ML20212L067
| ML20212L067 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/18/1986 |
| From: | Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20212L002 | List: |
| References | |
| NUDOCS 8701290309 | |
| Download: ML20212L067 (1) | |
Text
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ATTACHMENT 2
[ 'o UNITE 3 STATES
, ! o NUCLEAR REEULATORY COMMISSION
{ ,e REGION I o, ., 631 PARK AVENUE MING OF PHUS$1A, PENNSYLVANIA 19406
% . . . . ,o#
18 AUG G86 Docket No. 50-220 File No. RI-86-A-0080 Alleger's name and address removed
Subject:
Allegation Against Niagara Mohawk Power Corporation Qi) refers to your conversations with myself, Richard Matakas and Glenn Meyer on July 11, 14, 22, 23, 28 and 31, 1986 in which you expressed concerns -e k related to instrument and control department practices at Nine Mile Point Unit 1.
Since you have also brought these concerns to the licensee's attention, we l
provided the licensee with a copy of the summary of allegations which you reviewed on July 31, 1986.
. As you know, we have initiated Actio.1s to examine the facts and circumstances I s of your coneerns and you will be informed of our findings. If you have any l ) comments or corrections to yourTuly 22, 1986 statement transcript which we i
1 provided you on July 28, 1986, please inform us as soon as possible. .
W,th respec& to the matter of yout* harassment by your supervisor, the United States' Department of Labor (DOL) is responsible for handling such complaints, i
as we informed you when we provided you with a copy of NRC Form 3 on July 31, '
1986. Therefore, we suggestgou contact DOL for resolution of your com laint.
A copy of 00L's " Procedures for Handling of Discrimination Complaints'* n er' Federal Employee Protection Statute:," is enclosed.
In closing, let me assure you that we will evaluate and conduct an appropriate follow-up on the issues you have brought to our attention. It is concerned individuals such as yourself, who, in bringing such matters to our attention, assist us in meeting our responsibilities for protecting the health and safety of the public. Should you have additional questions, or if I can be of further assistance in this matter, please do not hesitate to contact me.
Sincerely,
-e -
G701290309 070122 45 PDR ADOCK 05000220
/. C. Linville, 2
O PDR r., Chief cactor Project ection 2C
.- Division of Reactor Projects
\ Enclosure m - _
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8cMa %,o, UNITED STATES
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NUCLEAR REGULATORY COMMISSION
[e REClON i o, 631 PARK AVENUE 6
g ,o'g KING OF PRUSSIA, PENNSYLVANIA 19406 SEP 3 01986 Docket No. 50-220 50-410 Niagara Mohawk Power Corporation ATTN: Mr. C. V. Mangan Senior Vice President 300 Erie Boulevard West Syracuse, New York 13202 Gentlemen: '
Subject:
Combined Inspection Nos. 50-220/86-16; 50-410/86-46 A routine radiological controls inspection was conducted by Region I staff members on August 4-7, 1986 at the Nine Mile Point Nuclear Station, Units 1 and
- 2. The findings of the inspection were discussed with Mr. T. Perkins.
The inspection covered the status of previous findings at Units 1 and 2 and preoperational testing at Unit 2. We also examined the circumstances and cor-ective actions surrounding: (1) the contaminated wound. sustained by a worker at Unit 1 on April 1, 1986 and; (2) an off-scale dosimeter event on April 29, 1986 which also occurred at Unit 1. These incidents were brought to our attention by concerned workers.
Our findings associated with the contaminated wound are discussed in the enclosed inspection report.
Our review identified a number of apparent violations associated with the off-scale dosimeter incident. These violations are discussed in the enclosed inspection report. Because of the programmatic significance of these viola-tions, we wish to schedule an enforcement conference to discuss them with you.
The need for and nature of appropriate enforcement action will be the subject of seperate correspondence following our conference. We will contact you to set up a mutually agreeable date.
Your cooperation with us in this matter is appreciated.
Sincerely,
+t The as T. rtin, trector Olvision o Radiation Safety and Safeguards
Enclosure:
Combined NRC Region ! Inspection Report Nos. 50-220/86-16 and 50-410/86-46
'b(cOffrU5fbT5@~Q9,
Niagara Mohawk Power Corporation 2 cc w/ enc 1:
T. E. Lempges, Vice President, Nuclear Generation J. A. Perry, Vice President, Quality Assurance T. Perkins, General Superintendent, Nuclear Generation W. Hansen, Manager of Quality Assurance T. Roman, Station Superintendent J. Aldrich, Supervisor, Operations W. Drews, Technical Superintendent Troy B. Conner, Jr. Esquire John W. Keib, Esquire Director, Power Division Public Document Room (PDR)
Local Pubite Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of New York bec w/ enc 1:
Region I Docket Room (with concurrences)
Management Assistant, DRMA (w/o enc 1)
DRP Section Chief Robert J. Bores, DRSS T
O i .
U.S. NUCLEAR REGULATORY COMMIS$10N REGION I Report Nos. 50-220/86-16 50-410/86-46 4
i Docket Nos. 50-220
{ 50-410 License Nos. DPR-63 CPPR-112 Licensee: Niacara Mohawk Power Corporation 300 Erie Boulevard West Syracuse. New York 13202 l
Facility Name: Nine Mile point Units 1&2 i Inspection At: Oswego. New York i Inspection Conducted: August 4-7, 1986 1
Inspectors: 2( dt J
~
glh{ E L R. L. Nimitz, Senior Rabiation Specialist dat'e i at MM k. __ &la S. Sherbini, Radiation'6)ecFalist da t'e Approved by: ,s+1. 2h 9 //[Ify/
M. Shanbaky, ChTif, Faci itiaFs date
- Radiation Protection Section Inspection Summary
- Inspection conducted on August 4-7.1986 (Combined Inspection Report No.~5D-220/86-16: 50-410/86-46).
l Areas Inspected: Routine, announced inspection of licensee radiological ;
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controls including: licensee's action on previous findings at Unit 1 and Unit 2; organization, staffing, training, and qualification of personnel, preoperational testing at Unit 2; bulletins, circulars and information notices, and worker concerns.
1 Results: Several problems were identified in the implementation of radiological controls for LPRM replacement at Unit 1 (Details, paragraph 5.0).
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Details
- 1. Persons Contacted 1.1 Niagara Mohawk j
- T. Perkins, General Superintendent T. Roman, Station Superintendent, Unit 1
- E. Leach, Superintendent, Chemistry / Radiation Protection Management
- P. Volza, Supervisor, Radiological Support .
- J. Duell, Supervisor, Chemistry and Radiation Protection
- R. Gerbig, Radiation Protection Supervisor, Unit 1
- T. Irving, ALARA Coordinator 1.2 U.S. Nuclear Regulatory Commission W. Cook, Senter Resident Inspector C. Marschall, Resident Inspector
- Denotes those individuals who attended the exit meeting on August 7, 1986. The inspector also contacted other licensee personnel.
- 2. Purpose The purpose of this routine, announced inspection was to review the following matters:
Unit 1 licensee's action on previous findings; worker's allegation relating to receipt of unnecessary exposure sustained via residual radioactivity contained in a laceration; circumstances and Itcensee evaluation associated with an instance of off-scale dosimetry on April 29, 1986.
Unit 2 licensee's action on previous findings; licensee's action on bullettes, circulars and information notices; preoperational testing of the following systems:
Control Room Vent 11atten System, Standby Gas Treatment System. -
Process and Area Radiation Monitoring System.
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3.0 Licensee's Action on Previous Findings 3.1 .(Closed) Follow-up Item (50-220/83-27-01) Licensee to establish and implement the long term Radiation Protection Technician Training and ;
Retraining Program. This and other matters associated with Train- l ing/ retraining matters is being reviewed in conjunction with fol-
- low-up items 50-220/86-16-01; 50-410/86-46-01 and is closed for administrative purposes.
3.2 (Closed) Follow up Item (50-220/84-22-03) Licensee's on-site radia-tion sources do not possess sufficient activity to calibrate high range radiation monitoring equipment. The licensee calibrates in-struments on-site to about 320 R/hr. Administrative controls have been placed on use of these instruments in excess of this dose rate.
The licensee now sends selected instruments (RO-7s) off-site for periodic calibration to 20,000 R/hr. This matter is closed.
3.3 (Closed) Follow-up Item (50-220/85-18-01) Licensee to establish and implement a program to train personnel in new procedures and proce-dure changes and establish a program as necessary to verify previous I
experience of contract radiological controls materials personnel.
This matter is being reviewed in conjunction with follow-up item 50-220/86-14-01; 50-410/86-46-01 and is closed for administrative purposes.
3.4 (Closed) Follow-up Item (50-410/85-32-04) Licensee to complete pre-operational and surveillance testing of area radiation monitors. The licensee has completed preoperational and surveillance testing of ARMS. Appropriate radiation sources were used to calibrate the detectors and test alarm function and trip values. This matter is closed.
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3.5 (0 pen) Follow-up Item (50-220/86-14-01; 50-410/86-46-01) Licensee to establish and implement a training and retraining program for Radia-tion Protection personnel. The following was identified:
A defined initial training program for Radiation Protection
, personnel has been established and implemented. Tasks for which an individual is to be qualified on based upon his/her scope of l responsibilities, have been identified, i i
t A program to train personnel in new procedures and procedure !
changes is in place. i
- h A sufficient number of personnel have been trained and qualified ;
to support Unit 2 fuel load activities.
The scope and frequency of retraining /requalification for l Radiation Protection personnel has been identified. ,
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4 Within the scope of the review, the following matters needing lic-ensee attention were identified:
Upgrade retraining /requalification program for Radiation Pro-tection personnel and associated procedures as described in the recently developed retraining /requalification action plan.
Establish administrative controls for the training / qualification of " temporary Radiation Protection workers." Current procedures do not address " temporary licensee radiation protection per-sonnel" (Note: These individuals are licensee employees and not contractors).
3.6 (Closed) Follow-up Item (50-410/85-47-01) Licensee to review and revise General Employee Training (GET). The licensee reviewed the GET Program for applicability to Unit 2. The program was revised to identify new emergency assembly areas. Appropriate personnel have been made aware of the new locations and will be retrained in these during their yearly GET requalification.
3.7 (Closed) Follow-up Item (50-410/85-20-07) Licensee to establish and implement a Radiation Protection and Radwaste Operator Training Pro r gram at Unit 2. The Radwaste Operator Training Program was reviewed during inspection 50-410/86-17 and found acceptable. The Radiation Protection personnel training program is discussed in item 3.5 above.
3.8 (0 pen) Follow-up Item (50-410/85-20-09) NRC to review Radiological Controls Organization for Unit 2. The Radiation Protection elements of the organization were reviewed and found acceptable to support fuel load. The chemistry organization and staf fing was reviewed in inspection 50-410/86-17. The Itcens6e was found to have trained an adequate number of radwaste operations personnel to support fuel load activittos. The adequacy of the Radwaste Organization including descriptions in appropriate administration procedures remains open.
3.9 (Closed) Follow-up Item (50-410/85-32-01) NRC to review selected aspects of training and qualification program for Unit 2 Radiation Protection personnel. The matters associated with this item are discussed in section 3.5 above.
- 4. Worker Allegation (Unit 1) (RI-86-A-0074) 4.1 General On June 16, 1986, a contractor worker, formerly employed at the Niagara Mohawk Power Station, Unit 1, contacted NRC Region I. The worker alleged that radioactive material, introduced into the tissue of his finger via a cut, still remained in the tissue and should have i
been removed. The worker was concerned that the remaining radio-activity was providing needless exposure to him.
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5 The worker was contacted by an NRC Region I Senior Radiation Specialist on June 16, 1986.
The licensee was contacted vis telephone on June 16, 1986 by a NRC Region I Senior Radiation Specialist to ascertain the general cir-cumstances, and licensee evaluation surrounding the event.
A letter dated June 27, 1986 detailing the worker's concerns and requesting review of the concerns was transmitted to the licensee from NRC Region I.
An on-site inspection, by a NRC Region I Senior Radiation Specialist, of the worker's concerns was initiated on August 4, 1986.
4.2 Description On April 1, 1986, while preparing a pipe for welding in the drywell at Nine Mile Poi..t Unit 1, a contractor worker sustained a small cut (6 mm) to the fif th (small) finger of his left hand. A radiation survey of the cut showed that it was contaminated, measuring 2000 cpm at h" using a pancake probe. An unusual event was declarea at the site (contaminated injured person), and the worker was transported to Oswego Hospital for treatment, arriving there at approximately 1730.
A radiation protection supervisor was pre:ent at the hospital to supervise radiation surveys and to advise the physician on radiolo-gical matters. After several trials, the physician succeeded in reducing the wound contamination down to 1000 cpm.
Based on an initial low dose estimate by the radiation protection supervisor, (50 mrem to the finger), the physician and the radiation protection supervisor agreed that it is not necessary to perform any -
additional tissue removal or decontamination.
Subsequent dose calculations by the licensee, using acceptable in-dustry methods (ICRP 2 cnd NCRP 39), indicated 8.4 rem per calendar quarter as compared to tne regulatory limit (10 CFR 20.101) of 18.75 rem per calendar quarter.
On July 18, 1986, the worker met with the licensee and a licensee's consultant (physician). In that meeting it was recommended that surgical removal of the remaining minute activity was not advisable because of potential complications. The health implications from the radiation dose to the finger vero discussed and characterized as of minor and remote ef fects. The worker was satisfied with this discus-sion and the eatter was considered closed.
The licensee is in the process of revising his procedures to provide for a clear guidance on dose calculations, wound surveys and survey documentation.
6 5.0 Off-Scale Oosimeter Allegation (RI-86-A-0080) 5.1 General On July 11, 1986 a Itcensee worker contacted NRC Region I to discuss concerns associated with under vessel work performed at Unit 1 in late April 1986. The work was associated with incore instrumentation connectors (LPRMs) located under the vessel. In subsequent discus-sions, the worker mentioned an off-scale dosimeter incident that had occurred under the vessel.
An on-site inspection of the off-scale dosimeter incident was ini-tiated by a NRC Region ! Senior Radiation Specialist on August 4, 1986. The incident was determined to have occurred on or about April 29, 1986.
5.2 Description (Off-Scale Oosimeter. Unit 1)
As a result of a need to perform work associated with LPRM connectors under the reactor vessel, itcensee !&C personnel met with ALARA per-sonnel in late February and early March of 1986 to discuss the work and any associated radiological concerns. The first radiatica work permit (RWP) was initiated at that time but not used. A pre-job ALARA review was issued for this work on March 7, 1986. A new RWP was issued on April 5, 1986 (RWP No. 2043). The RWP required work party personnel to: report to the radiation protection techni- ,
clan at the drywell entrance and review survey maps prior to entry into the drywell; wear dosimetry on the head; and effectively organ-ize tasks to reduce exposure.
A radiation survey performed at 9:20 a.m. on April 5, 1986 indicated dose rates of 120-150 mR/hr at the waist level, 300-450 mR/hr at the head level and 600-800 mR/hr at the flange area.
On April 20, 1986 at about 2:00 p.m. a second survey was performed under the vessel in the work area. The documented survey indicated 100 mR/hr to the ankle, 125 mR/hr to the waist and 150-200 mR/hr at the head level.
At about 3:30 p.m. on April 20, 1986 the affected 1&C technician made an initial entry on RWP No. 2043. The individual was signed in on the RWP for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and received an exposure of 45 millirem. The technician received a low exposure because most of the time he was ,,
testing cables outside the drywell.
During the initial entry on April 20, 1986, the !&C technician read his dosimetry only once. He verified that his rate of accumulation of esposure was comparable with the dose rate information provided to him by the health physics control point technician. During subse-quent work entries the !&C technician stated that he did not check
7 L his dosimeter reading while he worked under the reactor vessel. Hs only checked it when he was signing out at the end of his shift.
At about 3:30 p.m. on April 26, 1986, the technician signed in or RWP No. 2043. The individual worked for a total of 5.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> and received 190 mR. The individual made multiple entries and signed out ;
l at 11:15 p.m. on April 26, 1986. !
' At at'out 3:00 p.m. on April 27, 1986 the I&C technician again signed in on RWP No. 2043, remained signed in on the RWP for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and received 160 mR.
A radiation survey of the LPRM area under the vessel was performed !
l and documented shortly before 3:10 p.m. on April 28, 1986. The survey showed 60-110 mR/hr at about 2 feet above the work platform and 140-300 mR/hr at the head level. ,
On April 28, 1986, at 3:30 p.m., the I&C technician again signed in on RWP No. 2043 and was signed in for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The technician '
entered the drywell without re-zeroing his dosimeter. The dosimeter indicated about 300 mR. However, upon exiting the drywell, he found that his dosimeter was off-scale. The individual apparently did not inform the health physics personnel that his dosimeter was off-scale.
t He signed out "off-scale".
) i l Due to an apparent breakdown in communications or a direct failure to I inform dosimetry personnel that this pocket dostmeter had gone off-scale, the individual's TLD was not read at that time. The indivi- !
dual was issued a second TLO. The individual was not asked why he '
wanted a replacement TLO. No evaluation of the exposure causing the ;
pocket dosimeter to go off-scale or subsequent personnel exposure i l sustained was made at the time. (Note Radiological Control Super- ;
visory personnel were unaware of the off-scale dosimeter.) '
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' The individual signed in again that evening on RWP No. 2043 at 9:00
- p.m. on April 28, 1986 and worked until about 11:30 p.m. He signed out with 250 millirem. No dose evaluation was made prior to his I i entry.
