ML20206H072
| ML20206H072 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/15/1988 |
| From: | Morris K OMAHA PUBLIC POWER DISTRICT |
| To: | Lieberman J NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), NRC OFFICE OF ENFORCEMENT (OE) |
| References | |
| LIC-88-994, NUDOCS 8811230079 | |
| Download: ML20206H072 (10) | |
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Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 2247 402/536 4000 November 15, 1988 CERTIFIED Mall LIC 88-994 RETURN RECEIPT REQUESTED Hr. James Lieberman, Director Office of Enforcement U. S. Nuclear Regulatory Comission Attn: Document Control Desk Mail Station PI-137 Washington, DC 205P5
References:
1.
Docket No. 50-285 2.
Letter from NRC (R. D. Martin) to OPPD (R. L. Andrews) dated October 12, 1988
Dear Mr. Lieberman:
SUBJECT:
Response to Notice of Violation and Proposed Imposition of Civil Penalty (NRC Inspection Report 50 285/88-15 EA88-145)
Omaha Public Power District (OPPD) received the Notice of Violation and Proposed Imposition of Civil Penalty dated October 12, 1988.
The Notice of Violation involved two violations categorized in 'he aggregate as a Level II problem concerning the potential for a comon mode failure of the Safety injection and Refueling Water Tank (SIRWT) level controll.ecs, and a loss of containment integrity.
OPPD acknowledges the violations and does not contest the proposed civil penalty.
Accordingly, please find attached OPPD's response to the violations pursuant to 10 CFR *srt 2.201 and a check in the amount of $50,000.
OPPD identified both incidents, promptly reported them to the NRC and took l
corrective actions to resolve them.
In addition to the activities in place to address the specific events, a major undertaking has been initiated by 0 PPD to act leve excellence in all nuclear related activities.
The cornerstone of tlis effort was an independent and comprehensive appraisal of all nuclear related activities which was completed June 1988.
In response to the recomendations arising from this assessment a detailed action plan has been developed.
This action plan represents a major comitent by OPPD to improve all areas rel.ted to the operatior, :naintenance, engineering support, and management of the fort Calhoan Station.
This plan has been presented to the NRC and included implementation schedules.
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- J'ames Liebermann LIC-88-994 i
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OPPD is confident that these initiatifes will ensure the continued safe operation of the Fort Calhoun Station.
Pursuant to the provisions of Section i
182a of the Atomic Energy Act of 1954, as amended, this response is submitted j
under oath and affirmation.
If you have any questions concerning this matter, do not hesitato to contact us.
Sincerely,
.J.
Olvision Mar.ager 4
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Nuclear Operations l
Attachments KJM/sa c: LeBoeuf, Lamb Leiby & MacRae l
1333 New Hampshire Ave., N.W.
j Washington, DC 20036
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R. D. Martin, NRC Regional Administrator
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P. D. Milano, NRC Project Manager P. H. "arrell, NRC Senior Resident Inspector i
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0HNISSION In the Matter of
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Omaha Public Pc.<er District
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Docket No. 50-205 (Fcrt Calhoun Station
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Unit No. 1)
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AFFIDAVIT K. J. Morris, being duly sworn, hereby deposes and says that he is the Division Manager - Nuclear Operations of the Omaha Public Power District; that as such he is duly authorized te sign and file with the Nuclear Regulatory Commission the attached information concerning the response to Notice of Violation and Proposed Imposition of Civil Penalty (NRC Inspection Report 50-285/88-15 EA 88 145); that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information, and belief.
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D}i ivision Manager Nuclear Operations STATE OF NEBRA3Kr.
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ss COUNTY OF DOUGLAS)
Subscribed and sworn to before me, a Notary Public in and for the State of Nebraska on this ir day of November, 1988.
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t RESPONSE TO NOTICE OF VIOLATION i
During an NRC inspection conducted during i.he period April 6 through May 13, l
1988, violations of NRC req'Jirements were 1.ientified.
In accordance with the "General Statement of Policy and Pracedure for NRC Enforcement Actions," 10 CFR i
Part 2, Appendix C (1988), the Nuclear Regulatory Commission proposes to impose i
civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act), 42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:
A.
Safety In.iection and Refuelina Water Tank Level Switches 10 CFR Part 50, Appendix B, Criterion XI, Test Control, and the Fort Calhoun Station, Unit 1, Updated Final Safety Analysis Report, Appendix A, Approved Quality Assuranes Program, require a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed. Also, the test program shall include, as appropriate, preoperational and operational tests during plant operation.
Contrary to the above, as of April 1988, the licensee had not established an acequate program to assure that certain check valves were tested to demonstrate these valves would perfore satisfactorily in service.
