IR 05000219/1997003
| ML20210S004 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek, Crane |
| 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) | |
Text
August 20, 1997
SUBJECT:
RESPONSE TO CONFIRMATORY ACTION LETTER 1-97-008, DATED MARCH 4,1997
Dear Mr. Roche:
March 4, 1997
SUBJECT:
NRC INTEGRATED INSPECTION REPORT NO. 50 219/97 03, NOTICE OF
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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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Mr. Michael to determine whether further enforcement action is necessary to ensure compliance with
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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'
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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 hly 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 desi n 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)
l 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 ori 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 R 1,1 - General Observations............................... 29
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l S1-Conduct of Security and Safeguards Activities.................. 29 l
S1.1 General Observations................................ 29
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S2 Status of Security Facilities and Equipment..................... 30 S2.1 Protected Area (PA) Detection Aids..................... 30
S2.2._ Alarm Stations and Communications....................- 31
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S2.3 Testing, Maintenance and Compensatory Measures.......... 31
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SS Security and Safeguards Staff Training and Qualification (T&O)...... 32 S6 Security Organization and Adrninistration...................... 32 i
S7 Quality Assurance in Security and Safeguards Activities
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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...........
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M ANAG EMENT M EETING S...................................... 36 X1 Exit Meeting Summary......
............................ 30 X2 Review of Updated Final Safety Analysis Report (UFSAR)........... 30 ATT A C H M E N T 1....................................... 38
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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 desi n 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.
Conclusi nt Q
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
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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)
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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),
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Observations and Findinos A conference call was held on March 3,1997, to discuss to OCL component
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downgrade concerns that woro initially identified at TMI. Oyster Crook was similarly
affected, and discussions included actions for both TMl and Oyster Crook. Tho OCL 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
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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
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actions to which GPUN committed in order to resolve the OCL downgrado issues.
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
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program, as implemented, at TMl and Oyster Creek. At Oyster Creek, a process
was imptomented which reviewed planned daily maintenance and modification activities involving equipment that was downgraded in order to assure that the parts l
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of the correct quality classification were used.
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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
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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 acceptable. In some instances, the inspector questioned specific aspects of the
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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 commitments documented in_ the CAL.. These actions will be addressed upon c
licensee submittal and is an unresolved item. (URI 50 219/97 03 05)
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Conelysigna l
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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
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are identified. Further tracking of this issue will be accomplished via an unresolved -
item.
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