IR 05000334/1998080
| ML20249B639 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/18/1998 |
| From: | Meyer G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Cross J DUQUESNE LIGHT CO. |
| References | |
| 50-334-98-80, 50-412-98-80, NUDOCS 9806240004 | |
| Download: ML20249B639 (2) | |
Text
June 18, 1998
SUBJECT:
INTEGRATED INSPECTION 50-334/98-80, 50-412/98-80
Dear Mr. Cross:
- This letter refers..to your April 8,1998 correspondence, in response to our March 9,1998 letter.
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- Thank you for informing us of the corrective and preventive actions documented in your letter. These actions will be examined during a future inspection of your licensed program.
Your cooperation with us is appreciated.
Sincerely, Glenn W. Meyer, Chief Civil, Mecharleal and Materials Engineering Branch Division of Reactor Safety Docket Nos.: 50-33a; 50-412 cc w/o cv of Licensee Resnonse Letter:
Sushil C. Jain, Senior Vice President, Nuclear Services Group R. Brandt, Vice President, Nuclear Operations Group and Plant Manager R. LeGrand, Vice President, Operations Support Group B. Tuite, General Manager, Nuclear Operations Unit W. Kline, Manager, Nuclear Engineering Department M. Pergar, Acting Manager, Quality Services Unit J. Arias, Director, Safety & Licensing Department g
J. Macdonald, Manager, System and Performance Engineering
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ec w/cv of Licensee Resoonse Letter:
. M. Clancy, Mayor, Shippingport, PA
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Commonwealth of Pennsylvania-State of Ohio State of West Virginia 230038 P
9906240004 990618 PDR ADOCK 0S000334.
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Mr. J. Distribution w/cv of Licensee Resoonse Letter:
. Region i Docket Room (with concurrences)
Nuclear Safety information Center (NSIC)
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NRC Reeldent inspector H. Miller, RA/W. Axelson, DRA M. Evans, DRP N. Perry, DRP D. Haverkamp, DRP C. O'Daniell, DRP B. McCabe, OEDO R. Capra, PD1-2, NRR D. Brinkman, PDI 2, NRR V. Nerses, PDI-2, NRR R. Correia,' NRR F. Talbot, NRR DOCDESK Inspection Program Branch, NRR (IPAS)
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DRS fil e DOCUMENT NAME: G:\\ BRANCH 7\\REPLYLTR\\bv9880.rpy T 3,eceiwo a cow of this document, Indcate in the bou! 'C' = Copy without ettschment/ enclosure
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NAME DHaverkamp tW PKaufman(
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DATE 4 /IT/98 f AG98 hV LW98 NW y
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g Be ve Valley Power Station
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Shippingport, PA 15077 0004 or i e President Fan (724) 643 8 69 Nuclear Power Division April 8, 1998 L-98-066 U. S. Nuclear Regulatory Conunission Attention: Document Control Desk Washington, DC 20555-0001 Subject:
Beaver Valley Power Station, Unit No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Integrated Inspection Report 50-334/98-80 and 50-412/98-80 Reply to Notice of Violation In response to NRC correspondence dated March 9,1998, and in accordance with 10 CFR 2.201, the attached reply addresses the Notice of Violation transmitted with the subject inspection report.
If there are any questions concerning this response, please contact Mr. J. Arias at (412) 393-5203.
Sincerely, Sushil C. Jain Attachment c:
Mr. D. S. Brinkman, Sr. Project Manager Mr. D. M. Kern, Sr. Resident Inspector Mr. H. J. Miller, NRC Region I Administrator Mr. J. T. Wiggins, Director, Division of Reactor Safety, Region 1 Mr. N. S. Perry, Acting Chief, Reactor Projects Branch No. 7, Region 1 DEllVERING.
QUALliY ENERGY
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DUQUESNE LIGHT COMPANY
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Nuclear Power Division Beaver Valley Power Station, Unit No. I and No. 2 Repiv to Notice of Violation.
Integrated Inspection Report 50-334/98-80 and 50-412/98-80 Letter Dated March 9,1998 VIOLATION 1 (Severity Level IV, Supplement I)
Description of Violation (50-334/98-80-02)
10 CFR 50, Appendix B, Criterion XVI (Corrective Action), requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, conditions adverse to quality were not promptly corrected in that from January 1995 to January 23,1998, Duquesne Light Company failed to promptly correct defective equipment; i.e., residual heat removal (RHR) system flow control valve MOV-RH-758, which had been identified via a maintenance work request in January 1995 to have excessive leakage (more than 1000 gallons per minute).
