ML20199C703
| ML20199C703 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/14/1997 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 1-1142, 50-346-97-09, 50-346-97-9, NUDOCS 9711200100 | |
| Download: ML20199C703 (6) | |
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%01 N State Route 2 419-249-2300 John K. Wood Oak Hertre,oH 43449 FAX: 419-321-8337 Vce P. escort Nuclear Daws-Besse Docket Number 50-346 License Number NPF-3 Serial Number 1-1142 Noveinbe r 14, 1997 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555-0001
Subject:
Response to Inspection Report Number 50-346/97009 Ladies and Gentlemen:
Toledo Edison (TE) has received Inspection Report (IR) Number 50-346/97009 (TE Log Number 1-3903) and the two enclosed Notices of Violation which require a response. After discussion with the Senior Resident Inspector and Resident Inspector for the Nuclear Regulatory Commission at the Davis Besse Nuclear Power Station on October 2, and November 7,1997, it was agreed that the response to IR97009 would be submitted by November 14,1997. The responses to the violations contained in IR97009 are provided in Attachment I to this letter.
Should you have any questions or require additional information, please contact Mr. James L.
Freels, Manager - Regulatory Affairs, at (419) 321-8466.
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O' ifm[j Very truly yours,
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aj enclosure cc:
A. B. Beach, F.egional Administrator, NRC Region III S. J, Campbe11, DB-1 NRC Senior Resident Inspector A. G. Hansen, DB 1 NRC/NRR Project Manager Utility Radiological Safety Board-
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PDR ADLCK 05000346 G
Docket Number 50-366
'Licanda Numb 0r NPP-3
- Scrick Numbar 1-1142 Attachmset l' P ge 1 Renly to a No,Lipe of Violation ~ (50-346/97009-01[DRPl)
Alleced Vio1%cion I.
10 CFR Part 50, Appendix B, Criterion V,
" Instructions, Procedures, and Drawings," requires, in part, that " Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings."
Technical Specification 6.5.3.1.b specifies, in part, that " Temporary approval of changes to plant procedures.
.which clearly do not change the intent of the cpproved procedures, can be made by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license."
Contrary to the above, on August 15, 1997, the NRC identified that Section 6.8,3,c of administrative procedure DB-OP-00000, " Conduct of Operations,"
inappropriately allowed the Shift Supervisor to authorize,
'f needed, the performance of procedural steps out of the order specified by system operating procedures without the review and approval of a second member of plant management staff.
This is a Severity Level IV violation (Supplement 1).
Egapon for the Violation Prior to 1988, guidance on the use of procedures contained in the Davis-Besse Nuclear Power Station (DBNPS) Operations Section administrative procedure, stated that procedures are to be followed step-by-step.
In late 1988, during the development o' a new procedure, " Conduct of Operations" (DB-OP-00000),
guidance was in:orporated into DB-OP-00000 that allowed performance of system operating procedures out of sequence as long as the procedure intent was not compromised.
This was considered acceptable guidance for the manner in which procedures were to be implemented, as opposed to a change in a plant procedure
-that required temporary approval in accordance with Techaical Specification (TS) 6.5.3.1.b.
The procedure preparer and reviewers, responsible for developing DB-OP-0000, which was approved in November, 1988, did not properly interpret the impact on the TS administrative requirements for minor procedure changes. The process that allowed the Shift Supervisor to authorize
-performance of procedure steps out of sequence, that did not compromise the procedure intert, was not challenged because the process was seldom used.
Corrective Steos Taken and Results Achieved Upon being identified as a concern by a representative of the Nuclear Regulatory Commission (NRC), a Potential Condition Adverse to Quality Report (PCAQR) was initiated. A Standing Order (97-008) was issued on August 19, 1957, which implemented a review by a second Senior Reactor Operator, in addition to the current requirements of DB-OP-00000, for deviations in system i
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operating procedure step sequence. This interim requirement will remain in effect until the concern identified on the PCAQR is resolved.
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Docket Number 50-346 License Numbar NPF-3 Serial Number 1-1142 P ge-2
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-Corrective Steos that will be Tak n to Avoid Egrther violations 9
The staff of the DBNPS recognizes the importance of procedure adherence and had previously initiated action to develop a-new Nuclear Grot,p procedure, which is the highest tier of procedural guidance at the DBNPS, to promote the highest standard for procedure compliance issues.
