ihe NRC provided a concern related to contamination of tools and a procedure package while ;
doing AWO M2-91-06732 in July 1991, and the NRC promptly inspected the concern.
Background ;
I NRC Inspection Report 50-336/91-18 (IR91-18), section 4.0, documented the NRC review of various radiological control issues related to posting and contml of radiological areas, radiation monitor RM-8132, and a spent fuel pool area frisker. In part, that report found that _" posting l of contaminated, high airborne radiation and high radiation areas was observed to be appropriate ,
with respect to boundary identification, locking requirements, and hold points." The following '
NRC assessment of the employee concern related to AWO M2-91-06732 was donein July 1991, V
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but was not specifically described in IR91-18.
Assessment
. Two work activities in the vicinity of stack gas radiation monitor RM-8132B took place during i the period of July 2 - 3,1991. One activity was calibration of RM-8132B using surveillance procedure SP 2404AF. The other activity was replacement of the RM-8132B sample fan .
i assembly blower and drive belt per AWO M2-06732. De concern was that work done per
' AWO M246732, which had no radiation work permit (RWP) requirements, may have caused contammation of an I&C technician's tools and documentation for the SP 2404 AF work package.
= MP2 health physics (HP) was involved with the issue when the I&C technician exited the work i
area on the morning of July 3,1991. In addition to contamination found on the papers and tools, HP surveys found 15K dpm in the work area. Access to the work area was then controlled as a contaminated area. Other floor areas accessible to personnel were clean. There was no contamination of personnel. HP review of the maintenance work activity identified no contamination in the fan removed on July 3,1991, or in the maintenance shop. NNECO assigned decontamination personnel to clean the affected area.
HP concluded the maintenance work did not cause the contamination and that there were -
acceptable radiological work controls. HP did identify leakage from a ventilation housing near '
the job site as the source of the contamination. NNECO took action to contain any further leakage from the ventilation housing and to continue decontamination of the area.
On July 4,1991, the inspector reviewed HP survey sheets and observed the job site. The ,
t inspector found that NNECO removed most of.the contamination and that NNECO had appropriate barricades and postings to control access to the area. :
Conclusion Based on review of HP records and physical inspection of the work area, the inspector concluded NNECO adequately responded to the contamination event. De inspector had no further questions on the adequacy of work control, l I
17.3 ASI Curve for LHGR 4
As described in NRC Inspection Report 50-336/89-08, section 5.4, on May 1,1989, the licensee found that the incore analysis (INCA) program produced unexplained results during power ascension testing. For example, a coefficient error resulted in the INCA program being unable to precisely measure LHGR over the full length of the core. The licensee resolved this matter and declared the INCA system operable on May 6,1989. The NRC evaluated licensee actions and found no inadequacies.
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38 On May 5,1989, the NRC moeived an employee concern asserting the LHGR check using the curve for axial shape index (ASI) was inadequate because the excore detectors were not calibrated against the incore detectors at equilibrium xenon conditions. As described in NRC Ine: *iaa Report 50-336/89-11, section 11.1, the NRC did a follow-up inspection of this and other related matters, and concluded the following: "In summary, no safety inadequacy was .
found in the operations performed while an INCA program problem' existed, in the 'A' RPS channel input being bypassed to the high power averaging cucuitry, or in the incore/excore measurements taken before equilibrium xenon was reached."
NRb Inspection Report 50 336/89-13, section 2.2 (A.6.15, A.7.2, A.12.1,2, and 3), described additional NRC follow-up inspection and evaluation of employee concems related to the INCA program and licensee observance of LHGRs. The NRC concluded their were no safety i t3 concerns. Also, the licensee "napW the INCA program pmblem and acted in a 1 conservative manner while the problerr; existed." Based on this previous inspection effort, this concern is considered resolved.
17.4 Steam Generator level Calibration Procedust SP 2402D ]
The NRC received a concern in September 1989 related to the adequacy of steam generator level calibration surveillance procedure SP 2402D. The concern was that SP 2402D could not be done as written, but was previously completed in 1989. The NRC referred this issue to NNECO for resolution, as documented in NRC Inspection Report 50-336/89-23, Appendix A, item number A.25.02.