Reason for the Violation Recurrent problems with excessive leakage past MOV-RH-758 have been documented since 1981. Since that time, adjustments were made to the valve's actuator to minimize the leakby when the valve is in the closed position. However, these efforts were ineffective. The recurrent problem with this valve was inappropriately accepted, since L
the reactor plant component cooling (CCR) system return flow from the residual heat
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removal (RHR) heat exchangers was able to be throttled to compensate for the leakby deficiency on MOV-RH-758. However, this change to the CCR system configuration was not adequately evaluated or proceduralized until recently.
Throttling of these CCR system valves without procedural guidance or safety evaluation
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went unquestioned in the past due to a perception that throttling of butterfly valves to control cooling flow to system heat exchangers was an acceptable practice. Also, in this l
specific instance, the valves that were throttled were in series with similar valves that were designated as throttle valves by the Operating Manual. The designated throttle valves were not used due to accessibility concerns.
Prior to the initiation of the Condition Report Program in January 1997, the corrective action program was weak and fragmented.
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Reply to N tice of Vi:lation
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Corrective Actions Taken and Results Achieved -
l 1. Condition Report 972281 was issued on December 12, 1997, and identified the problem.with excessive leakage past MOV-RH-758 and the actions being taken'to compensate for this deficiency. Condition Report 980242 was issued on February 6, 1998,' and identified that there were no operating procedures or safety evaluations that addressed the throttling of valves CCR-251 and CCR-252 to compensate for the excessive leakage past MOV-RH-758.
2. Temporary Operating Procedure (TOP) 1-97-27 was performed on December 16, 1997, to determine the minimum flow position on MOV-RH-758.
The TOP determined that flow through MOV-RH-758 was at its minimum at 0% indication on the valve's controller. However, even at this position, excessive leakage through the valve occurs.
3. Ultrasonic examinations of the valve body and piping downstream of MOV-RH-758 were performed 'on February 5,1998, to determine if piping erosion was occurring due to the excessive leakby on MOV-RH-758. No indication of erosion was detected by these examinations.
4. Ultrasonic examinations of the piping downstream of th.; reactor plant component cooling water (CCR) valves CCR-251 and CCR-252 were performed on February 6,
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1998, to determine if piping erosion was occurring due to the throttling of these valves. No indication of erosion was detected by these examinations.
5. A safety evaluation and operating manual procedure revisions were prepared to review and proceduralize the practice of throttling CCR valves CCR-251 and CCR-252 as a means to compensate for the excessive leakage past MOV-RH-758. The safety evaluation was reviewed by the Onsite Safety Committee (OSC) on February 9,1998, and the operating manual procedure revisions were incorporated on February 11,1998.
6. A Basis for Continued Operation (BCO) was prepared that evaluated the effects of the degraded condition on MOV-RH-758 and the compensatory measures of
. throttling CCR-251 and CCR-252. This BCO was reviewed and approved by the OSC and the Nuclear Safety Review Board (NSRB) on February 27,1998.
7. The problem of excessive leakage past MOV-RH-758 was added to the operator workaround list on March 20,1998.
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Inspection Report 50-334/98-80 and 50-412/98-80 Page 3 Corrective Actions to Prevent Further Violations l-1. Design Change Request 3348 was written on February 16,1998, to initiate actions to l
designate a replacement for MOV-RH-758.
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~ 2. Further actions to attempt to minimize the excessive leakage past MOV-RH-758 will l
be performed prior to achieving Mode 4 from the current outage at Unit 1.
3. Depending on the results of the above efforts, repair or replacement of M6V-RH-758
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will be performed during the fuel-out window of the next refueling outage,1R13.
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This outage is currently scheduled to begin in March 1999. Note that due to the current extended outage at Unit 1, the scheduled start date of IR13 is likely to be delayed.-
4. A critique was conducted to provide a thorough review of the circumstances that delayed the implementation of corrective actions, and the factors that contributed to accepting the flow control deficiencies with MOV-RH-758. A critique report will be issued by April 29,1998. This report will identify the contributing factors which led I
to the failure of the plant staff to establish and maintain an environment that places a high priority on preserving a system's design configuration. The issues identified by this critique report will be documented by condition report and adhaed in accordance with the Condition Report Program.
5. An extent of condition review for the condition identified in CR 980242 to identify potentially similar issues will be completed by June 1,1998. Any similar issues involving the inappropriate throttling of valves to compensate for equipment deficiencies which are identified by this review will be documented by condition report and addressed in accordance with the Condition Report Program.