The new procedure, NG-DB-00225, " Procedure Use and Adherence," will incorporate specific guidance and requirements for deviating from the sequence of steps in a procedure.
Guidance contained in NG-DB-00225 will be reviewed for compliance with TG 6.5.3.1.b and will be approved in accordance with NG-NA-00115, " Control of Procedures." Administrative procedure NG-DB-00225 will be approved and implemented by December 19, 1997.
Date When Full Comoliance Will Be Achieved Full compliance was achieved on August 19, 1997, when Standing Order 97-008 was issued, which specified requirements in compliance with TS 6.5.3.1.b for deviating from a system operating procedure step sequence.
Penly to a Notice of Violation (50- 3 4 6 /97009-04 IDRP I)
Alleoed Violation II. 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Act.on," requires, d
in part, that " Measures be estabitshed to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are properly identified and corrected."
Contrary to the above, between May 1996 and June 1997, interim measures were not taken to aid operator response actions if a worst case Circulating Water line break were to occur. This situation existed until June 1997, when it was identified by the NRC, even thouga a probabilistic safety assessment, issued in May 1996, identified that operations personnel would not have had sufficient time to respond to a postulated worst case Circulating Water system linebreak before the associated flooding in the turbine building caused a completed loss of feedwater.
This is a Severity Level IV violation (Supplenunt 1).
Reason for the Violation
.The Updated Safety Analysis Report (USAR) for the DBNPS, as a result of Questions and Answers asked during initial licensing, described the capability of pressure switches on each circulating water (CW) pump train to detect a complete rupture'of the main condenser CW expansion joint on the inlet side of the condenser.
Level switches in the condenser pit sump were credited to notify the control room operator, via a computer alarm, in the event of a high sump level. This computer alarm was-considered satisfactory for breaks of a sufficiently small size However, the available operator response time was
-dependent on the leak rate.
For a maximum size break (300,000 gpm flow), the
- available response time assuming the pressure switches on the condenser inlet pipe did not detect the break is approximately 5 minutes.
For small break i
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l Dockst Number 50-346 l
Licenso Numbsr NPF-3
- Seriel Number 1-1142 Page 3 sizes (less than 50,000 gpm flow), the available response time is greater than 30 minutes. Toledo Edison, in PCAQR 95-0849, initiated on October 4,
- 1995, identified a conaern that a spectrum of line breaks of an intermediate size could negatively impact the plant Probabilistic Risk Assessment (PRA) core damage probability. To mitigate this concern, several short term corrective actions were proposed and implemented. These included:
1.
Placing plant personnel in a safe location with ready communications to the Control Room during CW pump startup or shutdown and while pump discharge valve movements are taking place. This action was taken because pipe ruptures are considered more like)y during these operations.
2.
Additional operator training on the impact of a CW system rupture and the range of plant equipment that could be affected.
3.
Inclusion of the CW inlet pipe expansion joints in the preventative maintenance program. As a result of this action, all expansion joints were inspected and one 108 inch CW expansion joint was replaced.
4.
Modification of the CW pump discharge valves by replacing the stem-to-disc shear pin with a high strength pin.
5.
Adjustment of the CW pump discharge valve operators to increase the closing torque.
As a long term corrective action, a plant modification was initiated on October 30, 1995, to provide protection for the full spectrum of break sizes.
This modification proposed installation of new level switches and the appropriate logic to shutdown the CW train in the event of a CW pipe rupture.
In April, 1996, a probabilistic safety assessment (PSA) was completed that projected a core damage frequency (CDF) increase of twenty-five percent from the CDP calculated in the DBNPS Individual Plant Examination.
The PSA performed was considered conservative by TE in two respects.
1.
No credit was given for the existing pressure switches, which are capable of detecting a maximum size break at the condenser inlet and initiate trips to the associated CW pumps.
2.
No credit was given for the operators to trip the CW pumps from the control room for any break. The evaluations available at the cime showed that the existing design permitted tripping of the CW pumps from the control room for a spectrum of pipe breaks.
If these two conservatisms are removed from the calculation of the CDF, the CDF due to CW ruptures would have been increased by approximately 10 percent.
The industry guidance available at that time was EPRI Report TR-105396, "PSA Applications Guide," (August 1995).