The individual turned in his extremity dosimetry and his second TLD badge at the end of his work shift that night. He did not turn in l his film badge, t
l (Note: At Nine Mile Point, personnel are routinely issued a film l badge along with an accompanying TLD badge. The TLD badge is read ,
! out under special circumstances (e.g. dose extensions) and is used to -
l verify pocket dosimeter readinfis. "he film badge is used as the dosimeter record and is normal'y processed the first and fifteenth of j eachmonth).
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8 A resurvey of the LPRM work area was performed at 3:25 p.m. on April 29, 1986. The survey results were essentially the same as the previous survey conducted on April 28, 1986 (i.e. 300 mR/hr maximum exposure to the head).
At about 1:30 p.m. on April 29, 1986, the I&C technician attempted to sign in on RWP No 2043 to enter the drywell and work on LPRMs.
However, radiation protection personnel would not let the technician enter because: 1) the computer generated exposure report had not been updated to reflect the technician's increase in quarterly ex-posure limit (1000 mR to 1500 mR) and 2) the control point was out of respirator face pieces. The control point received the exposure upgrade and additional face pieces. The technician entered the dry-well at about 2:30 p.m. (April 29,1986).
(Note: Based on subsequent licensee review, it appears that the individual, contrary to RWP requirements, did not obtain extremity dosimetry and entered and worked without the extremity dosimetry.)
The technician performed work on LpRMs in a kneeling / crouched position from 2:30 p.m. until about 4:30 p.m. (April 29,1986). At that time the technician left the area under the vessel and went outside the drywell, and near the access control point. The tech-nician left to clean the inner portion of his mask which was fogged.
The technician re-entered the drywell at about 4:40 p.m. (April 29, 1986). Due to fatigue, the technician stood up to work on one i particular LpRM and remained in this position until about 5:30 p.m.
In this position, the worker rested his head against a CR0 flange.
Unknowingly, the technician had ansted his head in an area with dose rates later measured to be between 500-1200 mR/hr. The area had not been be posted as such (e.g. with hot spot stickers). ,
(Note: The !&C technician had not been prohibited from working in l this manner. The technician did not request a radiation survey prior '
to standing up and placing his head on the CR0 flange. The techn-icianwasnotchallengedbythehealthphysicspersonnel.)
Thetechnicianwasinformedatthistime(5:30p.m.)viatheheadset that it was dinner time and work was halted. The technician exited the drywell, removed protective clothing and read his head pocket dosimeter. At this time he discovered his 500 mR pocket dostmeter to be off-scale. He informed the technicians at the access control oint of this instance. This individual signed out on the RWP as p'off-scaledosimeter"andwasdirectedtothedosimetryareatohave his TLD read.
The technician's TLD was read at 5:51 p.m. on April 29, 1986 and indicated 1266 mrem. This resulted in a total cumulative exposure of about 2050 mrem as compared to the regulatory limit of 3000 mrem per calendar quarter, i
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9 Upon learning of the incident at 6:30 p.m., the senior radiation protection technician overseeing drywell work, halted work on RWP No.2043. Additional radiation surveys were performed of this work area. The survey results indicated 300-800 mR/hr at I foot from the CR0 flanges and 500-1200 mR/hr on contact with the flanges.
A meeting was held on May 1, 1986 between the involved technician, l his supervisor and the radiation protection supervisor. A noncon-formance event transmittal (NET) was initiated.
On May 2 and 5, 1986, the involved technician briefed his co-workers regarding the incident and lessons learned.
5.3 Conclusions (Of f-scale desf rneter)
Evaluation of the information acquired during the inspection resulted in the following preliminary conclusions:
On April 28, 1986 at about 3:30 p.m., an !&C technician sus-tained an off-scale pocket dosimeter while working under the reactor vessel at Unit 1. Contrary to 10 CFR 20.201, no eval-uation was made of the potential exposure sustained by the individual. The individual re-entered the drywell and worked under the reactor vessel later that evening. The individual sustained an additional exposure that evening of 250 millirem.
Contrary to the established procedures (5-RRI-12) on April 28, 1986 at about 3:30 p.m.: 1) an individual's dosimeter went off-scale; and 2) the individual may have sustained a poten-tially high exposure; but the individuals film badge was not sent off for special reading. (5-RRI-12,Section4.0)
Contrary to established radiation protection procedures, on April 28, 1986 at about 11:30 p.m. an !&C technician did not turn-in his whole body film badge when he returned his extremity dosimetry. (5-RRI-3, Section 4.2)
Contrary to established radiation protection procedure, on April 29, 1986 at about 2:30 p.m. an !&C technician signed in on RWP No. 2043 and entered and worked under the RWP but did not obtain and use RWP required extremity dosimetry. (5-RP-2,Section7.7)
' On April 29, 1986 adequate radiation surveys, sufficient to ensure compliance with 10 CFR 20 dose limits were not made by an
!&C technician or requested to be made for him prior to his resting his head against a CR0 flange. As a result, the techn-ician unknowingly rested his head against a CR0 flange with measured dose rates of between 500 and 1200 mrem / hour.
10 l During the period April 26 - April 29, 1986 inadequate high 4
radiation area control was provided for LPRM work under the reactor vessel in that:
- 1) RWP No. 2043, which controlled LPRM work, did not contain a specific frequency for performing periodic radiation sur-t ve111ance of the LPRM work. The RWP indicated " intermit-tent" and intermittent was not defined.
- 2) A radiation monitoring device which continuously indicates the radiation dose rate in the area was not provided to the technician. The dose rate instrument present (a portable ;
Area Radiation Monitor)could not be used in areas entered by the technician (i.e. between CRD flanges).
- 3) An integrating, alarming dosimeter was not provided to the technician.
l During the period April 26 - April 29, 1986, an I&C technician made repeated entries under the reactor vessel to perform work i i
associate with LPRMs and was not adequately informed of the radiation levels emanating from CR0 flanges in his work area.
l The individual unknowingly rested his head against a CR0 flange, 4
the radiation emanating from the flange caused the dosimeter to
. go off-scale. Subsequent surveys of the CRD flanges indicated l contact dose rates of between 500-1200 mR/hr. (10CFR19.12) i Contrary to estabitshed radiation protection procedures, on April 28, 1986 an !&C technician failed to re-zero his dostmeter i prior to entering the Radiological Controlled Area to perform 1 RWP work. Anticipated exposure was such that an off-scale j reading could result. (Procedure S-RP-1, Section 5.6.4)
Contrary to established radiation protection procedures, per-sonnel did not initiate a Radiological Occurrence Report (ROR)
. on April 28, 1986 when an !&C technician's dosimeter went off-
! scale, due to the individual's failure to re-zero it prior to entering a high radiation area. Also, the technician did not 3
periodically check the dosimeter reading while working under l the vessel. (Procedure 5-RP-S,Section19.2.1) :
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} Contrary to established Radiation Protection Procedures, per-i sonnel did not issue a Radiological Incident Report on or about April 29, 1986 when it was determined that the !&C technician l
may have sustained excessive radiation exposure resulting fron deficiencies in his performance and/or the performance of the i
radiation protection group. The individual's dostmeter went off-scale during work on LPRMs under the reactor vessel.
(Procedure 5-RP-5, Section 19.2.1) l 1
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l The I&C technician was provided specific instruction regarding re-zeroing of his dosimeter (anytime the hair-line approaches 75% of full scale). The technician was tested on this matter and other matters in a May 28, 1985 Radiation Protection Exam.
The individual scored 87.5%. The individual answered the l
dosimeter re-zero question correct.
The general employee radiation protection training was not specific relative to the frequency of reading a self-reading pocket d)simeter. The applicable training module said that the
- " frequency is determined by the dose rate in the area." No l reading frequency was specified on the RWP used for LPRM work l
before or at the time of the off-scale dosimetry event (April 29,1986). It was added subsequent to the event.
, The licensee's high radiation area access control procedure l (S-RP-1) only provided for the use of a continuously indicating dose rate meter to meet technical specification high radiation area process controls (Technical Specification 6.12-la). The use of an alarming dosimeter or periodic surveys was not des-cribed in the procedure. Technical Specification options 6.12.1 l b and c respectively were not described. The licensee revised
, his procedure in March 1985 to allow use of these latter options l when " directed by supervision." Procedural description of implementation methodology was not described.
l 5.4 Corrective Actions (preliminary)
, Based on a September 4, 1986 telephone conversation between Messers.
l M. Shanbaky, R. Nimitz, and S. Sherbini of NRC Region I, and Messers.
l E. Leach, J. Duell, and P. Volza of Niagara Mohawk, the following actions were taken:
An evaluation of the off-scale dosimetry event was conducted by the Supervisor, Technical Support. The evaluation, dated August l
12, 1986 provided: a description of the event; some conclusions; and recommended corrective action.
A second evaluation of the off-scale dosimetry event was con-ducted by the Dosimetry Supervisor. The evaluation, dated j August 29, 1986, provided a description of irregularities associated with dosimetry turn-in, handling, and processing.
- A corporate health physics group investigation of the off-scale dosimetry incident and a list of recommendations, was sent to the Manager, Chemistry and Radiation Protection on August 27, i 1986.
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A meeting was held between corporate and site health physics personnel on August 28, 1986. A list of recommendations, action items, and responsible individuals was generated from the meet-ing.
Pending revision of applicable procedures, the licensee informed the dosimetry personnel in memoranda dated August 14 and 25, 1986 that no replacement dosimetry was to be issued to personnel without proper dose evaluations being performed (excluding normal badge change-outs). The memorandum also provided ins-truction as to what actions to take following notification of an off-scale dosimeter.
Pending procedure revisions to incorporate additional Technical Specification delineated high radiation area control options, a memorandum was issuedeto station personnel on August 27, 1986 to provide guidance relative to interface of workers with health physics personnel. The memorandum was issued to ensure workers and health physics personnel understand the work scope and location such that the current Technical Specification high radiation area control requirements contained in procedures are properly implemented.
6.0 Preoperational Testing (Unit 2) 6.1 General The inspector reviewed the status of the preoperational testing of the following systems:
Process and area radiation monitoring system (DRMS)
Safety related ventilation systems The review was with respect to criteria contained in the following:
Regulatory Guide 1.68, November 1978, "Preoperational and Initial Start-up Test Program for Water Cooled Power Reactors";
Final Safety Analysis Report, Section 6.5, " Habitability Systems";
Final Safety Analysis Report, Section 6.5, " Fission Product Removal and Control Systems";
Final Safety Analysis Report, (FSAR) Section 11.5, " Radiation Monitoring System";
FSAR, Section 12.3, " Radiation Protection Design Features";
13 Final Safety Analysis Report (FSAR), Chapter 14, " Initial Tests Program";
ANSI N42.18-1980, " Specification and Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity and Effluents"; and ANSI N13.1-1969, " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities".
The following matters were reviewed:
system status establishment of appropriate test procedures adninistrative control of testing and review of test results adequacy of testing The evaluation of the licensee's performance in the area was based on:
independent inspector observation of selected systems including walkdown of safety-related ventilation systems review of test results discussions with testing personnel.
6.2 Findings
6.2.1 Digital Radiation Monitoring System (DRMS)
This system consists of process and area radiation mont-tors.
Within the scope of the review the following was identi-fled:
Preoperational testing of all Technical Specification and fuel load required process monitors is not com-plete. These include: Above and Below Reactor >
Building Vent Monitors, Control Room Vent Monitor and Cooling Tower Blowdown. No testing deficiencies were identified.
All preoperational and surveillance testing of Tech-nical Specification required area radiation monitors is complete. Surveillance testing has been completed on all area radiation monitoring systems. No testing deficiencies were identified.
14 The licensee revised his FSAR to describe the actual location of some ARMS. The FSAR previously contained an incorrect location description for some monitors.
6.2.2 Safety Related Ventilation System Testing This system consists of the Control Room Emergency Venti-lation System and the Standby Gas Treatment (SBGT) System.
Within the scope of the review the following was identi-fled:
Preoperational Testing of the SBGT system, with the exception of in place leak testing, is complete. Two test exceptions remain open.
Preoperational Testing of the Control Room Emergency System with the exceptions of in place leak testing is complete.
Within the scope of the review, the following matters were brought to the licensee's attention:
System flow balancing is not complete Bolts were missing from covers of train A of the SBGT.
Also, material was found stored on top of train A.
Laboratory testing of a representative sample of charcoal from the systems was not performed. Con-sequently the systems do not meet operability require-ments of the Technical Specification. The licensee completed laboratory testing. The test results were l acceptable for the Control Room and SBGT systems.
In place testing results of by pass leakage of train B l
of the SBGT System did not meet Technical Specifi-cation requirements. The licensee performed sub-l sequent testing. Test results were acceptable.
7.0 Exit Meeting l The inspector met with licensee representatives, denoted in section 1 of i
the report, on August 7, 1986. The inspector discussed the purpose, l scope, and findings of the inspection.
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l
h 9 UNITED STATES oc. [,t*"8T0 j'g NUCLEAR REGULATORY COMMISSION l; e i nE:i:N I I 'f $31 PARK AVENUE Kino oP Prussia. PENNSYLVANIA 19406 Decencer IS, 1980 Docket No. 50-220 License No. DPR-63 Niagara Mohawk Power Corporation ATTN: Mr. B.G. Hooten Executive Director Nuclear Operations c/o Miss Catherine Seibert 300 Erie Boulevard West Syracuse, New York 13202 Gentlemen:
Subject:
Inspection Report No. 50-220/86-13 This refers to the special safety inspection conducted by Mr. G. Napuda of this office on Septe-ber 10-12 and 15-19,1986, at Nine Mile Point Unit 1, Oswegs, New York and your corporate offices in Syracuse, New York, of activities authorized by NRC License No. DPR-63 and to the discussions of our findings held by Mr. Napuda with Mr. T. Lempges and other members of your staff at the conclusion of the inspection.
Areas examined auring this inspection are described in the NRC Region I Inspection report which is enclosed with this letter. Withir. these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspector.
Based on the results of this inspection, we are concerned that the control for your operations, surveillance, maintenance and modification activities may be inadequate. Similar concerns were also identified in recent NRC Region I inspections (50-410/86-61, 50-220/86-07, 86-16 and 86-17) of both Units 1 and
- 2. The findings of the attached inspection report along with the findings of NRC inspection 50-410/86-61, 50-220/86-16 and 86-17 will be the subject of future enforcement conference. We will contact you, separately to set up a mutually agreeable date for this conference.
The need for and nature of appropriate enforcement action relative to the issues identified in the enclosed report will be considered after this conference .ind will be the subject of separate correspondence at a later time.
With regard to your actions for control of vendor technical information (discussed in Section 4.2 of the enclosed report), we believe that your current 1988-1989 schedule for completion of vendor technical manual validation and approval is not responsive to Generic Letter 83-28. You are therefore requested to reevaluate your planning in this regard and advise this office in writing of your actions to expedite the validation process.
cog n > M. r ni
%If_LLQ lGik g&
I Niagara Mohawk Power Corporation 2 Your cooperation with us in this matter is appreciated.
Sincerely VJJW
,afrr' Stewart D. Ebneter, Director Division of Reactor Safety
Enclosure:
NRC Region I Inspection Report No. 50-220/86-13 cc w/ encl:
T. E. Lempges, Vice President, Nuclear Generation J. A. Perry, Vice President, Quality Assurance T. Perkins, General Superintendent, Nuclear Generation W. Hansen, Manager of Quality Assurance T. Roman, Station Superintenaent J. Aldrich, Supervisor, Operations W. Drews, Technical Superintendent Troy B. Conner, Jr. Esquire John W. Keib, Esquire Director, Power Division Public Document Room (PDR)
Local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of New York o
/
/
o.
e U.S. NUCLEAR REGULATORY COMMISSION Region I Docket No. 50-220 License No. DPR-63 Category C Licensee: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility Name: Nine Mile Point Unit 1 Inspection At: Oswego and Syracuse, New York Inspection Conducted: September 10 - 19, 1986 Inspectors: ) /# /d' d' U. N4puda, Lead Reactor Engineer
/ dKe dfl D $
D. L. Taptiton, Sr. Technical Reviewer ' da'te b.YY '
d)G.HunterIII,ReactorEngineer lolI6l%
'date d k. A lo f /6 hfo D) L7 Harris, Jr. , pRC Consultant 'date
^9 0
()DgA. Beckman,' NRC Consultant l0
/ d4te Approved b : f4avD C Dr. P. K. Eapen/ Chief, Quality Assurance e
date z/. / 9 86 Section, OB, DRS Inspection Summary: Special announced inspection on September 10-19, 1986 (Report No. 50-220/86-13)
Areas Inspected: Licensee's actions to address the concerns identified in NRC Generic Letter 83-28 in the areas of Equipment Classification, Vendor Interface, Post Maintenance Testing, Plant Surveillances, and QA/QC Overview.
Results: Six violations were identified (iriadequate design / drawing control; failure to control measuring and test equipment; failure to control safety related maintenance activities; inadequate housekeeping; failure to implement surveillance procedures; inadequate inservice testing).
