ML20217B378
| ML20217B378 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/14/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20217B369 | List: |
| References | |
| 50-271-97-10, EA-97-531, NUDOCS 9804230024 | |
| Download: ML20217B378 (9) | |
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s ENCLOSURE NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF civil PENALTY l
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Vermont Yankee Nuclear Power Corporation Docket Nc. 50-271 Vermont Yankee Nuclear Power Station License No. DPR-28 EA 97-531 l
During an NRC inspection con 6 toted between September 29,1997, and November 20,1997, for which exits meetings warc held on October 3,1997, and November 20,1997, violations l
of NRC requirements were identified. In accordance with the " General Statement of Policy l
and Procedure for NRC Enforcement Actions," NUREG-1600, the NRC proposes to impose a civil penalty pursuant to Section 2.34 of the Atomic Energy Act of 1954, as amended (Act),
l 42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:
1 1.
VIOLATIONS ASSOCIATED WITH TORUS TEMPERATURE A.
10 CFR Part 50, Appendix B, Criterion Ill, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements j
and the design basis, as defined in 50.2 and as specified in the license l
application, for structures, systems, and components, are correctly translated i
into specifications, drawings, procedures and instructions and that measures l
shall be established for the identification and control of design interfaces and l
for coordination among participating design organizations.
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Contrary to the above, in 1982, the licensee failed to control the design interfaces and failed to assure that the design basis for the maximum torus temperature normal operating limit was correctly translated into specifications.
Specifically, the analyses to support a 1982 Technical Specification (TS) license amendment request (to increase the normal torus water temperature limit from 90*F to 100*F) did not consider the impact of this change on design basis analyses such as the emergency core cooling system (ECCS) pump net positive suction head (NPSH) margin calculations, loss of coolant accident (LOCA) containment analyses, ECCS piping stress and support load calculations, and equipment qualification. An initial torus temperature of 90*F was assumed in these analyses. (01013f 2This violation occurred beyond the five year statute of limitations period for assessing civil penalties; therefore, this violation was not considered for purposes of determining any civil penalty.
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Vermont Yankee Nuclear Power Corporation 2 B.
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, I
such as failures, malfunctions, deficiencies, deviations, defective material and l
equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that i
the cause of the condition is determined and corrective action taken to preclude i
repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.
Contrary to the above, between May 1994, and November 20,1997, the licensee failed to properly evaluate and correct, in a timely manner, an identified condition adverse to quality involving the discrepancy between the design basis and the TS limit for maximum normal torus temperature (described in Section I. A). Specifically, in May 1994, the licensee identified that the TS limit of 100 F for maximum torus normal operating temperature (in place as of June 6, i
1985, when TS Amendment 88 was issued) was not consistent with assumptions made in the Final Safety Analysis Report (FSAR) description of LOCA containment response, and did not properly evaluate and correct this condition adverse to quality in a timely manner. Specifically:
The analyses to support a safety evaluation pursuant to 10 CFR 50.59 were not initiated until November 1995. The safety evaluation and operability determination were not completed until April 8,1996.
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The nonconformance between the FSAR and the TS was not entered
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into the licensee's corrective action process until November 2,1995.
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Administrative controls to limit normal torus temperature to 90 F, consistent with the design basis, were not established until December 1,1995.
The residual heat removal (RHR) system operating procedure was not revised to reflect the 90 F administrative limit until April 1997.
On March 26,1996, the licensee initiated an internal event report that questioned the ability to technically justify plant operation with a torus temperature in excess of 90 F because of concerns associated with the core decay heat model, as well as the need to consider energy introduction into the containment from continued injection of feedwater.
Within one hour, the licensee reported the condition to the NRC as potential operation outside of the design basis pursuant to 10 CFR 50.72; however, the licensee only reviewed two years of plant operating logs to determine if torus temperature had actually exceeded 90 F.
Since no instances were identified, the licensee incorrectly concluded OFFICIAL RECORD COPY l
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Vermont Yankee Nuclear Power Corporation 3 that the condition was not reportable as a LER, pursuant to 10 CFR 50.73.