Specifically, check valves associated with the air-operated Safety /, C/,
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Injection and Refueling Water SIRW) tank low level switches (A/, B and D/FIC-383) were not tested (during operation to assure that they would perform their intended safety function under certain ar.cident conditions.
When tested in April 1988, these check valves were found to leak excessively.
Violations A and B have been categorized in the ggregate as a Severity Level III problem (Supplement I).
Civil Penalty - $50,000 (assessed equally between the violations).
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M SPONSE TO VIOLATION A 1.
Admission or Denial of the A11eaed Violation OPPD admits the violation occurred as stated.
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The Reasons for the Violation. If Admitted 1
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Prior L the 1987 water intrusion of the Fort Calhoun Station's Instrument i
i Air System, the Instrument Air System was viewed as an important auxiliary l
l system to support operations, however, the system is classified as a
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"nonsafety related system."
Therefore, tha follow.up actions in response i
to the 1985 SSONI Report first addressed items in "safety related" systems.
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In October 1987, is a result of the assessment of tho safety functions sup-f ported by the Instrument Air System, the Safety Injection and Refueling i
hater Tank (SIRWT) check valves were identified as a component requiring I
testing.
This assessment was con?ucted as one of the corrective actions 3
i following the September 1987 Diesel Generator (DG 2) exhaust damper j
failure.
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The SIRWT check valves were subsequently included in a revised IST program submitted to the NRC in December 1987.
Testing of these check valves was deferred at that time because:
1)
Previous testing of the Instrument Air System check valves performed during the 1987 outage and in the fall of 1987 provided confidence that the check valves would perform their intended function.
2)
Since no local isolations were prosided for, online testing would have required the isolation of an entire riser of air supply and affected other equipment on the riser, i
3)
The inadvertent initiation of engineered safeguards equipment was a possibility with onli a testing.
When the evaluation of the design basis and safety functions of the Instru-mont Air System accumulators was concluded on April 6, 1988, the conoquences of check valve failure and the decision to defer testing were reevaluated. A method to safely test the valves online was ascertained and testing of the check valves was performed on April 15, 1988. After the check valves failed to meet test acceptance criteria, it was discovered that the Crane Model 27 valve was an original design misapplication, in summary, the reasons the violation occurred are:
1)
Because the check valves were installed in a non:,afety grade system, I
the valves were not recognized as safety significant during the development of the testing program until a thorough review of the Instri.nent Air System was con'.ucted.
2)
The use of the Crant Model 27, horizontal lift check valve as the air accumulator isolation valve was a misapplication by the original i
system designer of a check v.1ve which was clearly identified for use in steam or water systems, and not air systems.
This error was not discovered t'efore 1987 becar.e of the original improper classifi-cation as nonsafety related Additionally, the original system design basis did not recegnize a i
requirement for testing or inspection nd, therefore, provisions were not made in the system configuration for testing or isolation of the check valves.
3.
The Corrective Steos that Have Been Taken and the Results Achieved l
1)
Equipment Changes The failed valves (Crane Model 27) were replaced with a different type e
check valve (NVPRO B 8C-1).
This work was completed on the same day as the check valves test failure, April 15, 1988.
The new check valves were bench tested to prove the proper functioning of the valve prior to installation, and the SIRWT Level System was returned to normal. A full functional test of the check valves and their l
associated accumulators will be completed prior to restart from the i
current refueling outage.
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A review was made by a walkdown of the Instrument Air System, to identify other Crane Model 27 valves with accumulators. A total of seventeen additional Crane Model 27 valves have been identified.
Sixteen additional valves were identified on nonsafety related systems. One valve was found on the accumulator for LCV-383-1 and LCV-383-2 which is safety related. At the time of testing, LCV-383-1 and LCV-383-2 were already on a temporary nitrogen backup system due to a change in the design requirements for these valves as determined by the Instrument Air System accumulators evaluation as reported in LER 88-009.
The Crane Model 27 valves that have been identified were originally classified as nonsafety related and were part of the i
original plant equipment. Modifications for the safety related valves will be completed prior to restart following the current refueling 4
outage. Modifications are planned for the nonsafety related valves.
These nonsafety related valves have been evaluated and they do nct present a safety concern.
t In order o facilitate functional testing of the SIRWT bubbler check 1
valves, the SIRWT level instrumentation system is being modified (MR FC-88-39) during the 1988 refueling outage. This will allow l
testing without isolation of the air riser.
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The abova modifications and testing will assure improved reliability 1
of the Instrument Air System and execution of the its associated j
safety functions.
