IR 05000219/1997003
| ML20210S004 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek, Three Mile Island |
| Issue date: | 08/20/1997 |
| From: | Miller H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| References | |
| 50-219-97-03, 50-219-97-3, 50-289-97-06, 50-289-97-6, CAL-1-97-08, CAL-1-97-8, NUDOCS 9709040306 | |
| Download: ML20210S004 (2) | |
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, August 20, 1997 Mr. Michael :Vice President and Director, OC GPU Nuclear Corporation Oyster Creek Nuclear Generating Station '
P.O. Box-388 Forked River, New Jersey 08731 SUBJECT: RESPONSE TO CONFIRMATORY ACTION LETTER 1-97-008, DATED MARCH 4,1997
Dear Mr. Roche:
This acknowledges receipt of your letter, dated April 30,1997, ret,ponding to actions -
1 and 2 requested in our Confirmatory Action Letter 1 97-008. In addition to reviewing
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reports and operability determinations, as described in NRC Integrated Inspection Report No. 50 219/97 03. The inspector found that the licensee's corrective actions were acceptable.
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Actions 3 and 4 requested in our Confirmatory Action Letter were addressed at a management meeting held on July 16,1997. The purpose of the meeting was to discuss the GPU Nuclear evaluation of the assessment of the equipment downgrading process, as documented in NRC Integrated Inspection Report No. 50-289/97 06.
Our inspection and meeting, along with your letter,is sufficient basis for us to conclude-you have satisfied the requested actions contained in our Confirmatory Action Letter.
Sincerely
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W Hubert J. Miller Regional Administrator Docket No.: 50-219
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Mr. Michael cc:
G. Busch, Manager, Site Licensing, Oyster Creek-M. Leggart, Manager, Corporate Licensing State of New Jersey Distribution:
Region 1 Docket Room (with concurrences)
Nuclear Safety Information Center (NSIC)
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NRC Resident inspector PUBLIC H. Miller, RA W. Axelson, DRA P. Eselgroth, DRP D. Haverkamp, DRP T. Kenny,DRS A. Keatley, DRP M. Oprendek, DRP
- B. Fewell, ORA W. Dean, OEDO P. Milano, NRR/PD l 3 R. Eaton, NRR/PD 1-3 R. Correla, NRR F. Talbot, NRR -
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Inspection Program Branch, NRR (IPAS)
D. Holody ORA DOCUMENT NAME: A: CAL 97008,0C Ti,essive a copy of thee document, Indicate in the bos: "C' = Copy without ettechment/ enclosure "E' = Copy with ettachment/ enclosure
"N" = No copy-l0FFICE RI/DRP , ;l c Rl/Qlg lE Rl/DRP . I Rl/@t#( l I
_ l NAME, DHAVERKAMPDf2// PKMdiROTH CHEhfituW - HMIfLER lDATE 08/18/97 08/ /9 /97 08/ W /97 08/ W /97 08/ /97 0FFICIAL-RECORD COPY
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t KING oF PRUSSIA. PENNSYLVANIA 19406-1416 March 4, 1997 License No. DPR 16 License No. DPR 50 CAL No.197 008 Mr. T. Gary Broughton President and Chief Executive Officer GPU Nuclear Corporation 1 Upper Pond Road Parsippany, NJ 07054
Dear Mr. Broughton:
SUBJECT: CONFIRMATORY ACTION LETTER 197 008 During an inspection covering the period January 6 to March 2,1997, NRC inspectors questioned the quality classification of selected plant components at Three Mile Island (TMI) Nuclear Station Unit 1. Subsequently, following licensee review, the NRC .
determined that numerous safety related components were improperly downgraded from the " nuclear safety related" classification to a lower tier classification without appropriate safety evaluations or other supporting engineering documentation. These equipment classification downgrades were initiated by GPU Nuclear and affected both TMI and Oyster Creek f acilities.