= T 7 s a 3 3 9 3 3 - 3 7y9
, DETAILS 1.0 Persons Contacted Niagara Mohawk Power Corporation (NMPC)
D. Balduzzi, Supt., Records Management W. Bandla, Asst. Supt. Operations
- J. Buckley, QC Supervisor R. Coon, Instrument and Control Supervisor K. Dahlberg, Maintenance Superintendent
- W. Drews, Technical Superintendent M. Falise, Supt., Mechanical Maintenance
- P. Francisco, Licensing Lead Engineer
- P. Frasier, Electrical Design Engineering
- F. Kawksley, Inservice Inspection Supt.
"W. James, Instrument and Control Supervisor
- T. Lempges, Vice President, Nuclear Generation R. Longo, Suprv, Mechanical Maintenance M. Masuicca, Asst. to Supervisor of Operations
- M. Mosier, NMPC Inspection Liaison
- R. Pasternak, Manager, Consulting, Nuclear Technology
- T. Perkins, General Superintendent "R. Randall, Supervisor, Technical Support
- T. Roman, Station Superintendent, NMP-1
- F. Stelter, Auditor
- K. Sweet, Supt., Electrical Maintenance United States Nuclear Reculatory Commission
- W. Cook, Senior Resident Inspector
- C. Marschall, Resident Inspector The inspectors also contacted other licensee administrative and technical personnel during the course of the inspection.
- Present at the September 19, 1986 Exit Meeting 2.0 Inspection Summary
2.1 Background
The reactor trip system, as part of the reactor protection system, is fundamental to reactor safety for all nuclear power reactor designs. Transient and accident analyses are predicated on the
.issumption that the reactor trip system will automatically initiate reactivity control systems on demand to assure that fuel design limits are not exceeded.
- 3 The design and regulatory philosophies for attaining the high reli-ability required of the reactor trip system have primarily relied on the use of redundancy, periodic testing, and quality assurance.
In February 1983 the Salem Nuclear Power Station experienced two failures of the reactor trip system on demand.
Regulatory and industry task forces were formed to determine the safety significance and generic implications of the events. Based on these findings, certain actions were required of all licensees.
These actions, transmitted in Generic Letter 83-28, fell into four areas.
(1) Post-Trip review, (2) equipment classification and vendor interface, (3) post-maintenance testing, and (4) reactor trip system reliability improvements.
NMPC submitted their response to Generic Letter 83-28 in letters listed in Attachment A. This inspection included the areas of equipment classification, vendor interface, post maintenance testing, and QA/QC interface.
2.2 Inspection Results Six violations with multiple examples were identified as detailed in paragraphs 3.2.a and 5.2.c. Two unresolved items were identified and are discussed in Paragraphs 4.2 and 5.2.c.
Except for the weaknesses detailed in the violations, and unresolved items, the licensee's actions were found to adequately address the concerns of the Generic Letter 83-28 and were consistent with their commitments provided in the letters listed in Attachment A.
3.0 Equipment Classification 3.1 Program Review The inspectors reviewed the applicable documents listed in Attach-ment A and determined that the Equipment Classification Program addressed the following:
Safety related components are identified as such in written procedures for procurement, maintenance, and modification.
Management oversight of source documentation for the designation of safety related systems and components.
i 1
. 4 Corporate level procedures for safety related component procurement, maintenance, and. modification.
Periodic Quality Assurance audits of activities impacting safety related equipment.
Corrective action program for safety related equipment.
3.2 Program Implementation The licensee uses the NMP-1 Q-List to identify safety related items at the " system" level. The Q-List specifies the boundaries of the safety related portions of each system by reference to the system drawings (P&ID, elementary diagram, etc.). Further definition is also provided by a computer based " component" level listing.
The Q-List was managed as a controlled document under the cognizance of the corporate QA Engineering staff.
A field comparison of system mechanical drawings and electrical elementary diagrams versus installed equipment was performed and certain components were selected for verification of the implement-ation of the licensee's Equipment Classification Program. For each component selected, the inspectors reviewed the adequacy of safety classification, procurement, receipt, storage and handling activi-ties as discussed in the following sections of this report.
- a. Reactor Protection System (RDS)
Recirculation Pump MG Sets A visual inspection of the Recirculation Pump Motor Generator Set field breakers confirmed correct component labeling.
Components were correctly classified in accordance with the component Q-List.
During the walkdown, all four breaker and control cubicles were found to contain loose nuts, bolts, relay covers, and terminal cover protectors (found on the floor inside the cubicles). Also, in three of the cubicles loose wires (equipped with terminal lugs) were observed lying on the floor.
Station Administrative Procedure 8.5, Housekeeping and Cleanliness Control, Appendix A, provided guidelines for maintaining acceptable fire protection and housekeeping
, conditions. The procedure required that " trash, rubbish, etc.
are not present in cable trays, in or on top of energized
- p. switchgear, instrument racks or control cabinets". The above Q represents a violation of the housekeeping and cleanliness
? requirements of AP-8.5 (50-220/86-13-01).
. 5
' Also, eight cables inside the cubicle could not be identified, due to the lack of cable tags.
The licensee's electrical supervisor responded to these findings by issuing work requests (WRs) to correct the above; the WRs (see Attachment A) were completed and the conditions corrected prior to the close of the inspection.
ATWS Control Panel On September 12, 1986 the inspector visually inspected the ATWS Instrument Panel No. 1S48. The licensee supplied drawing (C-34122-C/1, Revision 2) did not reflect the actual panel wiring configuration. The licensee was notified and their review of this condition established that the drawing had been updated from Revision 2 to Revision 3.
On Sept 15, 1986, the inspector was provided the Revision 3 drawing, reinspected the panel, and noted that the as instal-led wiring connections were still not in agreement with the drawings. The drawing revision 3 was verified by the inspector to be the most current available at the facility through the Document Control Group.
The inspector then requested the licensee to perform a com-plete panel wiring verification versus the latest available drawings. The licensee's verification effort confirmed and documented the nonconformities between the panel wiring and drawings on September 15, 1986. The licensee was again notified that the plant documents did not reflect the as-built condition of the panel.
Concurrent with the above verification effort, the Document Control Group found that a Revision 4 of the drawing had been entered in the document control computer system as a "pending revision" by NMPC (Syracuse) Engineering on the morning of September 15, 1986, which was after the inspector's walkdown.
The licensee explained that a plant modification had been done early in 1984 and that the drawing had not been updated.
The licensee was asked to provide the modification package or.
any documentation that would reflect the modification's as in-stalled configuration in the panel. No such information was available onsite to reflect the as-built condition of the panel.
On September 16, 1986 the Revision 4 drawing was delivered to the site by a NMPC (Syracuse Office) engineer and was verified by the inspector to reflect the as-built condition of the panel.
,, 6
Site. Administrative Procedure APN-9, Procedure for Station Permanent and Temporary Modifications and Replacements, Revision 3, (in effect at the time of the above modification), required, in Section 5.3.1 and 5.3.4, that drawings reflecting plant modifications be prepared and delivered to the. Station Superintendent in a timely manner. The above drawings were inadequately maintained in that revisions reflecting the installed plant modifications were not provided to the site until more than 24 months after completion of the modification.
' Failure to maintain drawings reflecting the as-built condition of the facility constitutes a viciation of administrative procedures for control of modifications. '(50-220/86-13-02)
Subsequent to this inspection, during the week of Septmber 22, 1986, the licensee notified the NRC Resident Inspectors that additional discrepancies had been identified in the ATWS Panel wiring (above). Apparently, the delay in issuing as-built drawings from the 1984 modifications had resulted in subsequent panel wiring modifications being performed with
. erroneous, unrevised drawings. As a result, an ATWS Panel "Lo.ss of Voltage" annunciator circuit was miswired, rendering it inoperable.
On a loss of voltage to the ATWS Panel, certain circuits, including the Recirculation Pump ATWS Trip, require manual reset action to rearm them. The annunciator is intended to alert the operators to perform the manual reset.
The inoperability of this annunciator could result in undetected inoperability of the ATWS Recirculation Pump Trip if the ATWS Panel circuits were deenergized and reenergized and the opera-tors failed to recognize that the condition required manual reset without the benefit of the annunciation. However, exist-ing restoration procedures and other operable annunciators appear to provide assurance that the manual reset would be accomplished. Other inspection efforts related to this matter were in progress and will be reported in inspection reports Nos.
50-220/86-17 and 86-18.
The NRC Resident Inspectors are conducting followup inspection in this regard; results of their inspection will be reported in NRC Region I Inspection Report 50-220/86-18.
The inspector also checked training records and verified that plant operators had been trained on the modification June 25, 1984.
REMOTE SHUTDOWN PANELS (RSP)
The inspector visually inspected the RSP Channel #12 and found i loose bolts, washers, screws, a box of fuses and retaining '
clips on the floor inside of the cubicle. Cables enter the top t
4
h 7 of the panol atd it was noted that the cable protector (bushing) on the top panel c.pening was partially removed. The licensee's electrical supervisor was informed of these conditions and took immediate action to correct the noted conditions by issuing WRs. The above is anotner example of poor housekeeping and part of violation (50-220/86-13-01).
Further drawing control discrepancies (see ATWS Control Panel above) were identified in that drawing C34813C, Sheet 6, Revision 1, showed field wiring connected to panel terminals R7, R8, R9, RIO, RII, 55, S6 and S7. However, physical inspec-tion determined that nothing was wired to the terminals. Drawing C34813C, Sheet 5, Revision 0, showed a shield wire for a drywell pressure instrument channel wired to terminal M4.
Physical inspection showed the wire actually landed on terminal M1. Also, drawing C34813C, Sheet 4, Revison 1, shows terminal A13 to be wired for an emergency condenser water level channel which did not exist in the panel; the field wire was actually terminated on point Bl.
The licensee representatives were unable to establish the reason for the discrepancies between the drawings and the as-installed wiring but stated that it appeared to be the result of a modification. The electrical supervisor issued a Drawing Charge Request to correct the drawings. This is an example of failure to control as-built drawing changes resulting from modification activities. (50-220/86-13-02)
The RSP channel #11 had the same type of drawing configuration control problems as channel #12 and is considered to be a further example of violation 50-220/86-13-02.
Also, during the inspection of Panel 11, the inspector found that three steel panel isolation barriers, which separate non-safety related equipment and wiring from safety related equipment and wiring, were removed and lying on the floor of the cubicle, apparently the result of in progress or impro-perly completed maintenance activities.
The licensee was unable to identify the reason for disassembly of the isolation barriers. No in progress activities were apparently authorized for the panel. The absence of the barriers is also a deviation from the electrical separation criteria requirements of the NMP-1 FSAR,Section IX 3.1.
Five SMB switch covers, and four relay covers were observed lying on the floor of the cubicle. Also a spare instrument was missing from the front of the panel. The electrical supervisor issued WRs for immediate corrective action.
i
8 Procedure AP-5.0, Procedure for Repairs, provides the require-ments for controlling and documenting maintenance activities on safety related systems, components, and structures and provides for restoration of systems components and structures upon completion of maintenance activities. Failure to control maintenance activities on safety related equipment in accordance with approved procedures constitutes a violation (50-220/86-13-03).
As a result of the discrepancies in as-built drawing accuracy, the sample of modification-affected drawings was expanded for field verification to determine whether the drawings accurately represented existing plant configuration and installed modifications. The expanded sample included:
Modification Drawing 83-08 C18017-0, Revision 26 82-80.5 C19859-C, Sh. 8&9, Revision 22 82-69 C19859-C, Sh. 2&5, Revision 23 The inspectors noted that the actual drawing revisions had not been made until 1984 for the 1982 modifications. No additional discrepancies were identified,
- b. 4160 V Safety Related Switchgear The under voltage and degraded voltage relays were reviewed for proper safety classification in the Q-List. The relays and timers were physically inspected in the presence of a plant electrician. All components were correctly labeled in accor-dance with plant documents. No violations were noted.
- 1. Mechanical Portions of the Feedwater/High Pressure Coolant Injection (HPCI), the Shutdown Cooling (SDC), Emergency Cooling (ECS), Containment Spray (CSS), and Control Rod Drive Hydraulic (CRDHS) systems were visually inspected and the following components were selected for verification of safety classification controls.
SOV 117/118 CRDHS Scram Pilot Valves CV126/1127 CRDHS Scram Valves SOV NC 16A Backup Scram Solenoid Valve SOV NC 168 Backup Scram Solenoid Valve 80-23/24 Containment Spray Pumps 80-35/16 Containment Spray Valves SOV 80-16C/D Containment Spray Solenoid Valves
o 9 80-33/34 Containment Spray Heat Exchanger 38-04A SDC Valve and Operator 38-01 SDC Valve and Operator FCV-30-10 SDC Valve 38-13 SDC Valve and Operator 39-10A ECS Valve and Operator 39-08A ECS Valve and Operator 39-06 ECF Valve and Operator NOTE: The above samples were also used for selection of work request and post maintenance testing review samples.
The inspectors reviewed MRs (listed in Attachment A) for proper safety classification, QC coverage, post mainten-ance test requirements, maintenance and test equipment used, spare parts used and appropriate supervisory reviews and approvals. A review of the Q-List determined that the above components were correctly classified in Maintenance Requests (MRs) and procedures (listed in Attachment A).
No violations were identified.
- d. Instruments and Controls Instrumentation and controls (I&C) associated with the Reactor Protection, Emergency Core Cooling, and Control Rod Drive Hydraulic systems were visually inspected and the following components selected for verification of safety classification.
PS 80-70/74 CSS Suction Alarm PT 80-69/75 CSS Pump Discharge Pressure 12K5 HPCI Initiation Logic Circuitry 11K5 HPCI Initiation Logic Circuitry 12K6 HPCI Initiation Logic Circuitry 11K6 HPCI Initiation Logic Circuitry NOTE: The above samples were also used for selection of work request and post maintenance testing review samples.
The field walk down verified that the above components identified were properly and clearly marked in the field and corresponded to the current revision of system diagrams. The licensee's Q-List properly designated both the systems and components reviewed.
. ~10 Work Requests ~(WR's) were examined for selected components of the RPS, HDCI, SDC, ECS, CSS, and CRDHS to verify proper safety classification of repairs and spare parts used. The WRs reviewed are listed in Attachment A. In all cases, safety classification of work and spare parts used was in agreement with the licensee's Q- List and supplementary indices.
No violations were identified.
- e. Purchasine and Storage The licensee's purchasing and storage activities were reviewed using the components selected. The purchase documents (listed in Attachment A) were complete and contained:
Proper safety classification Environmental requirements Seismic category National codes criteria (e.g. IEEE, ASME) 10 CFR 21 requirements Shelf life provisions requests Documentation (e.g. Certificates of Conformance, Test Reports) requirements Vendor technical information (e.g. manuals, storage
, conditions) requirements The onsite warehouse was toured. Cleanliness, storage environment, item identification, traceability, access restrictions, and shelf life controls were adequate ~for a sanple of items similar to or associated with reviewed Purchase Orders (P0s).
The licensee has engaged an agent to evaluate those items in stock subject to age deterioration and to establish a program for shelf life controls. The agent and licensee warehouse personnel have primary roles in this effort that include the following major program aspects:
contact of component vendors, composition manufacturers and the NSSS for shelf life information use of published data such as Military Standards l
l i
11 consideration of components containing internal parts subject to shelf life considerations marking of components and spare parts as to shelf life expiration dates and their inclusion in a listing removal of expired stored components and spare parts to a hold area until final determination of action is made continuation of this effort for items that were missed and any newly purchased items lacking shelf life information a requirement in purchase documents for vendors to supply shelf life informaiton No violations were identified, nowever the following concern is being followed by the Resident Inspectors and will be addressed further in Inspection Report 50-220/86-18.
The recent failure of a Hydraulic Control Unit Scram Outlet Valve Actuator diaphragm, Stock Number 95-26-028, was attributed by the licensee to aging deterioration. The shelf life provi-sions, purchase, length of service, and status of changeouts for these diaphragms were reviewed with the following results:
item composition is Buna-N and nylon '
stocked spares were purchased in 1975 (Purchase Order 87787) the item was not evaluated during the initial shelf life review an evaluation by the NSSS estimated shelf life to be seven years while the licensee's shelf life program age determination is five years the licensee had removed stocked items to a hold area pending determination of final action approximately half of the installed scram valves, purchased prior to 1970, still contain original diaphragms no service life expectancy determination had been made for this or other items subject to aging deterioration while in service
12 The licensee had not yet decided on a changeout program for these valves nor considered what other valves fall into this category.
4.0 Vendor Interface 4.1 Program Review The Vendor Technical Manual Upgrade Project Control Plan, Revision 0, details the specific actions, responsibilities, interfaces and sequential steps to implement the vendor information control portion of the Nuclear Utility Task Action Committee (NUTAC) proposal for a Vendor Equipment Technical Information Program (VETIP).
NMPC has contracted an agent to assist in the identification, evaluation, vendor contact, validation, approval, and issuance of accurate and controlled vendor manuals. The agent and Maintenance and Engineering Departments have primary roles in this project.
The following are the major aspects of this effort.
Manual inventory .
Reviews and comparison of manual and maintenance procedures and resolution of comments Determination of need for licensing and environmental qualification reviews Field verification of components in plant and determination of appropriate action if differences are found Tracking status of above, assignment of identifiers to validated manuals, and provision of program controls for 1
future updating No violations were identified.
4.2 Program Implementation Approximately 1600 vendor manuals were identified; over 500 vendor contacts were completed; plant procedures for almost 200 manuals were reviewed; and maintenance comments resolved for over 90.
However, engineering comments, master copy markups and final approval have not been completed for any manual. The projected completion date for the agent's activities is mid-1988 and the licensee review and final approval of all manuals is targeted for mid-1989.