Date When Full Compliance Will Be Achieved Full compliance will be achieved with the restoration of MOV-RH-758 to its design configuration. This will be completed by the end of the IR13 refueling outage which is currently scheduled to begin in March 1999. Note that due to the current extended outage at Unit 1, the scheduled start date of IR13 is likely to be delayed.
The critique report for the MOV-RH-758 issue will be issued by April 29,1998. Issues identified in the report will be addressed via the Condition Report Program.
The. extent of condition review for CR 980242'will be completed by June 1,1998.
Similar issues identified by this review will be addressed via the Condition Report Program.
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' Reply to Notice of Vi::lation Inspection Repod 50-334/98-80 and 50-412/98-80 t
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VIOLATION 2 (Severity Level IV, Supplement I)
Description of Violation (50-334/98-80-04)
10 CFR 50.59, " Changes, tests and experiments," in part, permits the licensee to make changes to its facility and procedures as described in the; safety analysis report and conduct tests or experiments not described in the safety analysis report without prior Commission approval, provided the change does not involve a change in the technical'
specifications or an Unreviewed Safety Question (USQ). The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USQ.
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Section 9.3 of the Beaver Valley Unit 1 UFSAR states that "the cooldown rate of the reactor coolant is controlled by regulating the reactor coolant flow through the tube side of the RHR heat exchangers."
Contrary to the above, on December 18,1997, Duquesne Light Company implemented a change to temporary operating procedure (1 TOP-97-28) to allow operators to control the cooldown and heatup rate of the reactor coolant (a procedure described in the safety analysis report) by a means other than by regulating reactor coolant flow through the tube side of the RHR heat exchangers and no safety evaluation was performed.
Specifically, the change allowed operators to regulate reactor coolant cooldown and heatup rates by throttling the component cooling water flow through the shell side of the RHR heat exchangers.
Reason for the Violation-The reason for the violation is considered to be improper use of the station non-intent review program. Operating Manual Change Notice (OMCN) 1-97-884 was made on December 18, 1997, to ITOP-97-28 to allow throttling of the reactor plant component cooling water (CCR) valves to control RHR temperature. However, this change was incorrectly processed as non-intent, based on existing procedure 10M-10.4.C, "RHR Shutdown," which allows for a brief period of throttling the CCR valves while the system is being cooled to ambient conditions. Because the desired steps for CCR valve throttling were found in this approved procedure, a more thorough UFSAR review was not perfonned.
Corrective Actions Taken and Results Achieved 1. A safety evaluation and operating manual procedure revisions were prepared to allow throttling of valves CCR-251 and CCR-252 as a means to compensate for the
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Page5 excessive leakage past MOV-RH-758. The safety evaluation was reviewed by the-OSC on February 9,1998, and the operating manual procedure revisions to IOM-10.4.A,B,C were incorporated on February 11,1998.
2. Condition Report 980240 was initiated to review the concern of the same individual completing the two review sign-offs for OMCN 1-97-884 which implemented the change to ITOP-97-28 on December 18,1997. As corrective action to this concern, a random sample of 751 OMCNs were reviewed to identify the extent of condition.
Three similar concerns were identified from this review. The original OMCNs for these cases were corrected and documented with the appropriate signatures.
E Corrective Actions to Prevent Further Violations 1. Nuclear Power Division Administrative Procedure 2.3, " Procedure Review and Approval," (NPDAP 2.3) was revised effective February 3,1998.
The revision incorporated specific guidance to address the definition of non-intent changes and specific guidance associated with the review and determination process for non-intent changes.
Specifically, this revision added performance of a 10 CFR 50.59 -
Applicability screen for all non-intent changes to safety related procedures concurrently with the non-intent change.
2. The guidance in NPDAP 2.3 will be further augmented to require a 50.59 Applicability screen for all non-intent procedure changes.
A 10 CFR 50.59 Applicability screen has and will continue to be performed for intent procedure changes. In order to allow adequate time to process the revision to NPDAP 2.3, this revision to NPDAP 2.3 will be effective by June 15,1998.
3. A review of the technical specification, UFSAR and BVPS administrative requirements for necessary signatures related to the review and approval of OMCNs will be reviewed as part of Module 3 oflicensed operator retraming program (LRT).
This training is scheduled to be completed by June 30,1998.
Date When Full Comoliance Will Be Achieved Full compliance with respect to the inadequate review of the procedure change was achieved with the completion of the safety evaluation and procedure revisions associated
. with the throttling of valves CCR-251 and CCR-252. These actions were completed by February 11,1998.