The NRC is concerned about the potential implications of the inappropriate equipment classification downgndes. Moreover, we are concerned about the poor implementation of the component classification process, as well as related weaknesses in procedure j adherence and communications. We are also concerned about the ineffective oversight of i the process by management, especially related to not taking prompt action to evaluate and resolve program problems identified by your own quality assurance activities. Because of the considerable extent of the process weaknesses, we also question the broader implications of these problems for other engineering processes.
Accordingly, pursuant to a telephone conversation between Mr. Arthur Rone and others of your staff and Mssrs. Eugene Kelly and Peter Eselgroth of my staff, on March 3,1997, it is our understanding that you would take severalimmediate and long term corrective actions.
Specifically, you committed to:
1. Immediately take steps to preclude additional inappropriate instances of equipment classification downgrades by; a. stopping. work on further programmatic equipment classification downgrades pending procedure and training upgrades, and
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T. G. Broughton 2 b. establishing a process to review planned day to-day maintenance and modification work involving equipment that was re-classified to assure that parts of the correct quality classification components are used.
By April 30,1997; 2. Determine the impact of the equipment classification downgrade program, as implemented, at TMt and Oyster Creek.
You should report the completion of this item by letter addressed to the Regional Administrator, NRC Region I, dated on or before April 30,1997.
Bv July 1.1997:
3. Perform an assessment, involving outside contractors and at least one outside member of your General Office Review Board, of the equipment classification downgrading process to determine:
a. the root causes of the weaknesses associated with the estabilshment and implementation of the process and appropriate corrective actions, and b. why past quality assurance findings in this area were not addressed in a timely manner.
4. Based on the results of your assessment, determine what assessments and changes to other engineering processes are needed.
You should be prepared to discuss your evaluation of the assessment in a management meeting with Region I staff on or before July 18,1997.
Pursuant to Section 182 of the Atomic Energy Act,42 U.S.C. 2232, and 10 CFR 2.204, you are required to:
1. Notify me immediately if your understanding differs from that set forth above.
2. Notify me if for any reason you cannot complete the actions within the specified schedule and advise me in writing of your modified schedule in advance of the change.
3. Notify me in writing when you have completed the actions addressed in this Confirmatory Action Letter.
Issuance of this Confirmatory Antion Letter does not preclude issuance of an Order formalizing the above commitmer.ts or requiring other actions on the part of the licensee.
Nor does it preclude the NRC tam taking enforcement action for violations of NRC requirements that may have prompted the issuance of this letter, in addition, failure to take the actions addressed in this Confirmatory Action Letter may result in enforcement action.
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T. G. Broughton 3 The responso directed by this letter are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub.L. 96 511.
In accordance with 10 CFR 2.700 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Docum nt Room.
Sincerely, l bort J. Miller egional Administrator Docket Nos. 50 219, 50 289 License Nos. DPR 16, DPR 50 cc:
M. J. Cross, Acting Vice President and Director TMl E. L. Blake, Shaw, Pittman, Potts and Trowbridge (Legal Counsel for GPUN)
Commonwealth of Pennsylvania J. C. Fornicola, Director, Licensing and Regulatory Aff airs TMI Alert (TMIA)
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J. S. Wetmore, Manager, TMI Regulatory Affairs M. B. Roche, Vice President and Director Oyster Creek G. Busch, Manager, Site Licensing, Oyster Creek M. Laggart, Manager, Corporate Licensing State of New Jersey
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6730 97-2135 April 30,1997 Mr. Hubert J. Miller, Administrator US Nuclear Regulatory Commission '
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Region I 475 AllendaleRoad King of Prussia, PA 19406 1415
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Subject: Oyster Creek Nuclear Generating Station (OCNOS)
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Facility License No. DPR-16 Docket No. 50-219 Response to Itent 2 of Confirmatory Action Letter (CAL) 1-97-008 Re: Quality Classification of Selected Plant Components at OCNGS
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Dear Mr. Miller:
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Item 2 in CAL l-97-008 follows:
"2. Determine the impact of the equipment classification downgrade program, as implemented, at TMI and Oyster Creek.