Assessment NNECO stated that during a training class for 1&C personnel in 1989, it identified two typographical errors in SP 2402D, section 7.7, associated with the Auviliary Feedwater (AFW) automatic AFW initiation system test. Steps 7.7.33 and 7.7.34 of the procedure referred to test pin "13/24," but the correct reference was 13/23. In addition, some hand switch nornenclature differences were noted.
4 NNECO completed SP 2402D on April 18,1989, per AWO M2-88-02316. The procedure in effect at that time was SP 2402D, revision 9, change 2, dated March 8,1989. NNECO issued SP 2402D, revision 9, changes 3 and 4 on August 23, 1989, and September 19, 1990, respectively. Changes 3 and 4 corrected the typographical errors, added a new step to avoid transmitter alignment problems at system pressure, and replaced hand switch numerical designations with noun names.
The inspector reviewed SP 2402D, revision 9, changes I through 5, and AWO M2-88-02316.
As documented on 1&C Form 2402D-1, revision 8, test results in steps 7.7.36, 7.7.T' and 7.7.38 met ay-;; = criteria. 'Ihose steps could not meet =~*penace criteria unless pw. ing steps 7.7.33 and 7.7.34 were done in a technically correct mar.ner.
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ACP-QA-3.02E, section 6.2, stated that " full and totai compliance is expected" for those procedures used to do surveillance and testing as specified in the MP2 TS. SP 2402D was used to meet several MP2 TS surveillance requirements. ,
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Conclusions Based on review of applicable documentation and discussion with cognizant NNECO personnel, !
i
) the inspector concluded NNECO satisfactorily completed SP 2402D per AWO M2-88-02316, l
l but attention to detail in procedural <=apli=* may act have been adequate in all respects The 4
hand switch nomenclature differences and typographical errors described in changes 3 and 4 ,
could have been identified and corrected prior to completion of SP 2402D in April 1989.
Because tes results met ep*mam criteria, the inspector concluded the above discrepancies were not fu:Wonally significant and did not compromise nuclear safety.
17.5 Rigging Practices for Two Too Holst ne NRC provided a concern regarding two ton hoist rigging practices used for MP2 polar crane i modification during the l'90 refueling outage. The concern asserted that there were electrical l
- cables over the crane cele, Fmre was interference between the crane cable and a guard rail, and there were sharp bends en N whip line when attached to load. De NRC promptly referred this ,
i issue to the Millstone Safety Office for resolution.
The NUSCO Safety Office stated that it promptly inspected this issue and took action to resolve l relevant concerns. Further, there were currently no similar unresolved industrial safety issues.
i-2 The inspector had no further concerns.
4 18.0 MANAGEMf'NT MEETINGS i
j On February 7,1992, an exit interview was madM-1 with NNECO's senior site representatives
- to summanze the observations and conclusions of this inspection. NNECO did not indicate this
! inspection involved any proprietary information.
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A'ITACHMENT 1 PERSONS CONTACTED NORTHEAST NUCLEAR ENERGY COMPANY !
i Mr. Jack Amatucci, MP2 I&C Engineer ;
Mr. Richard Armour, Senior Control Operator, MP2 Operations '
Mr. Teny Arnett, MP2 I&C ,
Mr. Ralph Bates, MP2 Engineenng Supervisor -
~~
.p. John Becker, MP2 I&C Manager
' Mr. Edward Bireley, Unit Services Mr. 'Ihomas Blanchard, MP2 Engineering i Mr. Steven K. Brinkman, MP2 Operations Mr. David Clark, MP2 Ope ations !
Mr. David L. Coleman, Mechanical Engineering i Mr. Bruce Danielson, Nuclear Training Mr. 'Ihomas Dembek, Emergency Planning !
Mr. Keith D. Deslandes, MP2 eng'meering i Mr. John F. Follett, NUSCO Safety ;
Mr. Richard Goldsmith, Nuclear Training Mr. Mark Heinonen, MP2 Maintenance General Supervisor t Mr. John Kiss, Nuclear Training '
Mr. Steve Main, Nuclear Records Supervisor !
Mr. Thomas W. McCance, Emergency Planning Mr. Michael Mullin, MP2 Shift Supervisor ,
Mr. Stephen Myers, Shift Supervisor, MP2 Operations l Mr. Charles Nelson, MP2 Operations j
Mr. John W. Riley, MP2 Engineering Manager .