A comprehensive review of plant operating logs was not performed until May 29,1997.
Although the assessment of the concerns identified in March 1996, determined that maximum torus temperature, as a result of a LOCA, was acceptable only if an initial torus temperature of 90oF was assumed, the i
l licensee had not requested a TS change as of November 20,1997.
(01023)
C.
10 CFR 50.73(a)(2) requires, in part, that licensees shall submit a Licensee Event Report (LER) within 30 days after the discovery of the event, for any event or condition that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant.
Contrary to the above, between June 28 and November 20,1997, the licensee failed to report a condition that resulted in operation outside the design basis of the plant. Specifically, on May 29,1997, the licensee discovered that the plant had operated with torus temperature above 90 F for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in two instances between 1985 and 1995 (a two-day period in August 1988, and an eight day period in 1993); however, as of November 20,1997, the licensee had not reported the condition to the NRC, a period in excess of 30 days. This condition was outside of the design basi.s of the plant in that an initial torus temperature of 90 F was used in design basis analyses for ECCS pump NPSH margin calculations, LOCA containment response, ECCS piping stress and support load calculations, and equipment qualification. Additionally, when continued feedwater injection was considered in the LOCA containment analysis, the peak torus water temperature was acceptable only if an initial torus temperr.ture of 90*F was assumed. (01033)
These violations in the aggregate constitute a Severity Level lll problem (Supplement 1).
Civil Penalty - $55,000.
II.
VIOLATIONS NOT ASSESSED A civil PENALTY A.
10 CFR Part 50, Appendix B, Criterion Ill, Design Control, requires, in part, that l
measures shall be established to assure that applicable regulatory requirements i
and the design basis, as defined in 50.2 and as specified in the license l
application, for structures, systems, and components, are correctly translated
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into specifications, drawings, procedures and instructions and that measures j
shall be established for the selection and review of materials, parts, equipment, i
and processes that are essential to the safety-related functions of the structures, systems and components. Criterion ill also requires that measures 1
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shall be established for the identification and control of design interfaces and for coordination among participating design organizations and that design control measures shall provide for verifying or checking the adequacy of design.
1.
Contrary to the above, prior to November 20,1997, the licensee failed to correctly translate the manufacturer's recommendations for RHR motor starting limits into operating instructions considering the expected RHR motor ambient temperature. Specifically, the limit for consecutive pump starts (three in five minutes) specified in the RHR operating procedure was based on a maximum ambient room temperature of 86
- F.
However, the maximum normal operating temperature for the RHR corner room is 109'F and the room temperature may reach as high as 155
- F during accidents. (02014)
This is a Severity Level IV violation (Supplement 1).
2.
Contrary to the above, prior to May 9,1997, the licensee failed to correctly select equipment in a subsystem essential to the safety-related function of the emergency diesel generators (EDGs). Specifically, air to the solenoid valves that operated the EDG service water cooling flow control valves (FCVs) was supplied from a nonsafety-related pressure regulator. Failure of the pressure regulator could have resulted in a malfunction of the solenoid valve which could have prevented the FCVs from opening. The failure of the flow control valve could cause a loss of all service water to the EDG which would prevent operation of the EDG. (03014)
This is a Severity Level IV violation (Supplement I).
3.
Contrary to the above, prior to November 20,1997, the licensee failed to correctly translate RHR flow specifications into procedures.
Specifically, the flow limitations specified in RHR operating procedure for minimum pump flow requirements did not consider instrument uncertainty in the specified limit. The RHR procedure included a I
minimum flow precaution of 2700 gpm.
However, considering uncertainty of the flow instrument, an indicated flow rate of 3920 gpm was required to ensure that the vendor recommended minimum flow of 2700 gpm was established. (04014)
This is a Severity Level IV violation (Supplement 1).
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8 Vermont Yankee Nuclear Power Corporation 5 4.
Contrary to the above, on December 6,1995, the licensee used incorrect design inputs in the calculation of NPSH nargin for the RHR pumps in calculation VYC-808, Rev. 2.