You should report the completion of this item by letter addressed to the Regional Administrator, NRC Region I, dated on or before April 30,1997."
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This letter responds to this specific item for OCNGS.
Immediate corTective action was taken in response to the issues raised concerning the quality classi6 cation list (QCL). Steps were taken to preclude additional inappropriate instances of equipment classification downgrades. Work on fbrther programmatic equipment classification
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downgrades was stopped pending procedure and training upgrades and a process was established to review planned day-to-day maintenance and modiScation werk involving equipment that was
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downgraded in order to assure that parts of the correct quality classl6 cation are used.
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Page 2 of 3 Downgraded equipment can be olaced in the following three categories: 1) items initially classl6ed as "nudear safety related" (NSR) and downgraded to " regulatory required" (RR) or "Other with QA" or "Other without QA"; 2) items initially dasal6ed as RR and downgraded to "Other with QA"; and 3) items initially classised as RR and downgraded to "Other without QA".
For the category of 605 items initially classlSed as NSR, item lists were posted and utilized to prevent unreviewsd activities until safety reviews could be conducted which con 6rmed the downgraded clasalEcation or nSirmed the original NSR dassl6 cation. Safety reviews on all these
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Items have been completed. During the review, the Oyster Creek Deviation Report corrective action process was used to address operability concems. As a result,16 items esquired re-dasel6 cation to NSR. The impact of the equipment classi6 cation downgrade propam on previously cleaslSed NSR ltems at Oyster Creek was detamined by evaluating the operability of -
- the speci6c components. Operability reviews were completed and concluded that none of the inappropriately classlSed NSR components a8boted operability.
All 1530 components that were downgraded kom RR to "Other with QA" had their classi6 cation retumed to RR. The impact of this downgrade was minimized for the following reasons.
Ahhough the el-ification was downgraded, materials and parts for the downgraded components were not programmatically downgraded and most of the parts remained at the RR classi6 cation.
For components in the preventive maintenance program, a review revealed that 15 of 1136 parts required material upgrading. All parts in this category had procedural requirements for ,
installation, testing, maintenance or surveillance. Existing maintenance piccedures required a check of new parts against installed parts and required post maintenance testing. Based on the above reviews, OPU Nuclear condudes that there was no adverse impact on equipment.
All 449 components downgraded kom RR to "Other without QA" were evaluated,- The evaluation concluded that 294 be returned to RR and the rest remain as "Other without QA". The impact of the downgrade on these 194 items was minimised for similar reasons to those downgraded to "Other with QA" from Rlt Materials and parts for the downgraded components were not programmatically downgraded and most remained at the RR classi6 cation. Existing maintenance procedures required a check of new parts against installed parts and required post maintenance testing. Based on the above reviews, OPU Nuclear condudes that there was no adverse impact on equipment.
GPU Nuclear took action to ensure that QCL denciendes with a potential to impact plant safety were addressed. For the components with the greatest potential impact on safety (NSR), a review of the deviations resulting kom the safety reviews indicated no a8 bot on the operability of those components. Most materials and parts for downgraded RR components remained at the RR classl6 cation. This, combined with the controls employed through maintenance procedures, provides reasonable assurance that repaired components are operable and reliable.
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6730 97<2135 '
. Page 3 of 3 Based on the above, OPU Nuclear concludes that, although de6clencies existed in the QCL, appropriate action was taken to minimize any affect on safe plant operation including the implementation of corrective action.
Ifyou should have any questions concerning the information in this letter, please contact Mr. Paul Caaya, Regulatory Affairs Department, at 609 9714139.
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Very truly yours, ;
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Michael hk '
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j Vice President and Director
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c: USNRC Document Control Desk Oyster Creek NRC Senior Resident inspector Oyster CreekNRC Project Many i
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July 3,1997 EA 97 296 Mr. Michael Vice President and Director 1 GPU Nuclear incorporated !