Mr. Robert F. Rowe, MP2 Maintenance Supervisor :
Mr. William R. Salen, MP2 I&C
[
Mr. Raymond Schleicher, MP2 I&C General Supervisor l Mr. Jeffrey Smith, MP2 Operations Manager Mr. Peter Smith, MP2 I&C Supervisor
'Ihe inspectors also contacted additional administrative, operations and technical personnel during !
the inspection.
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REFERENCES l l
Plant Design Change Request 2457-90, Service Water System Sample Valve Installation in SW !
Strainer D/P Tubing, revision 0,11/26/90 ;
i Work Order AWO M2-9415617, Install Swagelock " Tees" and Whitey Valves as Described in I PDCR No. 2 057-90, 12/12/90 l I
I&C Instruction 1.10, I&C Department Instrument loop Folder, revision 1,2/13/90 l
I&C Instruction 3.02, I&C Department Forms, revision 1, 6/14/90 l i
Maintenance Procedure IC 2422B, Gaseous Process Radiation Monitor Calibrations, revision 2, change 4, 3/18/91 ,
i I&C Form IC 2422B-1, Gaseous Process Radiation Monitor Calibration RM 8011 Calibration i Data Sheet, revision 2, change 1, 3/18/91 ,
i I&C Form IC 2422B-2, Gaseous Process Radiation Monitor Calibration RM 8434B Calibration l Data Sheet, revision 2, change 2, 3/18/91 l IC 2422D, Particulate Rulistion Monitor Calibration, revision 2, change 2, 3/18/91 ,
IC 2427A, Recorders Annual PM, revision 10, change 1, 5/30/90 i
IC 2428B, Volume Control Tank Make-Up System Calibration, revision 4,11/7/85_ _
I&C Form 2428B-1, Volume Control Tank Make-Up System Calibration Data Sheet,' revision l 5, 11/7/85 i i
IC 2436A, Safety Related Instrumentation Valve Line-Up Verification, revision 3,9/17/91 l
I&C Form IC 2436A-1, Safety Related Instrumentation Value line-Up Verification, revision 2, !
9/17/91 Station Procedure SP 2401F, Reactor Protection System High Power Trip Test, revision 8, >
change 2,12/11/91 SP 2401J,7hermal Margin / low Pressure Calculator Test, revision 10, change 5,1/17/91 SP 2401R, CEA Withdrawal Prohibit (CWP) Functional Test, revision 0,' change 1, 1/17/91 1
SP 2402D, Steam Generator level Calibration, revision 10, 10/31/90 SP 2402M, Functional Test of Auto-Aux. Feedwater Initiation Imgic, revision 5, 10/16/91 SP 2404 AF, Unit 2 Stack Gaseous Process Radiation Monitor RM-8132B, Calibration, revision 1, change 5, 7/5/91 SP 2404AR, Unit 2 Stack Gaseous High Range Radiation Monitor, RM 8168, Functional Test, revision 2,7/12/91 Form SP 2404AR-1, Stack Gas Radiation Monitor High Range Functional, revision 3, change 2, 9/11/91 I
SP 2404 AS, High Range Stack Gas Radiation Monitor RM 8168 Calibration, revision 0, change 1, 9/16/87 Form SP 2404AS-1, High Range Stack Gas Radiation Monitor RM 8168 Calibration, revision 0, change 1,9/16/87 SP 2604A, HPSI Pump Operability Fac. I, revision 6, change 2,11/21/91 SP 2604E, Facility 1 High Pressure Safety Injection System Alignment Check and Valve Operability Test, revision 7,4/10/91 SP 2604F, Facility 2 High Pressure Safety Injection System Alignment Check and Valve Operability Test, revision 7,4/10/91 SP 2619A, Control Room Shift Checks, revision 8,11/26/91 OPS Form 2330C-1, Chilled Water, revision 9,4/10/91 OPS Form 2305-1, Spent Fuel Pool Cooling, revision 7,12/17/91 OP 2306, Safety Injection Tanks, revision 12, 1/15/92 OP 2387E, Control Room Annunciator Response, revision 1,11/27/91 OPS Form 2619A-1, Control Room Daily Suncillance, revision 32,1/22/92 OPS Form 2604 A-1, High Pressure Safety Injection (HPSI) Pump Operability Test Data (Facility 1), revision 5, change 1,10/31/90 -s OPS Form 2604E-2, High Pressure Safety Injection System Valve Alignment Facility 1, revision 11, 4/10/91 2
I o 9: ,
OPS Form 2604F-2, Facility 2 High Pressure Safety injection System Valve Alignment, revision i 10, 4/10/91 [
OPS Form 2604L-2, LPSI Valve Alignment Check, Facility 1, revision 11, change 1, 5/15/91 l L
OPS Form 2604M-2, LPSI Valve Alignment Check, Facility 2, revision 10, change 1, 5/15/91 F
OPS Form 2619A-1,' Control Room Daily Surveillance, revision 30, change 1,7/23/91 4
OPS Form 2619A-2, Control Room Daily Surveillance, mode 3 & 4, revision 19, change 1, ,
9/16/91 !