13
' The licensee was informed that the status and anticipated completion dates of the project were not as advanced as anticipated by NRC, based on NRC's understanding of the NMPC responses to Generic Letter 83-28. The licensee was advised that the accepta-bility of the above would be unresolved pending NRC:RI review.
(50-220/86-13-04)
The following were reviewed with respect to vendor supplied manuals and supplementary information,
- a. Reactor Protection System The inspector reviewed the General Electric vendor manual on AK2-25 type breakers used on the Recirculation Pump MG Sets (part of the ATWS pump trip circuit). Currently NMP-1 does not interface vendor technical letters or service advisories with vendor manuals. However the SIL's and SAL's are reviewed by the plant staff for applicability to station procedures and activities.
The technical information supplied by GE suggests the AK breakers being used as Reactor Trip Switchgear breakers should be considered safety related. This licensee's breakers are non-safety related.
The licensee has not incorporated the suggested 20 ounce-inch force adjustment on the tripper bar of the breaker and uses the maximum force of 24 ounce-inches. Further advice from GE for AK breaker reliablity suggests the update of the front frame assemblies with the Mobil 28 lubricated bearings. Again the licensee considers the advisory to be applicable for Reactor Trip breakers only.
- b. Mechanical Systems The inspectors reviewed preventive and corrective maintenance procedures, surveillance tests, and post maintenance test instructions listed in Attachment A for technical accuracy and appropriate references to, and incorporation of vendor technical information. The inspectors determined that the documents reviewed appropriately incorporated such information.
The inspectors also reviewed the licensee's preventive maintenance program general implementation status per Administrative Procedure 8.1, Preventive Maintenance, through review of preventive maintenance schedules and completion status, interviews of key personnel, and review of completed maintenance records listed in Attachment A.
NSSS Vendor Recommendations The licensee is currently addressing NSSS vendor (GE) recommend-ations. The licensee has initiated action for GE's Rapid Infor-mation Communication Services Information Letter (RICSIL) 008 and
14
. Services Information Letter (SIL) 441 concerning Control Rod Drive Scram Anomalies. RICSIL 008 provided the initial notification of problems with scram solenoid valves repair kits.
SIL 441 updated the information in RICSIL 008 and advised that scram solenoid repair kits shipped prior to July 17, 1986 could be defective and should be returned to GE for reinspection. GE also advised that facilities with such kits already installed should continue to use Technical Specification (TS) surveillance testing to demonstrate Control Rod Drive Hydraulic System (CRDHS) operabil-ity. SIL 441 provided additional instructions for inspecting solenoid valve core springs and core assemblies to assure proper operation.
The licensee is in the process of gathering the affected repair kits for return to GE. Onsite inspections per SIL 441 will be performed on those kits retained as spares until the reinspected kits are returned from GE. The licensee informed the inspector that one half (previously one third) of the scram pilot valves would be rebuilt during each outage.
- c. Instrumentation and Controls Instrument calibration procedures and data listed in Attachment A were reviewed to determine adequate incorporation of vendor inform-ation. The review included a sample of data from recently complet-ed functional tests and calibrations.
No Violations were identified 5.0 Post Maintenance and Surveillance Testing 5.1 Program Review The inspectors reviewed the procedures in Attachment A to determine that the following were addressed during the performance of maintenance:
Written procedures provide for initiating requests for post maintenance testing.
Criteria and responsibilities for review and approval of maintenance.
Criteria and responsibilities for the basis of safety related/
non-safety related activity designated.
Criteria and responsibilities for inspection of post mainten-ance testing.
- 85 Methods for performing functional testing following mainten-ance.
Administrative controls for documentation of maintenance activities.
No violations were identified.
5.2 Program Implementation Post maintenance testing is initiated and tracked in accordance with Administrative Procedures AP-5.0, Procedure for Repair, and TDp-8, Post Maintenance Testing Criteria. The inspectors reviewed maintenance activities listed in Appendix A and compared the documentation of those activities with the requirements of the above procedures, the facility Technical Specifications, and avail-able vendor supplied information.
Except for the violations noted below, the licensee's program for assuring adequate post maintenance testing and surveillance testing appeared to be functioning acceptably. The existing practice
' includes retaining copies of the detailed records of such testing with the maintenance records, providing definitive confirmation of testing adequacy.
- a. Reactor protection System The inspector deterrined that the licensee had identified a reliability problem with the MG set voltage regulators and had initiated a plant modification to improve reliability. The inspector reviewed the plant modification package and discussed the licensee's plans with the Technical Services supervisor.
The modification package did not appear to include sufficient information to complete the modification. Examples included recommendations of testing without specifying what type of test to perform. The inspector was informed that the modification is scheduled for the next refueling outage which would give the licensee enough time to complete the modification package for final approval to work.
- b. 4160 V Safety Related Switchgear The inspector determined that the surveillance test procedure with respect to Technical Specification requirements for under voltage and degraded voltage relays was adequately written and the performance of the test showed the set points were checked and verified within the limits of the technical specifications.
16
- c. Mechanical Systems The inspector reviewed WRs listed in Attachment A for adequate post maintenance testing. Surveillance test procedures and completed surveillance tests performed as post maintenance tests, listed in Attachment A, were also reviewed for technical adequacy, completeness, acceptance criteria, maintenance and test equipment tracking, corrective action and appropriate supervisory review and approval.
During the review of the WRs, the inspector noted that complete listings or descriptions of post maintanence testing were not always provided on the WR cover sheet. In some cases, only one of several applicable procedures would be referenced. In other cases, no specific procedure was referenced. These discrepancies involved WRs issued prior to and during the 1986 refueling outage. The inspector was able to verify document-ation of adequate testing for all cases except as noted below.
Backup Scram Valve Testing In a July 31, 1984 letter to NRC in response to GL 83-28, Item 4.5.1, NMPC provided a justification for not performing on line functional testing of backup scram valves. In a letter dated December 31, 1984, as a result of their evaluation, the licensee advised that they would perform a qualitative test of the backup scram valves during each refueling outage.
The test involves interrupting power to the valves and monitoring the system upstream header pressure to ensure that the valves function properly. The inspector reviewed N1-ISFC-44.28, Qualitative Test of the Backuup Scram Valves, including data from the first (4/13/86) performance and found it accep-table except for a typographical error in the identification of Annunciator Window F-3-3-2. A temporary procedure changes was issued to correct the error.
The inspector also noted that the licensee uses N1-IMP-44.2, CRD Scram Valves Timing, to insure the proper operation of the scram valves after maintenance on either the scram pilot valve solenoids or the scram valve. The licensee informed the inspector that as a preventative maintenance measure, one half of the pilot and backup scram solenoid valves are repaired or replaced during each refueling outage. The inspector reviewed various WRs listed in Attachment A for performance results of the valve timing after maintenance and no discrepancies were identified.
. 17 Control cf Measurement and Test Equipment (M&TE)
The inspector reviewed calibration and usage information for a sample of M&TE used on safety related components by the I&C and Maintenance Departments.
The Maintenance Department appeared to be controlling M&TE in
-compliance with the site administrative procedures.
Administrative Procedure AP-8.4, Procedure for Control and Calibration of Equipment Used in Tests and Inspections, requires that the calibration records for such equipment be mair.tained at the facility. The I&C Department was unable to provide the inspector with a calibration data sheet for a Transmation Minitemp Calibrator, Serial Number 18497, previ-ously used on safety related equipment. The licensee stated that the calibration had been performed off site and had the data sheet, dated April 16, 1986, forwarded to the site after the inspector's request.
Failure to maintain the above calibration record onsite in accordance with AP-8.4, Section 6.3.1, is an example of failure to follow site administrative procedures (50-220/86-l 13-05).
AP-8.4, Section 6.3.2, also requires that, between M&TE calibrations, all plant devices calibrated or tested by a piece cf M&TE must be recorded in a usage log Gould Recorder, Serial Number 1155, was used for at least thirteen performanc-es of scram valve timing tests on November 4, 1985. The inspector reviewed the calibration / usage log for the Gould Recorder and determined that the above usage for scram valve testing was not logged. This constitutes a further example of failure to follow AP-8.4 (50-220/86-13-05).
While examining the above recorder, the inspector noted that the individual recorder channel modules (RMS, DC, frequency, etc.) appeared to be individually calibrated and each bore an individual calibration sticker and calibration due date.
I&C personnel informed the inspector that the modules were returned to the vendor for calibration with the recorder but i
that the modules were also interchanged with other recorders when needed to support tests requiring different recorder configurations. Individual module usage is not recorded on either plant procedure data sheets nor the calibration / usage log. The licensee was unable to substantiate prior usage of individual recorder modules for safety related activities.
1 J
1 e- ,n. , -- , - - - . , , , -
~-,,-a >--------------,-,,-r-m- - -- --,-,-----n--r n v-- ------ --~
18 Failure to maintain the calibration / usage log for each indi-vidually calibrated module is another example of failure to follow site adminsitrative procedure AP-8.4 (Section 6.3.2).
(50-220/86-13-05)
Pump and Valve Inservice Testing Periodic and post maintenance testing of pumps and valves is administered, in part, by the licensee's Pump and Valve Inservice Test (IST) Program. The inspectors reviewed imple-mentation of this program's requirements for selected pumps and valves in the Shutdown Cooling, Emergency Cooling, and HPCI System.
The licensee's IST Program, Revision 4, for the first 120 month inspection interval (since initial plant operation),
concluded at the end of the 1986 refueling outage on June 20, 1986. This progr:< was based upon Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition, Summer 1975 Addenda and is required by Technical Specif.ication 4.2.6.
The IST Program, Revision 7, for the second 120 month interval (beginning June 20, 1986) is based upon the 1983 edition, Summer 1983 Addenda, of ASME Section XI.
Both Revisions 4 and 7 of the IST Program had been submitted to NRC:NRR for approval; approval had not been granted at the time of this inspection. The licensee had implemented the programs pending approval. Further discussion of licensee implementation and administration of the programs is provided below.
Emergency Coolino System Valves Emergency Condenser Condensate Return Valve 39-06 (and its sister valve 39-01 in the #11 Train) are air-to-close, spring-to-open, normally closed isolation valves in the reactor con-densate return line from the ECS isolation condensers.
TS 3.7.2 lists the valves as Reactor Coolant Isolation valves
- that are required to open on an ECCS initiation signal and to close automatically to isolate the reactor on indication of an ECS downstream piping rupture.
The IST Program, Revision 4, Pump and Valve List Section, Page A-20, lists valves 39-01 and -06 as ASME XI Category A valves.
Category A valves are those for which seat leakage is limited to a specified maximum amount in the closed position for fulfill-ment of their function .
4
-- -------,-,>n-- ~ . - . , , , , , - , , , ,, - - - , - -
. . - _ = _ - - . . _ - _ - - - - - - - . .
- 19 ASME Section XI, Article IWV 3420 requires seat leakage testing of Category A valves at each refueling outage but not less than once every two years. Consistent with this requirement, IST_ '
Program, Revision 4, specifies the performance of periodic leak i rate testing. This categorization is consistent with the valves' Reactor Coolant isolation valve functions. However, IST Program, Revision 4, Table C1, Test Schedule, Page 8, lists the valves as Category B (valves for which seat leakage is inconsequential for fufilliiment of their function). This categorization appears erroneous.
4 The licensee was unable to demonstrate that individual seat leakage tests had been performed on either valve. The licensee demonstrated that the valves had been tested as boundary valves during a 5/29/84 containment integrated leak rate test per 10 CFR 50, Appendix J and stipulated that this latter test met the testing requirements for the valve. Further, the licensee maintained that the normally shut valves are exposed to reactor I. pressure during normal operations and that any leakage under those conditions is readily apparent. The inspector noted, however, that under the above conditions, the valves are not necessarily exposed to rated pressure across the valve seats.
Further, the second 120 month IST Program, Revision 7, includes
' both a valve seat leakage test requirement and a relief request to preclude testing valves 39-01 and 39-06. The relief request is based upon a 10CFR50 Appendix J, TS Change Request (dated August 27,1954) currently pending with NRC to exclude these valves from Appendix J, Type C local leak rate testing, j The inspector stated to the licensee that: (1) the above exemp-4 tion request had not yet been approved by NRC and therefore could not be implemented without formal written relief from NRC; (2) the Appendix J test requirements are exclusive of and are derived from technical bases that are different from the
, requirements of ASME XI; and (3) under the requirements of ASME
- XI, the valves must be considered as reactor coolant isolation valves requiring demonstration of seat integrity. ,
Failure to periodically test valves 39-01 and 39-06 constitutes a violation of TS 4.2.6 and the IST Program requirements (50-220/
l 86-13-06).
4 Shutdown Cooling System Valves Shutdown Cooling loop isolation valves 38-01, 38-02, 38-12, and 38-13 are also listed as Reactor Coolant Isolation Valves in T3 3.2.7. ,
4 1
i
- - , ~ ,a,--,- , , ,, , , . , - - ...,----,--m,,-- n,-- , , - - . , , - - - , , ,- ,,,--,._,.-,e,,,,,-n ,,..,,-,,,.r,----,,----r,,,,--,--c-
20
, The valves are identified as Category A valves in the IST Program Revision 4, Field Change 8. Prior to Field Change 6 the valves were subject to seat leakage testing.
The IST Program specified that they be tested per 10 CFR 50, Appendix J as they were then considered to perform the dual role of both containment and reactor coolant isolation valves. (NRC will, on a case basis, accept Appendix J, Type C, testing as an equivalent to the leak testing of ASME XI).
However, on May 31, 1985, the licensee issued Field Change 6 to the IST Program, Revision 4, deleting the leak rate test requirements based upon the submitted but unapproved August 27, 1984 10 CFR 50, Appendix J, TS Change Request to NRC.
Based on the above Field Change, no seat leakage testing required by ASME XI has been performed on the valves correspon-ding to either of the valve's dual roles as containment and reactor coolant isolation valves. Deletion of the seat leakage l
testing requirerrents and failure to perform the seat leakage testing on the subject valves consitutes a second example of the violation of TS 4.2.6 (50-220/86-13-05).
Further, valve 38-12 is a swing check valve. Both Revisions 4 and 7 of the IST program provide for valve exercise testing in only the forward flow direction. In its role as a reactor coolant isolation valve, 38-12 is required to seat in the
, reverse flow direction consistent with ASME XI, Article IWV3410 l (1974/75 Code) and Article IWV 3412 (1983 Code).
The licensee conducts forward flow check valve exercise testing of 38-12 by verifying system flow during normal system cperation in accordance with Operating Procedure OP-43. However, no provi-sions for reverse flow exercise testing the valve have been established or implemented. This is a third example of the violation of TS 4.2.6(50-220/86-13-06).
Overall Program Implementation As noted above, NRC:NRR approval for the IST Program has not been granted. By NPMC letter to NRC (Mangan to Zwolinski) dated December 11, 1985, the licensee forwarded Revision 7 of the program and stated that the "second 120 month interval is scheduled to begin after the 1986 refueling and maintenance outage".
At the time of this inspection, the licensee had not fully implemented the program although the above outage concluded on about June 20, 1986. For example, the program provides for i
- _ _ . . - . _ _ _ _ _ . _ _ _ . _ _ _ . . _ . _ _ _ _ . . _ _ . . m.__ - ._ -._ _ _ _ _ -_, _ _ _ . . - - _ - .
i 21
, testing of the Shutdown Cooling Pumps. The licensee advised that the procedures for these tests were under preparation and not yet implemented.
The acceptability of overall program implementation and its approval status remains unresolved and has been referred to NRC management for review and disposition. (50-220/86-13-07) l Feedwater/HPCI Pump Testing During the review of post maintenance surveillance testing, several discrepancies were identified in the licensee's implementation and documentation of the testing.
During 1980 - 1986 the licensee performed extensive and varied maintenance on the motor driven main feed (HPCI) pumps. For I
the #11 pump, this included replacement of the impeller in 1980 with a new design. During testing following another major overhaul during July, 1986, the licensee found that the
- 11 pump's performance no longer matched the existing pump l performance curve, apparently as a result of the modified l impeller and other maintenance and modification over the five year period.
l TS 4.1.8 requires that the HPCI pumps be demonstrated operable but provides no quantitative acceptance criteria. Licensee
- procedures N1-ST-03 and N1-ST-ICS (see Attachment A) are used l to demonstrate compliance with TS 4.1.8 and provide quanti-l tative acceptance criteria for total flow (3800 gpm at rated pressure) and agreement with the pump performance curves.
i These procedures' acceptance criteria were established in accordance with the TS Bases for 4.1.8.
l Because the pump could not meet the procedure and TS Bases I
criteria, the licensee obtained informal NRC:NRR concurrence (NMpC Memo 67061, Attachment A) to consider only total deliv-ered flow (3800 gpm) as a demonstration of pump operability
- pending receipt of new pump curves from the pump vendor.
On July 14, 1986, data was collected for both the #11 and #12 HPCI pumps to demonstrate compliance with TS 4.8.1, apparently without the use of either surveillance procedure (above) nor
, any other formal procedure. The data was recorded in the i shift log. The licensee was unable to explain why the exist-ing surveillance procedures had not been revised to reflect the new, though temporary, acceptance criteria and then used for performance of the surveillance.
22 TS 6.8.1 and Administrative Procedure 4.0, Administration of Operations, require that all operations shall be conducted in accordance with approved procedures. Conduct of the above surveillance activities without the use of an approved proce-dure constitutes one example of violation (50-220/86-13-08).