Oyster Creek Nuclear Generating Station P.O. Box 388 l Forked River, New Jersey 08731 SUBJECT: NRC INTEGRATED INSPECTION REPORT NO. 50 219/97 03, NOTICE OF !
VIOLATION
Dear Mr. Roche:
On May 25,1997, the NRC completed an integrated inspection at your Oyster Creek reactor facility. The enclosed report presents the results of that inspection.
During the six week period covered by this Inspection period, your conduct of activities at the Oyster Creek f acility was generally characterized by safety conscious operations, sound engineering and maintenance practices, and careful radiological work controls, However, we are concerned about several violations that occurred recently that were-caused by human performance problems and poor communications and which represent continuing examples of an adverse trend that past efforts have been ineffective in correcting. Some of the violation examples, as well as concerns identified while restoring an inoperable control rod to service, indicate a marked decline in performance related to verbal and written communications, in order to further discuss your current and planned corrective actions for the adverse trend in human performance, a management meeting has been scheduled for July 11,1997, as discussed between Mr. Sander Levin (GPU) and Mr.
Peter Eselgroth of my staff.
In addition, a separate violation was identified regarding the f ailure of your staff to complete a required 10 CFR 50,59 safety evaluation for a known sing'e failure vulnerability within the standby gas treatment system, which was first identified by-your staff in 1984
~ (UFSAR updated in December 1989). Your staff's current position regarding the need to perform a 10 CFR 50.59 safety evaluation when an original design configuration is subsequently found to not meet all assumed design criterin is inconsistent with the NRC's position that such a configuration necessitates a 10 CFR 50.59 safety evaluation. Please address this apparent inconsistency in your response to this violation.
These violations are cited in the enclosed Notice of Violation, and the circumstances surrounding the violations are described in detail in the enclosed report. Please note that you are required to respond to this letter and should follow the Instructions specified in the enclosed Notice when preparing your response. The NRC will use ynur response, in part,
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to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements, in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room (PDR).
We appreciate your cooperation.
Slocerely, Original Signed By: l Charles W. Hehl, Director Division of Reactor Projects Docket / License: 60 219/DPR 10 72 1004 Enclosures'
1. Notice of Violation I 2. NRC Inspection Report No. 60 219/97 03 I cc w/enci:
G. Busch, Manager, Site Regulatory Affairs, Oyster Creek M. Laggart, Manager, Corporate Regulatory Aff airs
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State of New Jersey-
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Ennineering
- Overall, licensee response to the identification and follow up for " slow" control rod start of motion times due to scram solenoid pl lot valve (SSPV) problems were very Good. However, the initial response to the test results was poor (untimely), and was due to inadequate communications between operations and engineering personnel. This resulted in a delay in evaluating the control rod test results. Once the actions were initiated, they were conducted conservatively and appropriately.
Engineering evaluation of the observed problems during the installation of the new SSPV diaphragms were thorough and probing, resulting in the further identification of an additional safety significant pilot valve seat problem. Operations and maintenance personnel effectively processed and implemented a large amount of tagouts and related work activities for this bl 0hly visible issue in an error free manner. (E1.1)
- The licensco's efforts in identifyin0 not positive suction head concerns with the core spray system while evaluating a desi0n chan00 were conservative and thorou0h.