i Drawing 25203-29115, sheets 37, 38 and 39, Switch, Adjustable Gage Pressure, 5/16/73 Drawing 25205-32007, Millstone' Site Emergency Operations Center Single Line Schedules &
Symbols Waterford Conn., revision 8,7/11/90 3
Drawing 25205-32008, Emergency Operations Center Auto. Transfer Switch Schematic l Waterford Conn., revision 1, 5/18/82 l Drawing 25205-39002, sheet 3, Transfer SW-CAB ASSY (Wiring Diagram), revision 1,5/18/82 3
Drawing 25203-39092, sheet 14C, Nuclear Measurements Corp., Power Flow Diagram', revision 2, 1/30/84 4
i Drawing 25203-39092, sheet 14E, Nuclear Measurements Corp., Power Flow Diagram, revision 1
Piping and Instrumentation Diagram (P&ID) 25203-26005, sheet 3, Condensate Storage & Aux. l Feed, revision 12, 1/14/91 P&ID 25203-26015, sheet 1, L.P. Safety Injection System, revision 9,10/17/91 ;
P&ID 25203-26015, sheet 2, High Pressure Safety Inj. Pumps, revision 5,10/17/91 l J
P&ID 25203-26022, sheet 1, R.B.C.C.W. System R.B.C.C.W. Pmps & Heat Exchangers, revision 21,11/21/90 ,
I P&ID 25203-26023, sheet 2, Spent Fuel Pool Cooling & Cleanup Sys, revision 4, 4/18/90 P&ID 25203-26027, sheet 2, Turb. Bldg. Intake Str., Whse. & D.G. Rms. Chilled Water System, revision 16, 11/21/90 Plant Design Change Record Evaluation PDCE MP2-90-032, Replacement of Magnahelic with 3
- - -, .. , , - . - , - - . - , - .~. - . - - -
- or l
~
i
'4 , , f Photohelic Switches for FIS-80ll, 8123, 8132, 8145, 8262, 8434, 9095, closcout date 9/13/91 .
PDCR 2-112-79, Install pressure Gauges, 7/13/79
{
Administrative Control Procedure ACP-QA-1.15, Management Program for Maintaining !
Emergency Preparedness, revision 12, 10/25/91 !
4
- ACP-QA-2.02C, Work Orders, revision 28,11/29/91
?
ACP-QA-2.06A, Station Tagging, revision 19, 1/15/92 ACP-QA-2.12, System Valve Alignment Control, revision 10, 5/29/90 !
ACP-QA-2.20, independent Verification, revision 2,10/2/90 !
ACP-QA-2.21, Administration of Plant Design Change Turnover and Preoperational Testing, !
revision 1,12/31/91 j ACP-QA-3.02A, Writer's Guide For Millstone Procedures, revision 2,4/13/90 .
i
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ACP-QA-3.10, Preparation, Review, and Disposition of Plant Design Change Records PDCRs l (NEO 3.03), revision 4,7/20/91 !
ACP-3.23, Control of Vendor Technical Manuals, revision 3, 7/16/91
)
ACP-QA-4.01, Plant Housekeeping, revision 15, 10/6/87 ,
ACP-6.01, Control Room Procedure, revision 22,7/5/91 l ACP-6.01A, Structured Communications NOP 2.18, revision 1,8/24/90 -,
ACP-6.22, System and Component labeling,' revision 0, 3/8/91 !
i ACP-QA-3.10, Preparation, Review, and Disposition of Plant Design Change Records PDCRs ,
(NEO 3.03), revision 4,7/20/91 !