Specifically, the licensee used a curve fit of the vendor's pump test data in calculating the required NPSH values rather than actual test data. Use of the curve fit data resulted in a nonconservative NPSH required value. (05014)
This is a Severity Level IV violation (Supplement 1).
5.
Contrary to the above, prior to November 20,1997, the licensee failed i
to update the heat exchanger fouling assumption used in RHR service water (RHRSW) room cooler thermal performance calculations after an inspection of cooler unit coils in April 1995 indicated that the assumption was incorrect. Specifically, Calculation VYC-1329 was not changed to reflect micro-fouling as the likely cause of the fouling, rather than tube plugging due to silt, after no evidence of silt fouling was found during the inspection. (OtiO14)
This is a Severity Level IV violation (Supplement 1).
6.
Engineering Instruction WE-103, " Engineering Calculations and Analyses," Rev.15, dated October 14,1994, section 4.1.4.2, stated that, when information from quality assurance (QA) design records was required, the licensee must ensure that the appropriate (governing) documents were used and that such documents were the latest approved revision obtained from the appropriate source.
Contrary to the above, in April 1997, the licensee failed to assure correct references and inputs were used in design calculations.
Specifically:
Calculation VYC-1349, Rev.1, dated April 30,1997, referenced drawing G 191372, Rev. 41; however, engineering had approved i
Rev. 42 of the drawing G-191372 on December 20,1996.
Calculation VYC-298, Rev.10, dated April 22,1997, referenced various drawings as listed in Section 3.0.5 (a) through (I), which were superseded by a later revision before the licensee issued Calculation VYC 298, Rev.10. The latest revisions of the drawings indicated some de load changes. (07014)
This is a Severity Level IV violation (Supplement 1).
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i B.
10 CFR Part 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, malfunctions, deficiencies, deviations, defective material and i
equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that i
the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the i
cause of the condition, and the corrective action taken shall be documented and
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i reported to appropriate levels of management.
J 1.
Contrary to the above, between November 13,1986, and May 1997, the licensee failed to properly evaluate and correct the nonconformance between the vendor recommended RHR pump minimum flow requirement of 2700 gpm and the installed minimum flow capacity of j
350 gpm. Specifically, in a letter dated November 13,1986, the vendor j
notified the licensee that the minimum flow for the RHR pumps should be increased to 2700 gpm for continuous operation (more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l
of operation in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and to 2075 gpm for intermittent operation.
However, the licensee failed to adequately correct this nonconformance, l
despite prior opoortunities, namely:
The licensee's response, dated May 8,1989, to IE Bulletin 88-04, which requested licensees to determine (and correct) whether the installed minimum flow capacity was adequate for pump operation, lacked the technical basis to conclude that the existing RHR pump minimum flow would be adequate during postulated accident scenarios during which the pump would operate for several hours under minimum flow conditions. The licensee's response did not provide either verification from the vendor or test results to demonstrate that minimum flow rates were adequate during the postulated accident scenarios. The i
vendor was unable to support the licensee's assertion that a cumulative arithmetic series of minimum flow events over the life of the plant (29,200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) had the same relationship to pump degradation as the length of a specific event (4 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of minimum flow operation during an accident).
The licensee added a precaution to the RHR operating and surveillance procedures, in 1987, to minimize operation of the RHR pumps in the minimum flow mode. These instructions did not edequately reflect the vendor recommendations, provided in i
November 1986, to increase the minimum flow to 2700 gpm.
Although, the RHR operating procedure was revised in May 1997, and additional instructions were provided to reflect the OFFICIAL RECORD COPY a: PROP-VY.DSN
h4 Vermont Yankee Nuclear Power Corporation 7 vendor recommendations, these instructions did not reflect the recommendation, provided in May 1997, that RHR pump operation should not be sustair.ad at a flow rate of 350 gpm for more than 30 seconds during surveillance tests. (08014)
This is a Severity Level IV violation (Supplement 1).
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2.
Administrative Procedure (AP) 0009, " Event Reports," a measure established by the licensee to implement the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, that event reports be initiated for unanalyzed conditions or potential conditions outside the design basis, and that the event be reviewed against the requirements of the Basis for Maintaining Operation (BMO) Guideline to determine if j
a BMO is required.