However, additionalinformetion is needed (licensco to complete an ongoing evaluation) to determine whether this discovery represents a condition in conflict l with the UFSAR, and is an unresolved item. (E1.2)
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A system engineer appropriately evaluated an elevated temperature indication for an electromatic relief valve to determine the impact on valve operability. (E1.3)
- The licensco did not perform a 10 CFR 50.59 ovaluation when the UFSAR was updated in December 1989 to reflect that the original design of the standby Das treatment system was vulnerable to a single failure. The vulnerability was not reviewed and assessed during the ori0 inal plant license process. When the licensee identified this single f ailure vulnerability in 1984, and subsequently incorporated in the UFSAR, the licensee did not review the implication of this on the technical specification basis. Although this failure to perform a 10 CFR 50.59 cvaluation occurred in December 1989,it is reflective of the licensee's current performance because the licensee's approach to this type of change to the UFSAR remained the same. The licensee has not taken actions to resolve this specific issue, and the f ailure to perform a 10 CFR 50.59 evaluation is a violation. (E2.1)
Problems identified with the implementation of the component classification process at Three Mile Island prompted a similar review for the common process at Oyster Creek. The licensee's actions taken to date concerning the Quality Classification List pro 0rammatic weaknesses were acceptable. Additionallicensee findings will be evaluated as they are identified, and further tracking and evaluation of this issue is an unresolved item. (E2.2)
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E2.1 Standby Gas Treatment System Single Failure (Violation 50 219/9 7 03-04) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 E2.2 Quality Classification List and Component Downgrade Program Deficiencies (Unresolved item 50 219/97 03 05) . . . . . . . . . . . . 27 E8 Miscellaneous Engineering issues (NCV 50 219/97 03 01e) . . . . . . . . . 29 EB.1 (Closed) eel 50 219/97 02 04: . ...................... 29 IV. PLANT SUPPORT (71707, 71750,92904) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 R1 Radiological Protection and Chemistry Controls . . . . . . . . . . . . . . . . . . 29
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R 1,1 - General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 l S1- Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . 29 l
' S1.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 S2 Status of Security Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . 30
S2.1 Protected Area (PA) Detection Aids . . . . . . . . . . . . . . . . . . . . . 30 S2.2._ Alarm Stations and Communications . . . . . . . . . . . . . . . . . . . .- 31 -- !
S2.3 Testing, Maintenance and Compensatory Measures . . . . . . . . . . 31 !
SS Security and Safeguards Staff Training and Qualification (T&O) . . . . . . 32 i
S6 Security Organization and Adrninistration . . . . . . . . . . . . . . . . . . . . . . 32 1 S7 Quality Assurance in Security and Safeguards Activities ........... 33 S7.1 Etfectiveness of Management Controls . . . . . . . . . . . . . . . . . . . 33 S8 Miscellaneous Security and Saf ety issues . . . . . . . . . . . . . . . . . . . . . . 33 S8.1 Ve hicle Barrier System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 S8.2 Bomb Blast Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 S8.3 Procedural Controls . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . 35 S8.4 Security Computer and Compensatory Measures . . . . . . . . . . . . 35 S8,5 Review of a Previously Identified Violation . . . . . . . . . . . .... 36 V. M ANAG EMENT M EETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 X1 Exit Meeting Summary . . . . . . ............................ 30 X2 Review of Updated Final Safety Analysis Report (UFSAR) . . . . . . . . . . . 30
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ATT A C H M E N T 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 ATT A C H M E NT 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 ATTAC H M E N T 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 vil
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NRC inspectors noted this vulnerabilRy in licensee documentation in 1989. The NRC findings were documented in inspection reports 50 219/89 00and 89 09, and, at the time, noted that the vulnerability of the system was not reflected in the UFSAR.
Durin0 this inspection, the inspector requested the documentation which ovaluated the change to the UFSAR. The licensco indicated that the single falluto vulnerability was original design, and a 10 CFR 50.59 ovaluation would not be required, since they were correcting or adding information in the UFSAR. However, the inspector questioned if the identification of this original dosion vulnerability was a chango to the system as described in the UFSAR, and would require a written 10 CFR 50.59 ovaluation to provido the basis for the actormination that this chan00 did not involve on unroviewed safety question.