ACP-QA-9.02, Station Surveillance Program, revision 20, 8/14/91 l
_ ACP-QA-9.02B, Unit 2 Surveillance Master Test Control List, revision 16, 10/5/90
{
i ACP-QA-10.04, Nuclear Plant Records, revision 31, 7/16/91 !
t i
Station Form SF 210, Tag Img Sheet, revision 8,1/15/92 '
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ggg- ema -- - - m,s -- -s~a-
i l
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NUREG-0737, Supplement 1, Requirements For Emergency Response Capability (Generic Letter No. 82-33),12/17/82 Northeast Utilities memo NSE-M-86-59, MPDC. T G Review of PDCR 2-112-79,6/20/86 Northeast Utilities memo PSE-SA-89-061, Millstone Unit No. 2 Evaluation of Pressure Gages ,
for PDCR-2-112-79, 3/6/89 l Northeast Utilities memo MCE-SA-91-105, Millstone Unit No. 2 REF 91 Evaluation of :
Pressure Gages, 10/28/91 i
Northeast Utilities memo NE-83-R-474 (CR 5127), Emergency Core Cooling System Operability '
Requirements, 9/23/83 Calculation 2-112-79-1067 GP, Evaluation of Pressure Gages for PDCR-2-Il2-79, revision 1, ,
t 6/20/89 Specification 7604-MS-64, Nuclear Code and Seismic Classification for Instrument Lines, .
Sampling Lines and Inline Instruments, revision 3,12/19/75 Specification 7604-MS-66, Design Guide For Seismic Class I Instrument Tubing Installation, ;
revision 3,10/12/73 Specification 7604-M-467 B, Pressure Switches,4/5/76 l
Instrument Index 7604-MS-60, Millstone Nuclear Power Station, 12/22/74 Custom Component Switches, Inc., Myumle Gage Pressure Switches Models 604 and 605GC r Maintenance and Parts Replacement Manual, VTM2-167-002A,7/23/73 l
I r
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a 69) 5 NORTHEAST UTILITIES
== l .e:IeU v$sS' c1nmi On.cu . ssio.n sir,si. 8:,iin. Conn.ci coi k -
P.O. BOX 270 HARTFORD CONNECTICUT 06141-0270 L L J C".' E.N '"". cow ~,,
, 5 (203) 665-5000 i
September 30, 1991 Docket No. 50-336 A09829 Mr. Charles W. Hehl, Director Division of Reactor Projects U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406
Dear Mr. Hehl:
Millstone Nuclear Power Station, Unit No. 2 RI-91-A,-0210 ?
l We have completed our review of an identified issue concerning activities at Millstone Station. 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 transmittal l 1etter and our response have received controlled and limited distribution on a "need-to-know" basis during the preparation of this response.
ISSUE 210-1:
" Millstone 2 turbine sampling system valves with a prefix of '2-S' are not labeled in the field. In addition, the sodium analyzer 'AE 7764 A5', is incorrectly labeled as 'AE 7764 A6' on sheet 3 of drawing 25203-26025."
\
RE00EST: I "Please discuss the validity of the above assertion. If the above conditions are valid, please notify us why this was not previously corrected by your labeling program and what corrective actions you have taken to prevent recur-rence. Please provide us with an assessment of the safety significance of any !
identified deficiencies, including any generic considerations."
RESPONSE 210-1:
The assertion concerning the sampling system valve labels is valid if the stated analyzer is AE7784 rather than AE 7764. We were made aware of the drawing problem in early September, 1991. The Millstone Unit No. 2 Engineering Department has identified that the Piping and Instrument Diagram (P&lD) does not agree with the as-built condition and is in the process of updating the drawings to reflect, the as-built conditions.
%umn t -
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Mr. Charles W. Hehl A09829/Page 2 September 30, 1991 In March 1991, a new Administrative Control Procedu e,M. ((MMt'daI3Id C.'d,.._..i--4.atm11ng ize 'twn written and issued to provide g6TUance on a standard-labeling program for the Millstone site. Also in March, a Controlled Routing (CR 8139) was issued to track the progress of each of the units in complying with the requirements of the ACP. The labeling effort includes valves, ma.ior components, and instrument and gauge labeling. The sampling system labeling valves identified in the assertion are part of this program. The valve efforts are expected to be completed by the end of 1992. Currently the Operations Department is holding discussions with other departments in the unit in an effort to establish who will normally operate and be responsible for labeling valves such as those associated with radiation monitors, sample systems, etc.