The inspector reviewed the above data, including verification of the licensee's calculated conversions and adjustments for pump head, fluid density, and mass flow. Updated pump curves based on analytical data had since been received and were also reviewed by the inspector. No further discrepancies were identified.
On June 18, 1986, procedure N1-ST-C3, Automatic Startup of HPCI System, was performed to demonstrate compliance with TS.
4.1.8a as part of the outage restart activities. The procedure Data Sheet, Return To Normal section (corresponding to proced-ure Section 8.0) provides for recording the " return to ser-vice" Feedwater System and Condensate System alignments. This section would typically be completed to document the overall system status upon completion.
For the June 18 performance, this section was initialed by the operator as complete and the entire procedure was signed off as satisfactory but no system lineup status was recorded.
This failure to record system lineup constitutes a second example of a violation of TS 6.8.1 and AP-4.0 for failure to properly implement procedures (50-220/86-13-08)
Containment Spray Raw Water pump Testing During the 1986 outage, the Raw Water Pump #112 underwent a major overhaul. Procedure N1-ST-Q6, Containment Spray and Raw Water pump Operability Test, was performed on August 14, 1986 to demonstrate compliance with TS 4.1.7 and to collect inser-vice test and related data. Procedure Section B.2.b.(4) requires observation of Raw Water Pump #112 motor current and provides an acceptable range of 54-66.5 amps.
During the above performance, a motor current of 70 amps was observed. The operator annotated the procedure acknowledging the unsatisfactory data, indicating that it was likely due to a change in the pump /metor characteristics following overhaul.
The procedure was signed off by the operator as " satisfactory, corrective action required" and " Step B-2-4-b - need an update on amps". No supervisory or engineering evaluation of the data was evident nor had the procedure been revised during or after the test to reflect a new baseline for the motor operating current.
23 The motor current acceptance criteria and the trending of motor current are not specific regulatory requirements, (i.e. not required by license conditions or regulation). However, the licensee had established them as criteria for satisfactory completion of the procedure demonstrating normal pump opera-tion.
Signoff of the procedure with unacceptable data and/or without resolution via a procedure revision is a third example of a failure to properly implement procedures (50-220/86-13-08).
- d. Instrumentation and Control Maintenance and surveillance activities for the equipment listed in Section 3.2 were reviewed to verify adequacy of post maintenance testing. In all cases reviewed, post maintenance testing was found to be thorough, adequate, and complete.
Those work requests reviewed are listed in Attachment A.
- e. Trendino and Corrective Action Trending of equipment failures, malfunctions, and maintenance is performed on an annual basis in accordance with Administrative Procedure AP-5.0, Procedure for Repair. The key plant departments are required to conduct a review of inservice test and inspection, corrective and preventive maintenance activities for similar failures, high failure incidence rates, etc. The departments then issue a report of their results and recommendations.
The Technical Superintendent then reviews the departmental reports and makes recommendations for program or equipment improvements to the Operations Assessment Committee for review and assignment. Assignment items are administered as Site Operations Review Committee (50RC) action items, and are tracked on a computerized open items listing and are required to be formally closed by the submittal of completed actions to the 50RC. The inspector reviewed the reports for the Mainten-ance, I&C, and Computer Department and discussed the status of the licensee's review with the Technical Support Supervisor.
The inspector noted that the licensee also maintains an active participation in the INPO Nuclear Plant Reliability Data System (NPRDS). The licensee has used the NPRDS data for specific failure studies and reviews but does not include an overall review of the plant specific or generic NPRDS data as part of the annual review per AP-5.0. In response to the inspector's comments, the licensee reviewed the above
24 practices and issued NMPC Memo NMP-20379 (Attachment A) directing how NPRDS data is to be incorporated in the annual reviews.
The inspector also reviewed other relevant programs which evaluate potential problem trends. Technical Support Ser-vices performs reviews of Occurrence Reports, NPRDS Input, reports to NRC and other operations assessment mechanisms and reports apparent trends and recommendations to the 50RC. These items are tracked by SORC as open action items.
The inspector reviewed the computer listing of SORC open items, finding that several dozen items remained open from 1982 - 1985. These items were not necessarily trending or operations assessment items but spanned many aspects of SORC activity.
Typical examples included:
50RC Item No. Title 82-86 Need test schedule for CO2 and Halon Systems 83-38 Need procedure change to exercise switches84-109 Verify that procedure for scram solenoids contains caution for use of loctite 83-44 Tech Spec change needed for contain-ment atmosphere post LOCA vent line modification 83-42 SE on Emergency Condenser mods will require changes in FSAR & procedures The Technical Superintendent demonstrated that the items remaining from 1982 - 1985 had not been formally closed pending assembly of cocumentation packages and that all the actual required actions had apparently been completed. One individual had been assigned to prepare the documentation packages and another to verify their completeness and close each item,
- 6. QA/QC Interface A Senior Supervisor - Audits and eight auditors, and a Supervisor - QA Services with a staff of six report to the Manager, Corporate Quality Assurance. An onsite Supervisor and seven auditors report to the offsite Senior Supervisor - QA Audits. This group is responsible for conducting audits required by Technical Specifications and the QA Program except for those associated with vendor evaluation and control.
25 A Supervisor - Quality Engineering and fifteen engineers, Supervisor -
Materials and Reliability Engineering with a staff of seven report to the Manager, Quality and Reliability Engineering. The Procurement Quality sub group is responsible for vendor surveys and evaluations, annual reevaluations, and vendor visits to inspect or conduct surveil-lances. They also establish and maintain the approved vendor list that is distributed as a controlled document.
The Supervisor - Quality Control and twelve inspectors, and Supervisor -
Operations Surveillance with a staff of seven (for Unit 1) report to the onsite Manager, Nuclear QA Operations. The QC group is responsible for l performing the inspection / witness points that are established in work procedures or WRs. The Operations Surveillance group conducts a second level monitoring / observation of ongoing activities such as maintenance, modification installation, and QC first level inspections.
The managers of the above three groups and a Manager, Non-Nuclear QA Operations report to the Vice President, Quality Assurance who reports directly to the President of the company.
Audits were conducted on functional areas that were reviewed during this inspection. Checklists were used, deficiencies were followed up, and
! the audits were completed within the general framework of the schedule.
l The Qualified Contractor List (approved vendors) was controlled and maintained. A sample of four vendors asssociated with the purchase orders listed in Attachment A were on the list, had been surveyed, evaulated, and reevaluated, and appropriate vendor visits conducted.
' Those WRs (Attachment A) classified as safety related had QC Group involvement as evidenced by the inspections and observations documented on Quality Control Inspection Checklists (QCIRs). Also, the QA Surveillance group conducted some second level overview of the functional areas reviewed during this inspection.
s ,
l No violations were identified.
l 7.0 '!Jnre' solved Items l ,
Unresolved items are matters about which more information is required in
~
n order to ascertain whether they are acceptable, deviations or viola-
..- tions. Two unresolved items were identified during this inspection and are discussed ir paragraphs 4.2 and 5.2.c.
8.0 Management Meetings ,
Licenseemanagementwas9[formedofthescopeandpurposeoftheinspec-tion at an entrance me~e ting conducted September 10, 1986. The findings of the inspection wereTdiscussed with licensee representatives during the course of the inspection. An exit meeting was conducted l , September 19, 1986 at the. conclusion of the inspection (see Paragraph 1
_. _m.__ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _
26
- for attendees) at which time the licensee management was informed of the inspection results. At no time during this inspection was written material provided to the licensee.
1 h..
ATTACHMENT A 1.0 References / Requirements f.
l NRC Generic Letter 83-28, " Generic Implications of Salem ATWS Events"
/[ 10 CFR 50, Appendix B PNPS Technical Specifications FSAR Quality Assurance Program ANSI N18.7-1972, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants.
ANSI N45.2.2-1972, Packing, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants.
ANSI N45.2.12-1974, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.
ASME Boiler and Pressure Vessel Code Section XI, 1974 edition.thru summer of 1975 addenda, and 1983 edition with addenda.
NMPC Ltr to NRC:NRR, GL 83-28 Supplemental Response and Schedule Update, dated 7/31/84 NMPC Ltr to NRC:NRR, Status Update - GL 83-28 Items, dated 4/30/84 NMPC Ltr to NRC:NRR, Status Update - GL 83-28 Items, dated 2/29/84 NMPC Ltr to NRC:NRR, Responses to GL 83-28, Items 2.1 and 2.2.2, dated 9/4/84 ,
t I
, , - . , - , , . - . - ~ . - - - ~ . . m- = ~ , - ,, m ,v,- - ,4 .m., - ,. m-.,.--,-%-. - - , , m.v.,-- ---
r--v - --, -- ,w ..-n .
2 Attachment A 1.0 References / Requirements Continued NRC Safety Evaluation, Generic Letter Items 3.1.1, 3.1.2, 3.2.1, 3.2.2, 4.5.1 for NMP-1 dated March 3, 1986 NMPC Ltr to NRC:NRR, Status of GL 83-28 Items, dated 12/31/85 NMPC Ltr to NRC:NRR, Clarification to Responses to GL 83-28 Items 2.1, 2.2, 3.1.3, 3.2.3 and 4.5.3, dated 7/2/85
~NMPC Ltr to NRC:NRR, Commitment for Refueling Outage Tests of Backup Scram Valves, dated 12/31/84 NMPC Ltr to NRC:NRR, Supplemental Response to GL 83-28, Items 2.1, 3.1, 2.2, dated 11/30/84 NMPC Ltr to NRC:NRR, Response to GL 83-38, dated 11/8/83.
NMPC Ltr +o NRC:NRR, Schedule Extension, GL 83-28 dated 1/30/85 Corrective Maintenance Review Report - I&C, 2/11/86 Corrective Maintenance Review Report - Electrical & Mechanical, 2/17/86 Corrective Maintenance Review Report - Computer Operations & Maintenance, 1/27/86 NMPC Memo, 6706I, NRC Telecon Re HPCI Operability, 7/15/86 Worthington Pump Division Ltr, Feedwater Pumps Impeller Redesign -
Estimated Pump Performance Curve, dated July 22, 1986.
NMPC Memo, MD-86182, Motor Driven Feedwater Pump Expected Performance Curves, 8/7/86 hMPC Memo, HPC1, Surveillance Testing Concerns, 8/1/86 Including Reactor Feel Pump curves (Head Flow, dP-THD, Ib/hr-GPM)
3 Attachment A 1.0 References / Requirements Continued NMPC Memo, 7026I, Notes of Telecon with NRC Regarding Implementation of_ Generic Requirements and App. J. SER, 8/28/86 NMPC Memo, 69 SSP, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 8/19/86 NMPC Memo, 68461, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 8/5/86 NMPC Memo, 6831V, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 7/31/86 NMPC Memo, 65361, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 6/20/86 NMPC Memo, 64811, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 6/11/86 NMPC Memo, 61481, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 4/30/86 NMPC Memo, 57241, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 3/7/86 NMPC MEMO, 5706I, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 3/8/86 NMPC Memo, NMP 14527, Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 10/1/85 NMPC Memo, MNP 13821, Notes of Telecon with NRC Regarding Appendix J Technical Specification Cnanges, 9/11/85 NMPC Memo, (No File #), Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 4/16/85 NMPC Memo, (No File #), Notes of Telecon with NRC Regarding Appendix J Technical Specification Changes, 4/11/85 IE Info Notice 86-78, Scram Solenoid Pilot Valve (SSPV) Rebuilt Kit Problems, 9/2/86
4 Attachment A 1.0 References / Requirements Continued IE Info Notice 86-71, Recent Identified Problems With Limitorque '
Operators, 8/18/86 IE Info Notice 86-34, Improper Assembly, Material Selection, And Test of Valves And Their Actuators, 5/13/86 2.0 Documents Reviewed 2.1 Procedures / Tests Nine Mile Point Unit 1 Q-List, Revision B AP-3.4.2, Operations Experience Assessment, Revision 2 TDP-5, Administration of Operations Engineering Assessment Items, Revision 1 TDP-8, Post Maintenance Testing Criteria, Revision 1 DCl-4, Instruction for Control of Station Manuals, Revision 3 N1-1CP-C-FWC-2, FW System Reactor Level Calibration, Revision 0 Performed 6/7/86 N1-1MP-38, Shutdown Cooling Calibration, Revision 4, Performed 6/10/86 N1-1MP-39, Emergency Cooling Timer Maintenance, Revision 0, Performed 3/29/86 N1-ISP-M-36, Lo2 & Lo3 Rx Level, Revision 0, Performed 8/18/86 N1-EPM-GEN-R-120&21, Limitorque Valve & MCC Inspection, Revision 0, Performed 6/11/86 N1-EMP-27.7, Maintenance of SDC Pump Motors, Revision 0 N1-EMP-44.18, Limitorque Preassembly of Type SMB&SB, Revision 2, Performed 5/24/86 N1-EMP-44.26, Replacement of HFA Relays, Revision 0, Performed 3/13/86 N1-NMP-6.2, Overhaul of CRD, Revision 0 N1-MMP-6.3, Installation of CRD, Revision 1 N1-MMP-8.1, Overhaul of SDC Pumps, Revision 0
5 Attachment A N1-MMP-8.2, Maintenance of SDC Valves, Revision 0 N1-MMP-8.3, Maintenance of SDC Valves, Revision 0 N1-MMP-8.4, Maintenance of SDC Valves, Revision 0 N1-MMP-8.5, Maintenance of SDC Valves, Revision 0 N1-MMP-9.1, Maintenance of Emergency Condenser Steam Isolation Valve, Revision 0 AP-8.4, Procedure for Control and Calibration of Equipment used in Tests and Inspections, Revision 4 NI-MMP-9.2, Maintenance of Emergency Condenser DC Motor Operated Steam Isolation Valves, Revision 0 N1-MMP-9.3, Overhaul of Emergency Condenser Condensate Isolation Valves AP-5.0, Procedure for Repair, Revision 7 AP-4.0, Administration of Operations, Revision 6 AP-2.0, Protection and Control of Procedures, Revision 6 AP-3.4.1, Administration of Technical and Safety Reviews, Revision 0 S-MI-GEN-002, Maintenance Instructions for Writing Procedures, Revision 0 5-IDP-PO, Outline for I&C Procedures, Revision 6 TDP-6, NPRDS Failure Reporting, Revision 1 TDP-9, Independent Safety Engineering Group, Revision 0 NEL-0146, Control & Distribution of Vendor Documents, Revision 0 AP-8.1, Preventive Maintenance AP-10-2.2, Procedure for Reporting Variations From Normal Plant Operations, Defects and Noncompliances, Revision 1 N1-EPM-C16, Data Sheet, Protective Relay Setting & Testing Checklist, Revision 7, Performed 2/11/86, 1/24/86 PTP-232.66, Preoperational Test - Replacement of RPS HFA Relays, 4-111A/40 and 4-111B/40, Revision 0, Performed 3/20/86 AP-8.5, Housekeeping and Cleanliness Contract, Revision 1
6 Attachment A N1-ST-C3, Automatic Startup of HPCI, Revision 2, Performed 6/15/86 and 6/18/86 Containment Spray System Description, Revision 0 Feed Water /HPCI System Description, Revision 0 Shutdown Cooling System Description, Revision 0 Emergency Cooling System Description, Revision 1 Control Rod Drive Hydraulic System Description, Revision 0 1986 Occurrence /LER Log (Nos. 86-01 to 86-1542)
SORC Open Item List, dated 9/11/86 N1-ST-ICS, HPCI Surveillance with Inoperable Component Test, Revision 3, Performed 7/1/86 NI-ST-03, HPCI Pump Operability Test, Revision 1, Performed 8/12/86 N1-MMP-7.1, Overhaul of Ele:trien1 FW Pumps Data Sheet, Performed 7/4-13/86 POT-223.12, Replacement of RPSHFA Relay 12K6, Revision 0, Performed 3/18/86 POT-223.60, Replacement of RPSHFA Relay 11K23, Revision 0, Performed 8/18/86 POT-223.9, Replacement of RPSHFA Relay 11K5, Revision 0, Performed 8/18/86 POT-232.11, Replacement of RPSHFA Relay 12K5, Revision 0, Performed 3/14/86 POT-233.10, Replacement of RPSHFA Relay 11K6, Revision 0, 3/13/86 N1-EPM-V6, Visual Inspection of HFA Relays for SIC 44, Revision 2 and IE Bulletin 84-02, Revision 1 N1-EPM-VS, Procedure for Testing MOV's Utilizing MOVATS, Revision 1 N1-ST-Q2, CRD Pumps Flow Rate Check, Revision 4, Performed 8/18/86 N1-ST-06, Containment Spray and Raw Water Pumps Operability Test, Revision 18, Performed 8/14/86
7 ' Attachment A N1-ST-R1, Control Rod Scram Insertion Time Test, Revision 7, Performed 8/8/86 N1-1CP-Q-80, Containment Spray Flow and Pressure Calibration, Revision 5, Performed 5/14/86 and 8/8/86 N1-1MP-44.2, CRD Scram Valve Timing, Revision 0, Performed 5/10/86 and 8/25/86 N1-ISP-C-25.3, Local Leakage Test Containment Spray Raw Water Heat Exchanger, Revision 5, Performed 3/11/86 and 5/20/86 N1-MMP-13.1, Maintenance of Containment Spray Pump, Revision 1, Performed 3/20/86 and 4/2/86 N1-MMP-13.6, Maintenance of Containment Spray Heat Exchangers, Revision 0, Performed 5/25/86 N1-MMP-13.7, Maintenance of Containment Spray Valves, Revision 0, Performed 4/2/86 and 5/10/86 NMP-13.4, Maintenance of Containment Spray Inlet Isolation Valves, 4 Revision 0, Performea 5/12/86 and 3/28/86 N1-IMP-50V-3, Repair / Replacement of Miscellaneous Solenoid Operated Valves, Revision 0 N1-IMP-SOV-2, Repair / Replacement of CRD Scram Pilot Solenoid Valves, Revision 0, Performed 5/3/86 N1-IMP-SOV-1, Replacement / Repair of Solenoid Operated Valves Which Control Primary Containment Isolation Valves, Revision 1 N1-ISP-C44.2B, Qualitative Test of the Backup Scram Valves, t Revision 0, Performed 4/13/86 DCI-1, Station Incoming / Outgoing Correspondence Control Instruction, Revision 2 DCI-3, NMP Drawing Control Instruction, Revision 7 l DCI-4, Instructions For Control Of Station Manuals, Revision 3, ND-100, Plant Modifications, Revision 1 ND-160, Drawing Change Control, Revision 0 ND-220, Plant Configuration Verification, Revision 0 1
l
h_
8 Attachment A TDP-8, Port Maintenance Testing Criteria, Revision 1 N1-IMP-38, Shutdown Cooling, Revision 4 N1-ST-Q7, Manual SCRAM Instrument Test, Revision 6 N1-EST-C5, Emergency Undervoltage Relay Surveillance Test, Revision 1 APN-9, Procedure For Station Permanent and Temporary Modifications-and Replacement, Revision 3 .