The BMO Guideline states, in part, that if a safety-related component or system is degraded relative to the Final Safety Analysis Report (FSAR) or other licensing commitment, then a BMO should be prepared.
Contrary to the above, prior to November 20,1997, the licensee failed to take appropriate measures to assure that a condition adverse to l
quality involving the service water (SW) and emergency diesel generator l
(EDG) support systems was appropriately evaluated and corrected.
I Specifically, the SW supply line to the circulating water and SW traveling screens, SW piping to the diesel generators outside the diesel generator rooms in the turbine building, fuel oil transfer lines routed on the exterior of the pump house, and the diesel exhausts, were not adequately protected from the effects of tornadoes, including tornado missile strikes. This constituted a degraded condition relative to commitments in the Preliminary Design Assessment Report (PDAR); however, no BMO was prepared. (09014)
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This is a Severity Level IV violation (Supplement 1).
C.
10 CFR 50.73(a)(2) requires, in part, that licensees shall submit a Licensee l
Event Report (LER) within 30 days after the discovery of the event, for any j
event or condition that alone could have prevented the fulfillment of the safety
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function of systems that are needed to remove residual heat.
Contrary to the above, between June 6 and November 20,1997, the licensee failed to report a condition that alone could have prevented the fulfillment of the safety function of the RHR system, a system needed to remove residual heat.
Specifically, on June 6,1997, the NRC identified that instrument uncertainty j
was not included in the' specification of RHR pump minimum flow requirements in system procedures and as of November 20,1997, the licensee had not i
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submitted a LER, a period in excess of 30 days. The failure to provide adequate
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instructions for RHR pump minimum flow requirements could have rr,sulted in failure of the RHR pumps which would have prevented the fulfillment cf a safety function c,f a system required to remove residual heat. (10014)
I This is a Severity Level IV violation (Supplement 1).
Pursuant to the provisions of 10 CFR 2.201, Vermont Yankee Nuclear Power Corporation (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this J
Notice of Violation and Proposed imposition of Civil Penalty (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each alleged violation: (1) admission or denial of the alleged violation, (2) the reasons for the violation if i
admitted, and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved, (4) the corrective steps that will be taken to avoid further violations, and (5) the date when full compliance will be achieved. If an adequate reply is not received within the time specified in this Notice, an Order or a Demand for Information may be issued as why the license should not be modified, suspended, or revoked or why such other action as may be proper should not be taken. Consideration may be given to extending the response time for good cause shown. Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.
Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commissic,n, with a check, draft, money order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or the cumulative amount of the civil penalties if more than one civil penalty is proposed, or may protest imposition of the civil penalty in whole or in part, by a written answer addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission. Should the Licensee fail to answer within the time specified, an order imposing the civil penalty will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty, in whole or in part, such answer should be clearly marked as an " Answer to a Notice of Violation" and may: (1) deny the violation (s) listed in this Notice, in whole or in part, (2) demonstrate extenuating circumstances, (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed. In addition to protesting the civil penalty in whole or in part, such answer may request remission or mitigation of the penalty.
In requesting mitigation of the proposed penalty, the factors addressed in Section VI.B.2 of the Enforcement Policy should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The attention of the Licensee is directed to the other provisions of 10 CFR 2.205, regarding the procedure for imposing a civil penalty.
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I Upon failure to pay any civil penalty due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be i
collected by civil action pursuant to Section 234c of the Act,42 U.S.C. 2282c.
The response noted above (Reply to Notice of Violation, letter with payment of civil penalty, and Answer to a Notice of Violation) should be addressed to: J. Lieberman, Director, Office l
of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555 Rockville Pike, Rockville, MD 20852-2738,with a copy to the Regional Administrator, U.S. Nuclear
~ Regulatory Commission, Region I and a copy to the NRC Resident inspector at the f acility that is the subject of this Notice.
I Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Dated at King of Prussia, Pennsylvania this 14thday of April,1998 i
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