As part of the progra:n implemented under ACP 6.22 a label request form has been generated which allows anyone finding a missing, incorrect, or deterio-rating label to bring it to the attention of the label coordinator or Opera-tions Manager for prorct action. ~
As part of the labe ing program under ACP-6.22, tt e initial labeling of a system is to be verified by a complete walkdown of the system using system checklists and P& ids to determine that all system ccm1ponents are labeled and that the label nomenciature matches the P&ID identification and system check-list description for each component. Thus, if other similar problems exist.
they should be routinely identified and corrected as part of this program.
ISSUE 210-2:
" Circuit changes had been made to the Millstone 2 n.ain generator hydrogen !
monitor calibrationwithout the preparation procedure of a modification package. As a result, the is inadequate, j changes have not made." and appropriate procedure and drawing RE00EST:
"Please discuss the validity of the above assertion. If the above conditions are valid, please nctify us of the corrective act'ons you have taken to prevent recurrence.
Please provide us with an assessment of the safety signi-ficance of any identified deficiencies, including any generic considerations."
RESPONSE: l l
We are still investigat'ng this matter and will respon.: when the investigation is complete. We currer:ly plan to respond by October :4,1991.
9 Mr. Charles W. Hehl A09829/Page 3 September 30, 1991 After our review and evaluation of this issue, we find that it did not present any indication of a compromise of nuclear safety. We appreciate the opportu-nity to respond and explain the basis of our actions. Please contact my staff if there are further questions on any of these matters.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY Edwaff J. Mrocefa Senitr Vice President cc: W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 E. C. Wenzinger, Chief Projects Branch No. 4, Division of Reactor Projects E. M. Kelly, Chief, Reactor Projects Section 4A J. T. Shedlosky, U.S. Nuclear Regulatory Comission, Millstone i
- . (9 8
-. \ e UNITED STATES NUCLEAR RESULATORY COMMISSl!N 8, p f REGION I 475 ALLENDALE ROAD KING oF PRUSSIA. PENNSYLVANIA 19406-1415
%**"*g OCT 0 91991 50-336 .T Docket Number: ;
File Number: RI-91-A-220 Northeast Nuclear Energy Company A*ITN: Mr. John F. Opeka Executive Vice President - Nuclear Engineering and Operations Group P.O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Opeka:
The U.S. Nuclear Regulatory Commission recently received information concerning activities at the Millstone Nuclear Power Facility, Unit 2. The details are enclosed for your review and follow-up.
We request that the results of your review and disposition of these matters 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. Please refer to file number
- filename
- when providing your response.
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 1. 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.
Your cooperation in this matter will be appreciated. We will gladly discuss any questions you have concerning this information.
Sincerely, ,
Ch s . Heh 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 h f
e%e ,
3 .
- v. .
Ietter to Northeast Nuclear Energy Company i
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bec:
Allegation File: RI-91-A 220 [
E. Kelly W. Raymond ,
T. Shedlosky t E. Conner '
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OCT 2 8 set
- Docket Number: 50-336 File Number: RI-91-A-0220
- Northeast Nuclear Energy Company A'ITN: Mr. John F. Opeka Executive Vice President - Nuclear P.O. Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Opeka:
The.U.S. Nuclear Regulatory Commission recently received information concerning activities at Millstone Unit 2. In a letter dated October 9,1991, we enclosed details regarding file number RI-91-A-0220 for your review and followup. We have added a new question to our request. Accordingly, enclosed is a revision of the enclosure to our October 9,1991 letter.
We request that the results of your review and disposition of these matters be submitted to Region I within 30 days of the date of mceipt 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.- Please refer to file number RI-91-A-220 when providing your response.
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.
Your cooperation in this matter will be appreciated. We will gladly discuss any questions you have concerning this information.
Si I k e=m ST: ,
A
/
C s W. Hehl, tor Division of Reactor Projects
Enclosure:
10 CFR 2.790(a) Information Issues and Requests cc w/o encl: '
I Public Document Room (PDR) f Local Public Document Room (LPDR)
State of Connecticut W'OK ! Sf? A
2 Northeast Nuclear Energy C.mpany 2 g g gg -
- bec w/ encl:
Allegation File, RI-91-A-0220 E. Conner files E. Kelly (Section Chief)
W. Raymond T. Shediosky l' Contractor's office files (REAGAN)
RI:DRP
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CO CurrenceS:
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