APN-17, Management of Station Records, Revision 5 N1-EPM-C12, Type AK Breaker / Motor Inspect and Breaker Load Test, Revision 2 2.2 Vendor Manuals Valve No. Valve Model Limitorque Operator 38-01&l3 AC Grave Model 900 SMB-2 38-02 DC Grave Model 900 SMB-3 38-09/10/11 AC Fisher Type 657AR 31-07/08 Rockwell Class 900 SMB-2 31-01/02
- Chapman #SP923' S0117/118 ASCO HVA 90 4052A Pumps NU-02A/B/C PacificModelHVCH FWP 11 & 12 Worthington 8WNC-141 2.3 Prints / Drawings 18012-C, Reactor Containment Spray System, Sheet 2, Revision 13 18016-C, Control Rod Drive, Sheet 1, Revision 13 18016-C, Control Rod Drive, Sheet 2, Revision 11 22008-C, Control Rod Drive Interconnection Wiring Diagram, Sheet 1, Revision 8 22008-C, Control Rod Drive Interconnection Wiring Diagram, Sheet 2, Revision 3 22008-C, Control Rod Drive Interconnection Wiring Diagram, Sheet 3, Revision 5 2200S-C, Control Rod Drive Interconnection Wiring Diagram, Sheet 4, Revision 4 18017-0, Emergency Cooling System, Revision 24 18018-C, Reactor Shutdown Cooling, Revision 10 18005-C, Feedwater Flow H.P., Sheet 1, Revision 11 18005-C, Feedwater Flow H.P. , Sheet 2, Revision 13 19859-C, Reactor Protection System Trip Diagram, Sheet 1, Revision 9
i 9 Attachment A 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 2, Revision 27 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 5, Revision 26 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 6, Revision 21 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 8, Revision 31 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 8A, Revision 8 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 9, Revision 16 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 10, Revision 19 19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 10A, Revision 5
-19859-C, Reactor Protection System Elementary Wiring Diagram, Sheet 17, Revision 11 C-19408-C/2, One-line Diagram Main and Secondary Connections, Revision 19 C-19409-C/1A, One-line Diagram Auxiliary System (Powerboards),
Revision 3 C-19409-C/3, One-line Diagram 4160V Emergency Boards 102/103, Revision 20 C-19410-C/11, 4.16kV Emergency Powerboard and DG #102 UV Relays, Revision 5 C-19410-C/12, 4.16kV Emergency Powerboard and DG #103 UV Relays, Revision 5 C-34122-C/1, Aux Control Cab 1548 ATWS Channel #11, 3/9/82, Revision 2 C-34122-C/1, Aux Control Cab 1S48 ATWS Channel #11, 8/29/82, Revision 4 C-34122-C/1, Aux Control Cab 1548 ATWS Channel #11, 7/1/86, Revision 3 C-34122-C/2, Aux Control Cab 1S48 ATWS Channel #12, 8/29/82, Revision 4 C-34122-C/2, Aux Control Cab 1548 ATWS Channel #12, 7/1/86, Revision 3 C-34122-C/2, Aux Control Cab 1S48 ATWS Channel #12, 3/9/82, Revision 2 C-34814-C/1, Remote Shutdown Panel #11, Revision 5 C-34814-C/2, Remote Shutdown Panel #11, Revision 9 C-34814-C/3, Remote Shutdown Panel #12, Revision 5 C-34814-C/4, Remote Shutdown Panel #12, Revision 7 C-34816-C/2, Remote Shutdown Panel #12, Revision 3
10 Attachment-A C-34816-C/4, Remote Shutdown Panel #12, Revision 1 C-34816-C/5, Remote Shutdown Panel #12, Revision 0
/
C-34816-C/6, Remote Shutdown Panel #12, Revision 1 2.4 Work Requests 37171, Replace HFA Relays - CS Valve & Recirc Pump Trip, 4/15/86 14168, Repair Limitorque Operators, 32-NG(08 B), 5/26/86 10582, Replace CR0 26-23 & Position Indicator, 4/25/86 100168, Repair #11 MDFP-Failed HPC1 Head / Flow Test, 7/14/86 014341, Recirc Valve 30-31 Would Not Stroke During HPC1 Test, 5/17/86 010238, Overhaul #12 MDFP, 3/86 023454, Overhaul #12 MDFP, 5/86 37070, Replace HFA Relays 11K5, 3/31/86 37065, Replace HFA Relays 11K23, 3/13/86 37150, Replace HFA Relays 12K6, 3/14/86 37097, Replace HFA Relays 11K6, 3/13/86 32408, HCU 14-19 Scram Outlet Valve CV127, Uternram Ieak, 4/29/85 32893, Core Spray Topping Pump 111, Inspection, 3/1/E6 34415, Core Spray Pump 111, Heater Repair, 8/27/85 32158, Containment Spray Pump 121, Rebuild, 12/17/85 32686, Containment Spray HX 111, Weld Repair, 5/12/85 32689, Containment Spray HX 121, Weld Repair, 5/20/85 32713, Containment Spray HX 111, Weld Inspection, 6/1/85 010056, Containment Spray HX, 121, Head Removal, 3/13/86 014456, Containment Spray HX 121, End Plate Removal For Tube Test, 5/28/86 010231, Containment Spray Block Valve 80-11, Leaking, 5/30/86 32172, HCU 42-19, Scram Pilot Valves 117/118, Rebuild, 11/4/85 30776, HCU 50-27, Scram Pilot Valve 117 Rebuild, 3/5/85 32173, HCU 42-23, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32174, HCU 42-27, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32176, HCU 42-35, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32175, HCU 42-31, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32177, HCU 46-23, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32178, HCU 46-35, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32179, HCU 42-39, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32180, HCU 42-43, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32181, HCU 26-51, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32182, HCU 22-51, Scram Pilot Valves 117/118, Rebuild, 11/4/85 32183, HCU 26-47, Scram Pilot Valves 117/118, Rebuild, 11/4/85-35550, HCU 38-11, Scram Pilot Valves 117/118 Rebuild, 11/4/85 010763, New Containment Spray HX 111, Cover Replacement, 5/29/86 010273, Containment Spray Pump 111, Rebuild, 6/15/86 011011, Rosemont Model 1153 Series D, E.Q. "0" Ring Replacement, 4/3/86 010987, Containment Spray Inlet IV 80-15, Binding, 4/21/86 010235, Containment Spray Inlet IV 80-16, Leaking, 4/6/86 31281, Containment Spray Valves, Leaking Solenoids, 4/17/85
11 Attachment A 010565, Containment Spray Breaker for 80-03, 5/29/86 10223, RPS Instruments (ATWS) IE Notice 85-02 Verify Prop, 3/9/86 10778, Replace 0-Rings On Transmitter PT 36-21C ATWS, 3/18/86 13877, Re-brush MG-set 162, 6/16/85 14094, RSP #11, Replace barriers & SBM SW Brackets, 9/17/86 14095, RPS #12 Repair / replace Tubing For Incomming Cable, 9/17/86 14096, RPS #11 Repalce Covers On Aux Relays And SMB SWs, 9/17/86 15725, Replace Burntout 2" Relay On MG-set 172, 6/11/86 16296, ATWS Level Xmitter. Test Valve Leaks, 6/20/86 24489, Trouble Shoot Shutdown Cooling Pump #13, 3/21/84 25084, Inspect Breaker On Shutdown Cooling Pump #11, 3/20/84 25935, Investigate Breaker On Reactor Shutdown Cooling Pump, 4/5/84 31305, Re-brush MG-set 172, 2/25/85 31433, Investigate LCVs Operation From Remote Shutdown, 2/15/84 31995, Re-brush MG-set 171, 7/1/85 31997, Re-brush MG-set 162, 7/1/85 31999, Re-brush MG-set 11, 7/1/85 33060, SDV Level Instrument Does Not Work, 6/3/85 33278, Replace Whitey Valves On Transmitters 36-21, A,B,C, 6/12/85 33914, Replace Light Switch On MG-set 131, 12/9/85 34853, Re-brush M3-set 172, 9/30/85 35570, SCRAM Pump Vol Water Lvl Inspect / repair / replace, 11/16/85 35837, RSP's Rx Water Temp, Charige Scales For Non-linear, 11/25/85 36033, RSP #11 Calibrate Rx Water Temp, 11/24/85 37171, Replace HFA Relays Core Spray Viv & Recirc Pump, 2/11/86 100181, Investigate MG-set 141 For Problems, 7/6/86 100623, Replace Voltage Regulator MG-set 141, 7/17/86 105004, RSP #11 Meter Internals Missing, Replace Meter, 9/17/86 105026, RSP #12, Check Gaitronics Wiring, Write DCR If Changed 105027, RSP #11 Check Gaitronics Wiring, Write DCR If Changed 105674, Lable Cable Numbers Correctly in Rx Recir Control, 9/16/86 105675, Clean Cabinet RY21C, RY21A, RY20A, RY210, RY200, 9/16/86 2.5 Purchase Orders and Associated WRs 40697, HFA Relays (WR 37070) 54333, Shutdown Cooling Pump Shaft Sleeve Collar (WR 22381) 22123, Bearings (WR 22381) 94420, Flexitallic Bonnet Gasket (WR 34932) 22014, Index Tube (WR 011312) 50106, Valve Poppet (WR 34932) 94434, Lubricant (WR 011312) 74123, Cylinder and Flange Assembly Tube (WR 11312) 21632, Rods and Tubes (WR 11312) 74124, 0-Rings Piston Tube (WR 11312) 87787, Scram Valve Diaphrams (WR 105641) 1-86-0369, Replacement of HFA Relays RPS-11/4-1118140, 4/6/86 1-86-1195, Overhaul of Electrical FW Pump #11, 7/4-13/86 1-86-0144, Replacement of HFA Relt,y 11KS
m l
12' Attachment A 1-85-257, Replacement of Air Diaphram on Scram Outlet Valve on HCU 14-19, 4/29/85 i 1-85-269, Containment Spray HX 111, Weld Repair, 5/13/85
.1-85-276, Containment Spray HX 111, Weld Examination, 5/31/85 1-86-0134, Containment Spray HX 121, Head Removal, 3/12/86 1-86-1037, Containment Spray HX 121, End Plate Installation, 5/27/86 1-86-1003, Containment Spray Block Valve 80-11, Picking Replacement, 5/25/86 1-85-0474, HCU 42-19 Scram Pilot Valve 117/118, Rebuild, 11/4/85 1-86-0402, New Containment Spray HX 111, Cover Replacement, 4/9/86 1-86-0287, Replace "0" Rings in Signal Transmitters, 4/4/86 1-85-121, SOV for 80-35 Replacement, 4/16/86 2.7 Audits / Vendor Control SY-RG-IN-86003, Syracuse Nuclear Technology SY-RG-IN-86011, Implementation of NRC Generic Letter 83-28, Salem ATWS Event SY-RG-IN-85014, Nuclear Design Engineering SY-RG-IN-86006, CPS Electric SRAB Audit B, Operations Activities GE Wilmington Nuclear Energy Products Division (survey, evaluations, visits,etc.)
GE Power System Management Business Department (visits, etc. at corporate offices and three manufacturing locations)
Worthington Pump Corporation (visits, etc.)
2.8 Cesion Change Requests N1Y86MX001LG040, RSP #12 Correct drawings for field installation, 9/18/86 2.9 Tr,aining NMP Unlicensed Operators training Curricula / schedule 1984, Revision 0 l
- g. .. ATTACHMENT 5 gg
,- gf f
'g UNITE 3 STATES g g NUCLEAR REGULATERY COMMISSION 3 j REGION I g 631 PARK AVENUE
, %, f KING oP PRUSSIA. PENNSYLVANIA 19406 19 NOV 1986 Docket No. 50-410 Niagara Mohawk Power Corporation ATTN: Mr. C. V. Mangan Senior Vice President 300 Erie Boulevard, West Syracuse, New York 13202 Gentlemen:
Subject:
Inspection Report No. 50-410/86-52 This refers to the special team inspection of the Nine Mile Point Unit 2 em-ployee concerns about quality assurance (QA) program effectiveness in the area of quality control (00) inspections and the Quality First Program (Q1P) in-vestigation of these concerns. This inspection was conducted on September 8-12, 1986, and a summary of our findings was presented by Mr. J. C. Linville to Mr. W. Donlon and members of your staff at the inspection's conclusion. The inspection on August 6-28, 1986 by Mr. R. Gramm which brought these matters to our attention, and management meetings on August 29, 1986 at the site and on September 4, 1986 in Region I to discuss them are also summarized.
Overall the inspection validated your position that the programmatic concerns identified in the Q1P investigation have not resulted in hardware deficiencies or impeded QA department personnel from identifying hardware deficiencies. In addition, there is evidence that QA overchecks of QC inspection activities have identified and corrected hardware deficiencies.
While no violations were identified during this inspection, several new pro-grammatic weaknesses were identified, in addition to those identified by your Q1P investigation, which require additional action to improve program effec-tiveness . The weaknesses are summarized in sections 1.1 and 1.2 of the en-closed inspection report. You are requested to respond to this letter provid-ing a description of your actions relative to these weaknesses.
The response requested by this letter is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
We appreciate your cooperation.
Sincerely, f
William . Kane, Director Division of Reactor Projects h.$h'h0 W &f '
Niagara Mohawk Power Corporaticn 2 9 NOV 1989
Enclosure:
NRC Region I Inspection Report No. 50-410/86-52 cc w/ encl:
Cannor & Wetterhahn John W. Keib, Esquire J. A. Perry, Vice President, Quality Assurance W. Hansen, Manager of Quality Assurance D. Quamme, NMP-2 froject Director C. Beckham, NMPC QA Manager T. J. Perkins, General Superintendent R. B. Abbott, Station Superintendent Department of Public Service, State of New York Public Document Room (PDR)
Local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of New York bec w/ encl:
Region I Docket Room (with concurrences)
Management Assistant, DRMA (w/o encl)
DRP Section Chief Region I SLO D/DRS M. Haughey, NRR, BWR Licensing P. McKee, DRP8, IE D
U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-410/86-52 Docket No. 50-410 License No. CPPR-112 Category B Licensee: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility: Nine Mile Point 2 Locations: Scriba, New York King of Prussia, Penr.sylvania Inspection Dates: August 26-28, 1986 and September 8-12, 1986 Meeting Dates: August 29, 986 and September 4,1986 Team Leader: _
1 //
C."Linvi l l e C ef RPS 2C, DRP 'date Inspectors: Dr. P. K. Eapen, Chief, QA Section, DRS R. Gramm, Senior Resident Inspector J. Stair, Resident Inspector Approved by: //fJ~f 8(.l, R. M. Gallo, Chief, Projects Branch 2 DRP date Inspection Summary: Inspections on August 26-29, and September 8-12, 1986 and Management Meetings on August 28, 1986, and September 4, 1986 (Report No. 50-410/86-52)
Special announced team inspection to evaluate the licensee's position that the programmatic quality assurance (QA) issues identified by Quality First Program (Q1P) concerns 86-64 A to G have not resulted in hardware deficiencies or im-peded the ability of QA department personnel to identify or correct hardware deficiencies. The inspection involved review of licensee records and inter-views with QA department personnel. It included 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> on site by 2 section chiefs, one senior resident inspector and one resident inspector.
O- b t4)
2 1.0 Inspection Conclusions 1.1 The following programmatic QA weaknesses were identified by the NRC in addition to those identified by the Q1P investigation:
(1) Q1P review of QA concerns lacked active independent management involvement similar to that normally provided by offsite review committees (Section 4.0).
(2) Bypassed QC hold points were not trended for frequency, repeat offenders, or repeat by discipline (Section 6.6).
(3) QC hold points in maintenance procedures and related QC check-lists were not consistent (Section 6.3).