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Program Upgrt. des I
i The SIRWT bubbler check valves were included in a revised IST program as a result of the Instrument Air System assessment, and submitted to the NRC in December 1987.
I A review of the valve operators which utilize check valves and air i
accumulators to perform their designed safe shutdown function was completed in April 1988. This review identified the safety related air check valves discussed above and verified their design i
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requirements.
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j A test program to p3rform functional testing of the safety related I
accumulators, including the associated check valves, is underway and will be completet prior to restart from the current refueling outage.
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In addition, these valves have bee.1 added to the surveillance test l
program.
Test procedures for periodic testing of the valves are being l
written.
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Management Reviews In order to arovide support in the review and analysis of the SIRWT I
bubbler chec( valve failure, the Management Investigative Safety Team i
(MIST) was initiated in accordance with Nuclear Production Division i
Procedure (NPD-QP-18) by the Di'ision Manager on April 28, 1988. The purpose was to collect and preserve event information, establish event documentation files, analyze event significance, identify root causes and recommer.d/ review corrective actions.
This team was composed of three managers and assisted by the Reactor Engineer and the Supervisor Reactor Performance Analysis.
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The MIST reviewed documents and conducted onsito interviews on both of the events. associated with Violations A and B.
A report, issued May 12, 1988, l
contained results, conclusions and recomendations from that t
investigation. Their report was reviewed by the Safety Audit and Review Committee at a July 13, 1988 meeting.
The major conclusions from this evaluation were presented to Region IV in the enforcement conference held in June 1988, and incorporated into the corrective actions taken.
f OPPD has completed a thorough review and upgrade of the Fort Calhoun Instrument Air System. This review included a System Functional Inspection and the design basis reconstitution.
This review, along with improvements in system maintenance, modification and evaluations, and testing scheduled i
to be completed prior to the end of the 1988 refueling outage, will ensure a reliable Instrument Air System, i
4.
The Corrective Steos that Will Be Taken to Avoid Further Violatig,na n
The corrective actions listed above will help avoid the potential for 4
future violations and equipment reliability problems ralated to Crane Model 27 check valves.
In May 1988, OPPD initiated an evaluation, the purpose of which is to enhance the existing Critical Quality Element (CQE) list, Surveillance Test and ISI programs of components in the Safety Injection System. A fault-tree based evaluation of the SI System has been developed to identify potentially safety reisted components. A preliminary e.aluation of the results of the SI fault-t"ee modeling will be availtble in January 1989.
This evaluation, along with the Design Basis Reconstitution project, and the Preventative Maintenance upgrade will ensure that a sufficient level of testing is conducted on plant equipment.
5.
The Date when Full Comoliance Will Be Achieved I
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OPPD is currently in full coa.pliance regarding the specific violation. The testing program is in place and the check valves and their associated accumulators will be functionally tested prior to restart from the current 4
l refueling outage.
Additional enhancements to the CQE identification process for the T,1 System is currently scheduled to be completed by June 1989.
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B. -Loss of Containment IntearitY 1.
Technical Specification 2.8, "Limiting Conditions for Operation-i Refueling Operations," requires certain conditions be satisfied during any refueling operations, including the requirement that all automatic containment isolation valves shall be operable or at least one valve in each line penetrating containment shall be closed.
Contrary to the above, refueling operations occurred during the period April 8-13, 1987, with a 3/8-inch containment pressure sensing instrument line Icft open to the outside atmosphere. This line had been open since March 27, 1987, when a maintenance technician perfonning Surveillance Test ST-CONT-3 removed and did not replace a test cap on this line. There is neither an automatic containment isolation valve for this instrument line nor was manual Isolation Valve A/IICV-742 closed such that no valve in the instrument line could be automatic. ally closed or was maintained closed to prevent an open path free the containment.
2.
Technical Specification 2.6(1).a. "Limited Condition for Operations-Containm'nt Systems," requires that containment integrity not be violated unless the reactor is in cold shutdown condition.
Technical Specification, Definitions-Containment Integrity, defines containment integrity, in part, by requiring that (5) The uncontrolled containment leakage satisfies Specification 3.5."
Technical Specification 3.5(4).c, ' Surveillance Requirement-Containment Tests," requires, in part, that "The combined leakage rate of all penetrations and valves subject to Type B and Type C tests shall be less than or equal to 0.6 La."
Contrary to the above, during the period May 28, 1987 to April 19, 1988, while the reactor was not in a cold shutdown condition, the leakage rate (as calculated by the licensee) through the uncapped containment pressure sensing instrument line connected to Penetration M-38, subject to Type C testing, during a design basis a cident would have exceeded 0.5 La by a factor of approximately seven (418,000 vs.
l 62,451 standard cubic centimeters per minute based on 0.6 La).