This guidance when applied to the DBNPS, indicated that changes in the base line risk of up to 12 percent could be treated as non-risk significant. Based on these considerations and this industry guidance, Toledo Edison concluded that the level switch modification could be implemented no later than the eleventh refueling outage (11RFO) without any additional interim measurea.
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Dockst Number 50-346 i
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Licenzo Numbar NPF-3 I
- Scrial Numbar 1-1142 Page 4 Plant Modification (MOD) 95-0057 was initiated to install four level switches to trip the associated CW pump if a high level in detected in the condenser pit area, and to provide a new annunciator alarm in the control room to alert the operators of the flooding condition. This modification was also to increase the stroke time of the outlet valve to provide additional torque to close against the CW pump shutoff head.
Except for increasing the stroke time of the valve, the other portions of MOD 95-0057 were expected to be implemented during the operating cycle.
Actual preparation of the MOD was approved on July 10, 1996, for conceptual design to support implementation during cycle 11.
On September 11, 1996, the proposed design was approved for implementation during cycle 11.
The design was completed and the package was issued for implementation on December 6, 1996.
However, when the MOD was issued for implementation, the design package required electrical bus outages, which normally occur during a RFO, to implement the MOD.
Personnel responsible for implementation of the MOD requested changes to the design package so that the modification work could be performed during the operating cycle.
Subsequently, on February 4, 1997, a schedule change was approvoa to allow partial implementation during cycle 11, with final tie-in to the CW system during a potential Reactor Coolant Pump outage or during the 11RFO.
Scheduling decisions were influenced by the assessments contained in PCAQR 95-0849.
Although the CDF increase was discussed in PCAQR 95-0649, the discussion did not cause the MOD to be treated with a high priority.
A forced outage occurred in May, 1997, due to the failure of the Main Transformer.
Since MOD 95-0057 was not implemented during the forced outage, the Independent Safety Engineering unit generated a PCAQR (97-0802).
Re-evaluation of the impact on CDF followed, which included discussions with the NRC Senior Resident Inspector.
Although substantial interim corrective actione had been taken to reduce the probability of the event, the NRC Senior Resident Inspector expressed a concern that additional corrective actions had not yet been taken to reduce the consequences of the event.
As a result, additional compensating actions were taken which were assessed to have minimal positive effect on the CDF.
It was then determined that the MOD could be completed with the plant at power. The required portions of MOD 95-0057 were revised and implemented with the plant on line.
These actions were completed on August 2, 1997.
Toledo Edison has concluded that there were interdepartmental comuunication and teamwork weaknesses with regard to the assessment of an increase in CDF greater than considered non-risk significant within current industry guidelines. This resulted in installation of the new level switches being completed later than could have been achieved.
Safety assessment personnel were aware of the conservatism in the PSA, and as stated above some personnel.
had the perception that the actions already taken as part of PCAQR 45-0849 reduced the CDF increase to approximately 10 percent beyond the base CDF for the DBNPS.
Considering the conservatism of the analysis and the short term actions that were completed, schedaling decisions made for implementation of the MOD were considered to be acceptable.
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Docket Number 50-346 Liccnsa Number hPF-3 Scrial Numb 2r 1-1142 i
Page 5 Corrective Steos Taken av Results Achieved Implementation of plant MOD 95-0057 to install leval switches and the appropriate logic to initiate shutdown of the CW train in the event of a CW pipe rupture was completed on August 2, 1997.
Corrective Steps that will be Taken to Avoid Further Violations The procedure " Potential Condition Adverse to Quality Reporting,"
(NG-NA-00702), will be revised to ensure that a PCAQR identifying an increase in CDF beyond the non-risk significant criteria defined in EPRI Report TR-105396, "PSA Applications Guide," (August 1995), will be elevated to a category requiring management oversight. This change will be reviewed and discussed with Engineering & Services supervisors.
These actions will be completed by February 16, 1998.
The DBNPS monitors the development of guidance published by the NRC and the nuclear industry on the use of PRA and PSA techniques.
When changes in the use of these techniques are proposed, the proposed changes will be evaluated against the policies and procedures of the DBNPS for enhancement of the site policies and procedures.
Date When Full Compliance Will Be Achieved on August 2, 1997, when the new level switches and associated logic were installed, the concern was relieved over the possible negative impacts of a CW pipe rupture in the condenser pit.
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