1.2 As a result of NRC reviews the following QA program weaknesses pre-viously identified by the licensee's Q1P investigation were clarified and confirmed:
(1) Limits on troubleshooting activities performed under Deficiency Reports (DRs) and Work Rcouests (WRs) are not well defined, and QA Engineering is not providing clearly established inspection attributes to QC inspectors. (Sections 5.0, 6.1 and 6.2)
(2) While QA department personnel have been adoquately trained on QA procedures, there has been no formal training on implementing policy guidance disseminated by numerous memoranda. In addit-ion, there is no vehicle in use for continuing training of new contractor personnel or on new policy guidance. (Section 5.0, 6.6 and 6.9) 1.3 None of the 26 QA department personnel interviewed by inspection team members knew of any uncorrected hardware deficiencies or believed that they had been impeded from identifying or correcting any hard-ware deficiencies because of programmatic QA weaknesses.
1.4 NRC reviews noted evidence that QA overchecks of QC inspection act-ivities have been identifying and assuring correction of hardware deficiencies. (Section 7.0)
2.0 Background
2.1 QC Inspector Certification On August 20, 1986 the NRC received allegations, that NMPC had im-properly certified two QC inspectors, that NMPC QC inspectors had performed inspections outside their certified discipline, and that the NMPC Quality First Program had not been responsive. NRC Region I conducted an allegation review panel to document recipt of the con-cerns and determine appropriate followup actions. As a result, on
3 August 26 and 27 an NRC inspector reviewed NMPC QA procedures, inter-viewed QC personnel, reviewed QA records, and accompanied QC person-nel during the conduct of inspections. Conclusions reached relative to the allegations are described in section 8.0.
2.2 Quality First Program Concern 86-64 Subsequently, on August 28, an NRC inspector reviewed Quality First Program (Q1P) records and noted concerns 86-64 A thru G, which add-ressed the following issues regarding NMPC QC activities:
(1) Lack of management support.
(2) Lack of supervisory guidance.
(3) Lack of procedural training.
(4) Lack of procedural adherence by Startup and Test, and QA.
(5) Inconsistent procedural implementation.
(6) Inadecuate QA communication channels.
(7) Lack of QA management feedback to inspector suggestions or concerns.
(8) Premature Deficiency Report closure.
(9) Bypassed QC witness points.
The Q1P findings had been forwarded to senior NMPC management for resolution. On August 28, Region I management was informed of the above concerns regarding the NMPC QA/QC program implementation.
After becoming aware of these concerns, management meetings between Region I and NMPC were held on August 29 and September 4. A list of attendees is included as Attachment 1.
NMPC concluded that while the concerns of the NMP2 employees had raised some valid programmatic concerns which were being addressed, these concerns nad not resulted in any unsatisfactory hardware.
These conclusions were based on 01P interviews with all QC inspec-
, tors, both NMpC personnel and contractors, which did not identify any unresolved hardware issues. In addition, NMPC stated that the QC function was overchecked by QA audits, QA surveillances, and QA engineering. The licensee position was documented in a letter to NRC Region I dated September 1, 1986 and in Attachment 2 to this report which was presented at the Septemoer 4, 1986 meeting.
NMPC's initial corrective actions in response to the Q1P concerns included meetings between the Vice President, QA and all levels in the QA department to explain departmental policy relative to the concerns addressed and to direct field involvement of QA engineers in defining inspection criteria for troubleshooting activities.
d 4
In response to a question during the September 4, 1986 meeting re-garding how the licensee could be sure that QA department personnel were not being impeded from identifying hardware issues by these programmatic weaknesses, the licensee conducted a survey of all QA department personnel with negative results as described in a September 9, 1986 letter to NRC Region I 3.0 Inspection Method and Scope The inspectors independently evaluated the concerns of former Nine Mile Point Unit 2 (NMP2) employees related to programmatic weaknesses in quality control (QC) inspections of preoperational testing and the impact of these concerns on the acceptability of installed and tested hardware.
These concerns had previously been evaluated by the licensee's Quality First Program (01P) as concern 86-64, which concluded that the concerns re-presented valid programmatic quality assurance (QA) program weaknesses but had not adversely affected the quality of the installed hardware. Also, the inspection evaluated portions of the QA program, other than QC, to determine whether their overcheck functions had been effectively per-formed.
The inspectors reviewed the background information developed by the 01P investigation, interviewed a sample of NMP2 QC inspectors including con-tractors, NMPC QC inspectors, supervision and management, and verified the resolution of selected quality issues for which the acceptability could not be determined through the QC interview and records review process. In addition, the inspectors interviewed a sample of personnel from other QA organizational elements including NMP2 Startup/ Operations Surveillance, NMP2 Quality Engineering, and NMP2 QA Audits to determine whether the pro-grammatic weaknesses identified by Q1P in the QC group impacted the groups providing the overchecks of the QC program. >
4.0 Q1P Review of Concerns 86-64 A to G Based on a review of the files and discussions with the Q1P manager and interviewer, the Q1P review of concerns 86-64 A to G generally proceeded in accordance with QAP 16.70. The planning of the investigation of the concerns was thorough and the conclusions were sound. However, several of the QA personnel interviewed by NRC inspectors questioned the independence of Q1P in reviewing concerns relating to the QA program. The provisions to assure independent management oversight of such concerns by the Adminis-trative Assistant to the President involved merely informing him of such activities rather than active involvement.
It is not evident that the corrective action plan, proposed by the Vice President, QA for the valid Q1P concerns, received the independent manage-ment review and approval required by QAP 16.70 before it was submitted to the NRC by letter dated September 9,1986. However, active participation by the President and Senior Vice President in the corrective action plan after the NRC became involved is acknowledged. The licensee response to this issue will be reviewed during a subsequent inspection. (50-410/
86-52-01)
5 5,0 NRC Interviews During this inspection, 26 QA department personnel were interviewed by the NRC including 12 QC inspectors, 3 QA Engineers, 2 QA Surveillance in-spectors, 2 QA Auditors, and 7 QA management and supervisory personnel.
The interviews included questions in the following areas:
Position Responsibilities Qualifications
- Training QA Program Effectiveness QC Inspection Activities QA Surveillance Activities QA Audits QA Engineering Activities Supervisory Support and Guidance Working relationships with other parts of QA Department Working relationships with other parts of the site organization
- Work experiences Concerns None of the interviewees knew of any uncorrected hardware deficiencies.
Only one individual believed that there could be hardware deficiencies which were not identified related to the control of troubleshooting and work activities as discussed in Sections 6.1 and 6.2. The examples dis-cussed do not represent work on safety-related equipment without any QA/QC oversight because there were Deficiency Reports and Work Requests to cover the work, but the guidance to the workers and the inspection criteria were so non-specific that the acceptability of the work required considerable judgement and continuous QC coverage. The individual acknowledged that these issues had been appropriately dispositioned, but was concerned that work of unacceptable quality could occur when the work scope and inspec-tion criteria are not clearly cefined in troubleshooting activities. Half of the interviewees expressed this general concern without specific exam-ples. The licensees letter of September 9, 1986 addressed the inspection criteria concern by committing to field involvement of Quality Engineers in developing inspection criteria for troubleshooting, but did not add-ress the responsibility of the line organization to place appropriate limits on workers for such work to assure appropriate engineering review i
__ _, _ _ _ - - ~ , _ _ _ _ _ _ _ _ - . . _ _ - _ , _ . . _ . _ - - . _ _ _ _ _ _ _ . -
6 prior to work completion. The licensee response to this issue will be reviewed during a subsequent inspection. (50-410/86-52-02)
While there is evidence of formal training on QA procedures, as discussed in section 6.4, half of the interviewees in the QC group believed that the policy guidance for implementing the QA procedures was inconsistent, poorly communicated and confusing. This situation has been aggravated by reas-signment of supervisory and management personnel, a reorganization which split the Quality Engineering and Quality Control functions, the turnover of many contractor QC inspectors, the transition from a rigidly defined construction QA program to a very flexible operations QA program, and the replacement of formal QA Instructions for implementing QA procedures by uncontrolled memoranda, verbal guidance or no guidance at all. Although the licensee corrective action described in the September 9, 1986 letter addressed these concerns, on a one time basis, there is no continuing program to assure that new employees receive this information or to assure that future policy changes are appropriately communicated. The licensee response to this issue will be reviewed during a subsequent inspection.
(50-410/66-52-03)
Followup cr specific concerns raised during the interviews is discussed in section 6.0.
6.0 NRC Followup of 01P Concerns and Specific Interviewee Concerns 6.1 Control of Troubleshocting Activities As documented in Q1P concern 86-64F, and discussed with an NRC in-spector during an interview, a Quality Control Inspector discovered individuals cutting a hole in safety related panel No. 028 associated with the high pressure core spray system diesel generator, to install a temporary modification for preoperational testing without any docu-mentation to support that work. The NRC inspector spoke with one Startup and Test individual and reviewed the Q1P documentation pack-age relative to this concern. Documentation had been prepared con-cerning this temporary modification in advance in the form of Eng-ineering and Design Coordination Reports (E & DCRs) 258352 and 258352A; Deficiency Reports (DRs) 11173, 17823, and 19281; Inter-office Correspondence (IOCs), ESEG 86-5-12 and 86-5-7; and Problem Report (PR) 0421 A. DR 18649 which controlled the work had been im-properly signed off as completed prior to work completion leading the QC inspector to believe that no documentation covering the work exis-ted. Although the work performed in this case was acceptable, it was not clear to the QC inspector that it would have been without his involvement.
While the finally accepted work was adequate, this is an example for which better definition of the work scope and QC inspection require-ments was necessary to assure the quality of safety-related equip-ment.
7 l 6.2 control of Work Activities During an interview, the NRC was informed of a concern relating to the improper control of troubleshooting activities on a RHR service water sample pump 25WPCAB238. The inspector reviewed Work Request (WR) 103572, QCIR 2-86-3862 and Corrective Action Request (CAR) 86-1012. The extent to which the maintenance personnel cauld dis-assemble the component was not specified on the WR. The pump and motor were apparently disassembled without the reference to the associated vendor manual. This work was an example where direct involvement of Quality Engineers and work scope definition was neces-t sary to specify the extent of troubleshooting activities. Without the involvement of the QC inspector, it was not clear to the con-cerned QC inspector that the work on the pump would have been accept-j able.
l This is another example of a case where the finally accepted work was :
adequate, but for which better definition of the work scope and OC inspection requirements was necessary to asure the quality of safety related equipment.
l 6.3 Inconsistent Hold points Between QC Checklists and Maintenance 1
Procedures During an interview, the NRC was informed that NMPC maintenance pro-cedures and the applicable NMPC QC checklists do not contain consis-tent hold points. The inspector reviewed the following checklists and procedures:
1
' QC Checklist and Maintenance Title Procedure No. o
! Overhaul of CRD Hydraulic N2-MMP-30.3 Control Units Overhaul of Control Rod N2-MMP-30.8 l
Drive Maintenance. of LPCS N2-MMP-32.2 Pressure Pump Maintenance of HPCS Pump N2-MMP-33.1 Maintenance of HPCS Pressure N2-MMP-33.4 Pump Overhaul of Reactor Core N2-MMP-35.1 Isolation Cooling Pump l
8 g
t -
The inspector identified the following discrepancies: :
QC Checklist MMP-30.8 attributes 1 through 10 did not have asso-ciated QC holdpoints in the maintenance procedure MMP-30.8 data sheet, f
Attribute 2 in QC checklists MMP-32.2; 33.1, and 33.4 was not contained in the associated maintenance procedure data sheet. ,
QC holdpoints in the maintenance procedure data sheets were not included on the associated QC checklists as exemplified by item ;
7.26.2 and 7.27.2 of MMP-30.8, and Items 7.1.14, 7.3.9, 7.3.10 t and 7.5.6 of MMP-33.1.
The inspector's review confirmed the expressed concern that the QC checklists and associated maintenance data sheets are inconsistent.
The licensee corrective actions in this area will be reviewed at a later date. (50-410/86-52-04) 6.4 QC Personnel Training As a followup to 01P concerns 86-64 C and E, the inspectors inter- l viewed site QA and QC personnel and reviewed the adequacy of training '
and qualification of these personnel. All these personnel were qualified to ANSI N 45.2.6 Level II. The inspector reviewed the training and qualification records of five QC personnel and verified that their education and experience levels met those specified in ANSI N 45.2.6.
Licensee Procedure QAP 2.10 (Revision 7) establishes the training requirements for QA and QC personnel. This program requires initial, on-the-job, and continued training as well as periodic reading assign-ments. Through a review of selected training records and technical discussion with personnel, the inspector determined that the per-sonnel were qualified and trained in accordance with QAP 2.10, were knowledgeable in the technical requirements of the activities that they monitor, and kept their knowledge level current by completing 1 the required reading of the procedures. In addition to the initial reading of the procedures, the personnel were required to reread these procedures when revised. The inspector noted that the reading was required by the supervisors and proficiency was verified by the supervisors or lead personnel. Adequate time was allocated for the ,
initial reading and rereading. On the average about six procedures had been read by the personnel. In addition to the reading, certain l
key procedures were discussed with personnel during formal classroom !
training. The inspector noted that the QC personnel training records i reviewed indicated attendance at three or more formal training sessions since June 1986, i
t i
i
9 The inspectors determined that the QA and QC personnel were trained on QAPs and that the training records were maintained in accordance with QAP 2.10. No discrepancies were identified.
6.5 Closing of Deficiency Reports (ors) and Policy Guidance l Q1P concerns 86-64 B and G included several examples in which Defici-ency Reports (DRs) had been prematurely closed. The following closed l DRs were reviewed by the inspector for which no associated inspection i j was recorded in the QC DR logbook: l O_R Startup and Test Explanation Computer Status I
10051 Closed Work performed on WCR 9284 10075 Closed Deficiency Report 11861 issued 10108 Closeo Cancelled 10109 Closed Cancelled l 10118 Closed QC determined no inspection necessary l l 12132 Closed Worked on multiple DRs 12449 Closed OC determined no inspection necessary l The inspector reviewed the QC files of open QC checklists for DRs and i Work Requests (WRs). The following documents were reviewed: i Open DR in ,
QC File OCIR Explanation !
15093 2-86-1024 Work previously performed for :'
checklist from EMP-114-1 17088 2-86-1687 Work previously performed with satisfactory checklist '
Open WR in Computer
__Q C File _ Status Explanation 3684 Closed QCIR 2-85-1012 and NCR 2-86-0036 previously issued 100979 Closed QCIR 2-86-2852 and OR 21202 previously issued 103300 Closed Inspector found no leaks so WR was closed without any work i
! 10 The inspector determined that while the QC logs and files were not consistent making it difficult to verify closure, the ors and WRs were properly closed in all cases with appropriate QC involvement.
The inspector reviewed Startup Administrative Procedure (SAP) 121A,
" Deficiency Reporting System". The procedure requires a final QA review of CRs to ensure proper completion. The inspector discussed the review of DRs with associated Quality Engineering (QE) personnel.
The inspector reviewed several logs that indicated QE personnel were fulfilling the procedure requirements. Two programmatic concerns were identified:
The QE staff does not maintain a trend history of DRs that are found unsatisfactory during their document review. For example, instances of bypassed QC holdpoints detected by QE review are not explicitly flagged and trended, although the items are re-solved on a case by case basis.
The inspector was presented QA policy memo 86-001 that described the mechanics of the QE review process. This memo was an example where the general Quality Assurance Procedures (QAPs) did not contain sufficient in'plementing direction for QA personnel, which resulted in the issuance of the informal memo and verbal instructions.
The licensee responses and corrective actions to the above concerns will be reviewed at a later date. (50-410/86-52-05) 6.6 Follow up on Cables with Outer Jackets Cut During an NRC interview, one QC inspector stated that in September 1985 he identified that contractor personnel were cutting into the .
outer jackets of installed cables while trimming back the fire pro-tection foam from the pGCC cables. The contractor personnel used sharp knives to trim back the foam and one cable was cut through to the conductor. DR 05681 E&DCR C 45072 and NMPC NCR 2-85-007 were generated to identify and resolve the concerns associated with cut cables. QC issued Surveillance Report SR-85-10317 to address this concern. The QC inspector who originated this surveillance report did not believe that he had all the facts necessary to accept the corrective actions in December 1985. At that time he believed that the corrective actions were limited to the items discussed in the surveillance report and that an investigation by QC had not been performed to cetermine the existence of similar problems in other cables. Based on the above, the QC inspector requested that his supervisor resolve this surveillance report. The supervisor closed this surveillance report on December 4,1985 based on his understandin; of planned corrective action in this area which was later cocumented in a Report of a Problem dated May 27, 1986. However, the supervisor did not expiain his reasons for closing this surveillance report to the QC inspector.
l
11 The NRC inspector reviewed this issue to assess the adequacy of licensee actions when a surveillance report was closed out by the initiator's supervisor. This concern was reported to the licensee's architect engineer for resolution in a Report of a Problem dated May 27, 1986. An engineering evaluation was performed, and a deci-sion was made to randomly inspect eighty cables from all affected areas. This inspection was completed on August 12, 1986, and it identified four additional cables that were cut but not through the jacket. The inspector determined that the cuts identified in these four cables were within the allowables established in the licensee specifications. The inspector determined that this licensee action was adequate to resolve the concern identified in the QC Surveillance Report.