3.
Technical Specification 2.15 and associated Table 2-3, footnote f, t
"Limiting Condition for Operations - Instrument Operating Requirements for Engineered Safety features" require, with regard to the containment high pressure channels, that if one channel becomes t
inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability.
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Contrary to the above, when the test cap for the containment pressure instrument sensing line connected to Penetration M-38 was discovered missing on A>ril 18, 1988, at approximately 3:15 p.m. by licensee personnel, tie engineered safety features channel associated with redundant Instruments A/PC-742-) and A/PC-742-2 were placed in a tripped or bypassed condition.
The missing test cap was not replaced until approximately 9:45 a.m. on April 19, 1988.
RESPONSE TO VIOLATION B 1.
Admission or Denial of the A11eaed V1.olation OPPD admits the violation occurred as stated.
2.
Ihr reasons for the Violation if Admitted This event was identified by OPPD and reported to the NRC in Licensee Event Report No. 50 285/88-011 (LER 88 011) dated April 19, 1988.
The LER was revised May 12, 1988 (Rev 1), May 27, 1988 (Rev 2) and July 1, 1988 (Rev 3) to reflect status changes.
In response to this event an investigation was conducted. The Management investigative Safety Team (MIST) was activated by the Nuclear Production Division Manager to evaluate the event to determine root cause(s) and to assess the adequacy of corrective actions.
In accordance with existing procedures, the MIST consists of a team of experienced senior managers whom are provided the authority and resources to conduct a detailed investigation.
In addition, the Plant Review Committee Chairman requested the Human Performance Evaluation System (HPES) coordinator to conduct an investigation focused on the human related root causes of the event.
The MIST investigation concluded that inadequate procedures represented the primary root cause for the failure to reinstall the swagelok cap.
Failure to take immediate corrective action upon discovery was attributed to a t
misunderstanding between the engineer who identified the alssing cap and a supervisor.
This misunderstanding was related to the position of the isolation valve just upstream of the missing cap.
These factors resulted in a lack of aw3reness of the safety significance of the missing cap.
Environmental factors were also identified as a contributor; e.g., on the day the cap was left off a snow storm lead to a job shutdown.
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The Corrective Steos That Have Been Taken and Results Achieved
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i Containment integrity was restored at about 9:45 a.m. April 19, 1988 when the cap was replaced. The status of all other mechanical containment l
penetrations were also verified through documented and independently verified walkdowns as detailed in the LER 88 011.
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The Surveillance Test, "Containment Isolation Valves Leaka9i Rate Test, Type C." (ST CONT-3) has been upgraded to include detailed v awings of all test tees and requires procedural signoffs for removal and rtinstallation of the caps.
The Surveillance Test "Containment Isolation Valves Leakage Rate Test, Type B," (ST CONT-2) has also been reviewed and revised.
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Safety analysis calculations were conducted by the NSSS vendor assuming a design basis accident (DBA) with the test tee uncapped.
These calculations usedtheUpdatedSafetyAnalysisReport(USAR)assumptionswiththe exception that the Technical Specification leak rate of 0.1 percent was assumed rather than the 0.2 percent as stated in the USAR.
Calculations combined the design leakage from the containment and the leakage from this open penetration to calculate the site boundary doses over the first two (2)hoursoftheevent. The site boundary doses during the first two (2) hours of the event were calculated to be 258.6 rem to the thyroid and 6.4 rem whole body if a DBA had occurred.
This compares to the 10 CFR 100 limits of 300 rem to the thyroid and 25 rem whole body.
A training program for improved safety awareness has been developed and is being implemented as part of a major action )lan to upgrade all aspects of the operation and management of the Fort Calioun Station.
This action plan was developed in response to a comprehensive independent audit of the conduct of operations, maintenance and administration of the Plant.
4.
The Corrective Steps That Will be Taken to Avoid Further Violations As d? tailed in LER 88-11, a checklist for double verification of conts nment penetrations is currently being developed and will be implemected cr ur to the end of the refueling outage that is currently in progress.
This checklist will be used to verify containment integrity prior to restart from the current and future refueling outages or outages where containment integrity ma, have been compromised.
The training program for improved safety awareness has been initiated for Plant Review Committee members and implementation of the )rogram will be expanded to cover first line supervisors for the Fort Calloun Station i
during 1989. Another major upgrade is planned for training of supervisors j
to im) rove their communications skills and diagnostic capabilities, which will )e completed in 1989.
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5.
The Date When Full ComDliance _Will be Achieved OPPD has been in full compliance since April 19, 1988, when the test tee cap was replaced, t
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