' The revision to NPDAP 2.3 will be issued by June 15,1998.
The review of the requirements for the review and approval of OMCNs will be discussed in Module 3 of LRT, which is scheduled to be completed by June 30,1998.
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' Reply to N:tice of Vi:lation
. Inspection Report 50-334/98-80 and 50-412/98-80
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Page 6 VIOLATION 3 (Severity Level IV, Supplement I)
Description of Violation (50-334(412)/98-80-03)
10 CFR 50.59, " Changes, tests and experiments," in part, permits the licensee to make changes to its facility and procedures as described in the safety analysis report without prior Commission approval, provided the change does not involve a change in the technical specifications or.an unreviewed safety question (USQ). The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USQ.
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Contrary to the above, in December 1994 Duquesne Light Company made a change to
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allow storage of propane next to the auxiliary intake structure building without performing a safety evaluation that was adequate to determine that the change did not
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involve a USQ. Specifically, the modification added three 1000 gallon propane storage tanks adjacent to the auxiliary river intake structure, which contains systems that provide the altemate heat sink requirements for the Beaver Valley units. (The Unit I auxiliary river water system and the Unit 2 standby service water system, which are described in Unit 1 UFSAR, Section 9.16, and Unit 2 UFSAR, Section 9.4, are located inside the auxiliary river intake structure.) The safety analysis was deficient, in that it did not evaluate the potential hazards or impact on equipment associated with the storage of large quantities of liquid propane next to the auxiliary river intake structure.
Additionally, instead of three 1000 gallon propane tanks, Duquesne Light Company actually stored 4000 gallons of propane in a tanker truck adjacent to the auxiliary intake structure building without performing an evaluation of the larger amount or the change in storage methods. Without these evaluations, there was insufficient bases to determine that the change did not involve a USQ.
Reason for the Violation The safety evaluation did not adequately address the potential hazards associated with the storage oflarge quantities ofliquid propane next to the alternate intake structure due to a lack of attention to details. In addition, the non-safety related nature of this design change (DCP 2133) and the alternate intake structure contributed to lack of depth and quality of the safety evaluation for the design change.
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i Page 7 Corrective Action Taken and Results Achieved 1.. A revised safety evaluation for DCP 2133 was approved on February 18,1998, which thoroughly evaluated the potential hazards of storing large quantities of liquid
. propane next to the alternate intake structure. This safety evaluation also considered the use of a tanker truck as the method of storage, including the storage of up to 4000 gallons ofliquid propane.
2. Since the time of the initial safety evaluation for DCP 2133 in December 1994, the l
quality and depth of safety evaluations has improved at Beaver Valley through the use of feedback offered during OSC reviews and by improved annual requalification training for ' safety evaluation preparers. It is now recognized that, regardless of the
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safety significance, safety evaluations must be robust and fully address the applicable technical specifications and the Updated Final Safety Analysis Report.
3. Design Change Package (DCP) 2133 was processed via controlling procedure NEAP 2.19, " Minor Design Change Control." Since January 1,1995, NEAP 2.19 has not been used for new design changes. Design changes are now processed via NEAP 2.2, " Design Change Control," regardless of the scope or complexity of the change. Since DCP 2133 was initiated prior to January 1,1995, it and other active minor DCPs were allowed to proceed to completion under NEAP 2.19.
NEAP 2.2 is a highly prescriptive, formatted approach of checklists and questions designed to address the major issues of design change control. This use of checklists is a major departure from NEAP 2.19 and offers a documented " checks and balance" system of process control.
4. Beginning in mid 1997, Safety Culture training was provided to engineering personnel to foster a questioning attitude.
Corrective Actions to Prevent Further Violations 1. A self-assessment will be conducted of design changes initiated in the time frame of December 1994 to June 1997. The self-assessment sample will include active Minor Design Change Packages as well as standard Design Change Packages. Emphasis of the review will be upon the scope of the safety evaluation versus the actual design scope and die quality of the safety evaluation. The self-assessment will be completed by June 30,1998.
2. Beginning in 1998,' Safety Culture training is being provided to operations and maintenance personnel to foster a questioning attitude. This training is expected to
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be completed by December 31,1998.
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Reply to Notice of Violation
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Inspection Report 50-334/98-80 and 50-412/98-80
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Date When Full Comoliance Will Be Achieved
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Full compliance was achieved with the approval of the revised safety evaluation for DCP 2133 on February 18,1998.
The self-assessment of design changes discussed above will be completed by June 30, 1998.
Safety Culture training for operations and maintenance personnel is expected'to be completed by December 31,1998.
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