The inspector furnished a copy of the licensee's engineering eval-uation and addit'onal cable sampling to the QC inspector who had initiated the original QC Surveillance Report. The QC inspector reviewed this additional material and stated that the new analysis and samplino inspection were adequate to address his original con-cern. He also stated that this additional information would enable him to discuss the disposition of this concern in a positive manner should the issue surface again. The inspector had no further quest-ions in this regard.
6.7 Use of Uncalibrated Measuring and Test Equipment During Retests A concern was expressed to the NRC during an interview that Quality Control (QC) inspectors did not verify the use of calibrated measu-ring and test equipment (M&TE) during retests. An NRC inspector interviewed the OC supervisor and was informed that the requirement for M&TE verification applied to retests. The inspector reviewed the NMPC Quality Control Inspection Report (OC1R) files. Based upon the review of OCIR$ 2-86-1314, 2-86-5180 and 2-86-5462 that documented the use of calibrated M&TE during retest activities, the review of additional OCIRS, and the statements of QC supervision, this concern was not substantiated.
6.8 Inadequate Verification of Receipt inspection One OC inspector expressed a concern during an NRC interview that OC inspectors were directed by letter NM QA 1735 dated December 16, 1985 not to perform the pre-installation verification (P!V) or to verify receipt inspection of parts from ministock. In addition, he believed that the stockroom had been directed to stop performing receipt in-spection of spare parts. Discussions with individuals in the Stores Receipt and Inspection department, along with a review of applicable procedures provided assurance that all parts were inspected upon receipt. In addition, memorandums issued provided direction to verify receipt inspection of parts. NMPC letter NM QA 1735 dated December 16, 1935, which instructed that the pIV not continue, was issued to prevent redundancy of inspe: tion since the verification would be per-
i
. .v l 12 formed at installation under the Niagara Mohawk Quality Program. In addition, memorandum 9M STQA 85-13 dated December 20, 1985 provided direction to verify receipt inspection of parts. While this concern l was unsubstantiated, it is another example of the need for training on policy guidance.
! 7.0 Followup on QA Overchecks >
L '
To determine the involvement and effectiveness of NMpC QA overcheck func-tions of the test and maintenance activities, the inspectors reviewed the following documents regarding audits, surveillances and corrective actions
( and discussed them with appropriate managers and supervisors:
NMPC Audits NM-RE-IN-86005, 86008, 86014, and 86017 NMPC Surveillance Reports 86-10510, 10603, 10626, 10640, 10641, l 10677, 10605, 10577, 10594, 10452, 10294, 10131, 10124, 10271, 10368, 10453, and 10445.
Corrective Action Requests 86-1000, 2001, 2002, 1003, 2004, 2005, 2006, 1007, 1008, 1009, 1010 and 1014.
The QA audits were detailed and focused on documented compliance with QA program requirements and Preoperational Test procedure requirements. While the audits met the requirements of 10 CFR 50, Appendix B, their usefulness in assessing the effectiveness of program implementation was limited since they involved little in-process observation of work activities or interviews with personnel implementing the programs and focused on after the fact review of records demonstrating compliance with requirements.
The QA surveillance program was well-structured and implemented to assess programmatic ef fectiveness and identify hardware problems, and assure .
l appropriate corrective action. The surveillances documented an extensive
! review of preoperational test results. In several instances, test eng-ineers had not processed necessary Deficiency Reports (DRs). As a result l of the surveillance program review the ors were subsequently issued. The l surveillance function also had identified problems in controlling the use of red plastic screws to isolate circuits in the control room area and assured correction of the problem. The concerns detected and the scope of corrective actions indicated that satisfactory oversight of the field activities has been maintained.
8.0 QC Inspector certification As discussed in section 2.0, the NRC had previously been informed that NMPC had allegedly improperly certified QC inspectors. The inspector re- i viewed the following documents: f l SWEC and NMPC QC certification records.
QAD 2.60, " Qualification and Certification of QA Inspection and t i Test Personnel". l l I I I l !
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. 13 x s .
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v l - " QAp 2.10,*PTraining".
FSAR Table 1.8-1, page 64-67.
NMPC Quality Assurance Topical Report.
RegulatoryGuide1.5Si"QualificationofNuclearPowerPlantIn-spection, Examination, sand Testing Personnel".
ANSI N45.2.6, " Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants.
- The inspector reviewed the certifications of 6 QC inspectors and deter-mined that they were in accordance with the applicable requirements. The inspector accompanied an electrical inspector during the conduct of a megger test on the Division I diesel generator output breaker. The in- 4 spector revieked the associated work documents and found the QC inspector knowledgeable of the inspection requirements.
The inspector also r6 viewed approximately 220 QCIRs and identified no in-stances of QC, trispectors performing inspections outside the scope of their >
, certification. The in'spector reviewed thirteen additional QCIRs performed by a mechanical QC N pector identified by the alleger as unqualified for i some of-the insp&ctlb
! activities were withins he scope i.the performed and of the QCverified that the inspector's inspectionThe certification.
l inspector determined that the NMPC certification program is in conformance i with commitments and requirements, that NMPC inspectors are properly cert-ified, and that the inspectors are performing work within their capabili-ties. No discrepancies were identified and allegation RI-86-A-099 is not l s substantia.ted.
! 9.0 Exit In_terview The scope and, findings summarized on the inspection cover sheet under l Inspection Results'were discussed with the President, Vice President, QA, and other attendees ide~ntified in Attachment 1 at an exit meeting on September 12, 1986. Based on NRC Region I review of thfs report and dis-cussion held with licensee management at the exit meeting, it was deter-
' mined that this report does not contain information subject to 10 CFR 2.790 restrictions.
1
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t ATTACHMENT 1 Meeting Attendees
- Meeting NMPC !
key 12 4 W. Donlon President 2 4 C. Mangan Senior Vice President 12 4 T. Lempges Vice President, Nuclear Generation 1234 J. Perry Vice President, QA 34 W. Hansen Manager, Nuclear QA Operations 34 C. Beckham QA Engineer 4 G. Doyle Startup Quality Engineer 34 J. Buckley QC Supervisor 4 J. Shepherd QC Lead 4 P. Wilde QA Surveillance Supervisor S4 A. Kovac Q1P Manager 34 E. Manning Q1P Interviewer 34 T. Gamon Q1P Interviewer 34 L. Fenton QA Auditor lead ^
12 4 T. Perkins General Superintendent 4 R. Abbott Station Superintendent 4 G. Afflerbach Work Control Manager 4 K. Dahlberg Maintenance Superintendent 4 R. Mattock Deputy Project Director 34 M. Ray Manager, Special Projects 34 I. Weakley Special Projects 4 G. Griffith Licensing 4 K. Ward Manager Consultant Nuclear Design 34 B. Hooten Special Consultant 4 M.-Boyle Nuclear Compliance and Verification 2 4 J. Beratta Security 3 J. Dillon Audit /Q1P Senior Supervisor 2 G. Wilson System Attorney 2 F. McCarthy Security 2 D. Kerr Corporate Performance Services 2 S. Wilczek Manager Nuclear Technology R. Pasternak 1 2 Manager Nuclear Consulting Services s 2 C. Gresock Manager Nuclear Design
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ATTACHMENT 1 2 s
- Meeting NRC key 2 T. E. Murley Regional Administrator 2 J. Allan Deputy Regional Administrator 1234 W. F. Kane Director, Division of Reactor Projects (DRP) 2 5. Eoneter Director, Division of Reactor Safety (DRS) 2 T. Martin Director, Division of Radiation Safety and Safeguards (DRSS) 2 W. Johnston Deputy Director, DRS 12 S. Collins Deputy Director, DRP 2 R. Gallo Chief, Projects Branch 2, DRP 234 J. Linville Chief, Reactor Projects Section, 2C DRP 234 P. K. Eapen Chief, Quality Assurance Section, DRS 2 G. Meyer Project Engineer 34 W. Cook Senior Resident Inspector 2 4 C. Marschall Resident Inspector 34 W. Schmidt Resident Inspector OTHER 4 P. Eddy PSC Site Representative 4 P. MacEwan NYSEG Manager 4 J. Drake SWEC Startup Special Projects Supervisor 3 K. Roenick PSC Site Representative 2 M. Wetterhahn Attorney, Conner and Wetthahn
" Meeting key 1-August 29, 1986 Management Meeting Attendees 2-September 4, 1986 Management Meeting Attendees 3-September 8, 1986 Entrance Meeting Attendees 4-September 12, 1986 Exit Meeting Attendees
4 9
ATTACHMENT 2 NIAGARA MOHAWK QUALITY FIRST PROGRAM ACTlVITIES PRESENTATION TO U.S. NRC REGION I
( SEPTEMBER 4, 1986 e
Q1P CONCERNS 86-00064A-G ;
RECEIVED FROM THREE QC CONTRACTORS AT TIME OF TERMINATION CONCERN NO. INVESTIGATION CONCERN DRRULTS A. NMPC QC MGT. PREVENTING NCR ISSUANCE VALID WHEN DR'S OR PR'S ARE NOT APPROVED.
B./G. PR'S* AND DR'S CLOSED WITHOUT PROPER VALID RESOLUTION OF OUTSTANDING IR'S. * (INVALID)
SU&T PERFORMING UNAUTHORIZED SIGNOFF OF DR'S.
C. LACK OF SUPPORT FROM NM QC MGT. TO VALID SUPPORT QC PEOPLE IMPLEMENTING QC PROGRAM.
D. QCIR NO. 2-86-0044 CLASSIFIED CAT 1, INVALID UNSAT. ITEMS IDENTIFIED. QCIR IMPROPERLY CLASSED CAT II TO ELIMINATE REINSPECTION.
E. NM QC PEOPLE UNAWARE OF CHANGES IN VALID SAP'S AND QAP'S* (i.e. SAP 1.21A) * (INVALID)
F.
SU&T GIVING VERBAL DIRECTION AND/OR INVALID ISSUING MEMOS TO GIVE ENGINEERING DIRECTION AND RESOLUTION OF PR'S AND DR'S AND TEMPORARY MODS. (i.e. 2FPM-PNL129).
, CHRONOLOGY OF CONCERN 86-64 DATES EVENTS 7/2&3 RECEIVED VERBAL CONCERNS FROM THREE CONTRACTOR QC PEOPLE DURING TERMINATION.
"C" PROGRAMMATIC & ID'ED AS FOLIDWS:
- SU&T DOMINATES QC ACTIVITIES
- LACK OF COMMUNICATION
- PROCEDURES HARD TO WORK TO
- NO JOB DEFINITIONS 7/3-8 RECEIVED ANONYMOUS CALLS THAT INFORMATION GIVEN BY THREE QC PEOPLE WAS LEGITIMATE.
- DECIDED TO INTERVIEW MORE PEOPLE
- QUESTIONS AT INTERVIEW RELATE TO CONCERNS
- DECIDED TO USE 2 INTERVIEWERS / INTERVIEW 7/10 INTERVIEWED 22 PEOPLE.
7/11-31 INTERVIEWED REMAINING PEOPLE.
k
- ALLOWED INTERVIEWEES TO DISCUSS ANYTHING.
- IF RELATED TO CONCERNS - ASKED IF AWARE OF PROBLEMS IN AREA.
- IF CONVERSATION DID NOT ADDRESS CONCERN, THEN QUESTIONS ASKED.
- FOLLOWING INTERVIEW, INTERVIEWERS SHARED NOTES, IDEAS, AND CLASSIFIED THEIR PERCEPTION OF INTERVIEW.
- TOTAL 29 PEOPLE INTERVIEWED, 17 WERE CONTRACT AND 12 WERE NMPC DIRECT EMPLOYEES.
7/14-8/16 Q1P INITIAL INTERVIEWER & INVESTIGATORS ARRIVE AT PLAN OF ACTION TO VALIDATE CONCERNS.
- START INVESTIGATIONS
- TALK TO PEOPLE, REVIEW OILTECTIVE EVIDENCE (DOCUMENTS), CONDUCT INTERVIEWS.
- ARRIVE AT CONCLUSIONS
- WRITE INVESTIGATIVE REPORTS AND RESULTS -
VALID OR INVALID.
1 8/19 MGR. Q1P & INTERVIEWED MET VICE PRESIDENT-Q.A. &
VERBALLY PRESENTED RESULTS AND BACKGROUND MATERIAL FOR V.P.-Q.A.8S ACTION.
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CHRONOLOGY OF CONCERN 86-64 -
(cont'd.)
DATES EVENTS I
8/21 Q1P FORMAL TRANSMITTAL OF ALL CONCERNS TO V.P.-Q.A.
FOR RESPONSE ON A,B,C,E & G.
8/21 V.P.-Q. A. MET WITH MGR. NUCLEAR QA OPERATIONS AND QC SUPERVISOR. DISCUSSED EACH CONCERN, BRAINSTORMED EACH AND DRAFTED RESPONSES.
8/25 DRAFT LETTER AND REVISED RESPONSES FROM MGR. NUCLEAR QA OPERATIONS RECEIVED BY V.P.-Q. A.
8/26&27 LETTER AND RESPONSES STRENGTHENED AND FIRMED UP.
8/28 LETTER WITH RESPONSE DATED 8/27 SIGNED BY V.P.-Q.A.
AND DISTRIBUTED.
8/28 NRC RESIDENT REVIEWED Q1P FILE MATERIAL.
8/29 NRC REGION I MEETING WITH NMPC SENIOR MANAGEMENT
( REGARDING Q1P CONCERN.
8/31 V.P.-Q.A. LETTER TO MANAGER LICENSING AND RESPONSE RECEIVED ON REPORTABILITY AND IMPACT ON CERTIFICATION OF COMPLETION OF UNIT 2.
9/1 V.P.-Q.A. ISSUED RESPONSE SUPPLEMENT LETTER TO Q1P.
9/1 NM PRESIDENT SIGNED LETTER TO MR. KANE ON Q1P ACTIVITIES.
9/2 LETTER TO MR. KANE RECEIVED AT REGION I.
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- RaviDion 1 9/2/86 ATTACHMENT
.( yactorm Formina Ramis for Conclusion (Ref. Concerns 86-00064)
- 1. The Quality First Program investigation showed that each concernee and every other person interviewed specifically stated that they knew of no hardware problem which was not supported for correction by QC supervision.
- 2. All items identified were determined not reportable per 10 CFR 50.55(e).
- 3. QC inspection activity is overchecked by many specific programs.
- a. Audits - Since June, 1985, 12 audits have been performed that are test and hardware oriented.
- b. Surveillances - Since January, 1986, approximately 700 surveillances have been performed of testing and quality control activities.
- c. There are reviews of completed inspection documents performed routinely.
I
- ' d. Quality Engineering reviews DR's and WR's to verify that any required inspections were performed consistent with the work description.
I e. Some specific components received functional tests.
- f. Preoperational/ Acceptance tests were often used to verify the individual components / equipment integrity on a system basis following inspection activities.
- g. Work documentation such as DR's routinely require
" retest" of some conditions following completion of work and inspection activities.
- h. Start Up personnel are certified per ANSI N45.2.6 to perform testing functions.
- i. Component and system acceptance is based on recorded data and evaluated by a certified Level III Test Engineer.
1
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NQA OPS.~SECTION LOWER % POSITIVE RESPONSE NMPC NUCLEAR OUESTIONNAIRE ITEMS 9M1 2H1
4.
NUCLEAR DIVISION GIVES PROPER ATTENTION TO QUALITY / SAFETY X
- RELATED ISSUES.
- 21. Q1P PROVIDES ADEQUATE CONFIDENTIALITY. X i
- 7. OPEN AND HONEST COMMUNICATIONS.REGARDING QUALITY AND X
) SAFETY CONCERNS.
J 20. IF SUPERVISOR DOES NOT RESOLVE CONCERN, WOULD USE Q1P. X i
- 33. DURING OUTAGES RELUCTANT TO REPORT CONCERNS TO NMPC.
j 24. Q1P ADMINISTERED PROPERLY. X
- 12. MY SUPERVISOR SUPPORTED BY BOSS TO RESOLVE PROBLEMS. X 1 14. IF SUPERVISOR DOES NOT RESOLVE CONCERN, FEEL FREE TO GO X s
TO OTHER MANAGEMENT LEVELS.
i
- 23. I FEEL PEER PRESSURE IF I USE QlP. X
- 11. MY SUPERVISOR GIVES'ME FEEDBACK ON RESOLUTION OF PROBLEMS. X
- 13. MY SUPERVISOR RESPONSIVE TO SAFETY / QUALITY RELATED X IDEAS.
- 10. WHEN CALLED TO SUPERVISOR'S ATTENTION, SUPERVISOR TAKES X ACTION ON SAFETY / QUALITY ISSUES.
- 8. MY SUPERVISOR ENCOURAGES IDENTIFICATION OF PROBLEMS. X l
O1P CONCERN 86-00064C
( HiVESTIGATION REPORT PARAGRAPH V.P. Q.A. RESPONSE OF 8/27/86 BUrT.FT NO.
- 1. - DOMINATION BY START UP 1, 2 , 3
- 2. IANGUAGE NOT MENTIONED *
- 3. SYSTEM COMPLEX / CONFUSING 1, 2, 3, 5
- 4. POOR COMMUNICATIONS 1, 2, 3, 4, 5
- NOTE: V.P.-Q.A. LETTER DATED 9/1/86 TO MANAGER Q1P - SUBJECT, RESPONSE SUPPLEMENT, ADDRESSES THIS SPECIFIC ISSUE.
(
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