The inspector also considered this chan00 to bo in conflict with the TS basis for primary containment intourity (TS 4.5), which indicated that the limits for containment leakage were based on 90% charcoal filter officiency. However, calculations performed by the licensco indicated that the single f ailuro vulnerability could reduce filter officiency to 78E The licensco indicated that they did not consider the 90% SGTS officiency in the TS basis to be in conflict with the sinolo f ailure vulnerability of the SGTS in the UFSAR, because the TS basis reflected the desi0n capability of the system rather the system performanco considorin0 a postulated failuro. The inspector concludod that the licensoo's position with roGard to TS basis could bo incorrect, and that the TS basis, specifically the basis for primary containment leakago limits, must consider the limitin0 single f ailuro vulnerability in order to assure public health and safety, c. ConclusiQ nt The licensoo did not perform a 10 CFR 50.59 ovaluation when the UFSAR was updated to reflect the original design of the SGTS to be vulnerable to a singlo ,
f ailure. Although this was the original design, it may not have been reviewed during I the originallicenso as havin0 a single f ailuro vulnerability. As a result, when this single f ailuro vulnerability was identified in 1984, and subsequently incorporated in the UFSAR, the licensoo did not review the implication of this on the TS basis.
Although this f ailure to perform a 10 CFR 50.59 ovaluation occurred in December 1989,it is reflective of the licensee's current performanco because the licensco's approach to this type of change to the UFSAR remained the samo, and no actions have been taken to resolve this specific issue. The failure to perform a 10 CFR 50.59 ovaluation is a violation, fVIO 50 219/97 03 04)
EL2 Quality Classification List anj Component Downarade Pronram Deficiencies (Unresolved item 50 219/97 03-05)
a. Insnection Scoco (37551,40500)
The inspector reviewed the licensee's quality classification list (OCL) in response to
, programmatic questions identified at Three Mile Island (TMI), which are also
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applicable to Oyster Creek. The inspector participated in a telephone conference
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call between GPUN and the NRC, interviewed engineering and operations personnel, I and revlewed selected documentation (e.g., safety evaluations, operability determinations, deviation reports), l i
b. Observations and Findinos
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A conference call was held on March 3,1997, to discuss to OCL component
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affected, and discussions included actions for both TMl and Oyster Crook. Tho OCL
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issue was rotated to downgrading the equipment classification of important
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components and/or_ systems without processing the appropriato evaluations or other supporting engineering documentailon.
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The NRC issued a Confirmatory Action Letter (CAL) to GPUN on March 4,1997, to document the NRC's understanding of the immediato and long term corrective ,
actions to which GPUN committed in order to resolve the OCL downgrado issues.
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l By lotter dated April 30,1997, GPUN responded to one of the items listed in the
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CAL, which was to determino the impact of the equipment classification downgrado program, as implemented, at TMl and Oyster Creek. At Oyster Creek, a process 1 was imptomented which reviewed planned daily maintenance and modification
, activities involving equipment that was downgraded in order to assure that the parts l
- of the correct quality classification were used. '
Oyster Creek described the several categories in which prior OCL downgrades were l performed. For the 605 items initially classified as " nuclear safety related" (NSR)
and subsequently downgraded to " regulatory required" (RR) or "Other," a list of those items was generated to provent work on those components until safety reviews were completed to either 1) confirm the adequacy of the downgrado, or 2)
restore the original NSR classification. The licensee's deviation report system, a corrective action process, was used to evaluate operability concerns. As a result of !
the safety reviews,16 (of the 605) required re classification to NSR.
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- The inspector reviewed selected safety reviews, deviation reports and operability determinations, . Overall, the inspector found that the licensee's reviews were
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acceptable. In some instances, the inspector questioned specific aspects of the .
above documentation. There were no current operability questions identiflod. '
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There are additional actions to be completed by the licensee as per the c commitments documented in_ the CAL.. These actions will be addressed upon
licensee submittal and is an unresolved item. (URI 50 219/97 03 05)
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[ The licensee's corrective actions taken to date concerning the OCL programmatic
, weakness were acceptable. Additionallicensee findings will be evaluated as'they are identified. Further tracking of this issue will be accomplished via an unresolved -
item.
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