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Matched Legal Cases: ['art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50']

NRC: EA-97-531 - Vermont Yankee (Vermont Yankee Nuclear Power Corp.)
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EA-97-531 - Vermont Yankee (Vermont Yankee Nuclear Power Corp.)
EA 97-531
Mr. Donald A. Reid
RD 5, Box 169
SUBJECT: NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY
- $55, 000
(NRC Inspection Report No. 50-271/97-10)
This letter refers to the subject NRC inspection conducted between September29 and November 20, 1997, at the Vermont Yankee Nuclear Power Stationin Brattleboro, Vermont, the findings of which were discussed with Mr.G. Maret and other members of your staff at exit meetings on October 3and November 20, 1997. The purpose of the inspection was to follow upon the findings of the NRC Architect/Engineer (A/E) team inspection conductedbetween May 5 and June 13, 1997. During the follow-up inspection, apparentviolations were identified associated with control of the design process,corrective actions for conditions adverse to quality, and reportability.The inspection reports for the A/E team inspection and the followup inspectionwere sent to you previously on August 27, 1997, and February 5, 1998,respectively. On March 2, 1998, a predecisional enforcement conference(conference) was conducted with Mr. R. Barkhurst, you, and other membersof the Vermont Yankee staff, to discuss the violations, their causes,and your corrective actions.
Based on the findings of the inspection and information provided duringthe conference, twelve violations are being cited and are described inthe enclosed Notice of Violation and Proposed Imposition of Civil Penalty(Notice). These violations involved a number of failures to: (1) properlytranslate the design basis of the plant into specifications, procedures,and instructions, contrary to 10 CFR Part 50, Appendix B, Criterion III;(2) promptly correct design deficiencies once they were identified, contraryto 10 CFR Part 50, Appendix B, Criterion XVI; and (3) report conditionsto the NRC in Licensee Event Reports (LERs), pursuant to 10 CFR 50.73.
The three most significant of these violations are set forth in sectionI of the Notice and relate to the Technical Specification (TS) limit formaximum normal torus operating temperature. In 1982, you submitted a TSlicense amendment request to increase the normal torus water temperaturelimit from 90F to 100F. The NRC approved the request in 1985; however,the analyses performed to support the change were inadequate in that theyfailed to consider the impact of the change on all of the affected designbasis analyses, namely, the emergency core cooling system (ECCS) pumpnet positive suction head (NPSH) margin calculations, loss of coolantaccident (LOCA) containment analyses, ECCS piping stress and support loadcalculations, and equipment qualification. Specifically, no evaluationwas performed to demonstrate that these previously performed analyseswere still acceptable assuming an initial torus temperature of 100F.
In May 1994, you identified that the TS limit of 100F for maximum torusnormal operating temperature (in place as of June 6, 1985, when a TS amendmentwas issued) was not consistent with assumptions made in the Final SafetyAnalysis Report (FSAR) description of LOCA containment response; however,you did not properly evaluate and correct this condition adverse to qualityin a timely manner. Specifically, you did not initiate formal analysesand calculations to support a safety evaluation of this condition untilNovember 1995, and did not complete the safety evaluation and operabilitydetermination until April 1996. More importantly, you did not put in placeinterim administrative controls to limit torus temperature to 90F, consistentwith the design basis, until December 1995. Although you identified, inMarch 1996, that you might not be able to justify plant operation witha torus temperature in excess of 90F, and later confirmed that you couldnot operate with torus temperature above this limit, you did not initiatethe necessary detailed design basis reviews until 1997, and, as of November1997, you had not requested a TS amendment to correct the nonconservativetorus temperature limit.
Although you identified, in March 1996, that operation of the plant withtorus temperature above 90F was a condition potentially outside of thedesign basis, you did not perform a comprehensive review to determineif torus temperature had actually exceeded 90F until May 1997, when questionedby the A/E inspection team. At that time, you identified that the planthad operated with torus temperature above 90F for greater than 24 hoursin two instances between 1985 and 1995; however, you did not report thiscondition to the NRC in an LER within 30 days, as required.
While no actual safety consequences resulted from operation of the plantwith torus temperature above 90F, if a LOCA had occurred, with an initialtorus temperature above 90F, there was not high confidence that the emergencycore cooling systems and the containment would have been able to performtheir safety functions. Additionally, the failure to take prompt actionto evaluate and correct an identified discrepancy between the FSAR andthe Technical Specifications is a significant concern because the NRCrelies upon licensees to operate the plant within the design basis andto promptly identify and report nonconforming conditions. These violationscollectively represent breakdowns in your processes for design control,corrective actions, and reportability; therefore, these violations, setforth in Section I of the enclosed Notice, are classified in the aggregateas a Severity Level III problem in accordance with the "General Statementof Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy),NUREG-1600.
In addition to your failure to ensure that the design basis was maintainedwhen you increased the torus temperature limit, the NRC identified otherviolations involving design deficiencies and errors in design calculations.These violations are set forth in Section II.A of the Notice. For example,a nonsafety-related component was used in a subsystem essential to thesafety-related function of the emergency diesel generators (EDGs). Specifically,air to the solenoid valves that operated the EDG service water coolingflow control valves (FCVs) was supplied from a nonsafety-related pressureregulator. Failure of the pressure regulator could have resulted in amalfunction that could have prevented operation of the EDGs due to a lossof all service water. In another case, incorrect design inputs were usedin the calculation of NPSH margin for the RHR pumps. Nonconservative datawas used in lieu of actual test data.
The NRC also identified other violations involving your failure to taketimely, effective action to correct conditions adverse to quality, andyour failure to report a condition that could have prevented the fulfillmentof a safety function. These violations are set forth in Sections II.Band II.C of the Notice. For example, you failed to properly evaluate andcorrect a nonconformance between the vendor recommended residual heatremoval (RHR) pump minimum flow requirement and the installed minimumflow capacity. Your determination that the installed minimum flow capacitywas adequate lacked the technical basis to conclude that the pumps wouldbe able to operate for several hours under minimum flow conditions duringpostulated accident scenarios because it was not supported by verificationfrom the vendor or test results. Additionally, the precautions added tothe RHR operating and surveillance procedures concerning RHR pump minimumflow operation did not adequately reflect the vendor recommendations.
The violations cited in Section II of the enclosed Notice have been classifiedindividually at Severity Level IV in accordance with the Enforcement Policy.
Following the A/E inspection, the NRC was concerned that, at the timeof the inspection, it did not appear that your design basis document (DBD)reviews would have identified the design issues identified by the A/Eteam. At the conference, you stated that you had committed to performthe DBD reviews and had identified the need for DBD validation prior toissuance of the NRC's 50.54(f) letter regarding the adequacy and availabilityof design basis information. However, while the DBD reviews were in progress,the validation effort had not been fully defined at the time of the A/Einspection. You indicated that the validation effort would have been designedto identify the type of problems identified by the A/E team. Based onthe findings of our follow-up inspection and the information providedat the conference, we have no current concern with your DBD validationeffort. We will review your DBD program again as part of our followupto your response to the violations described in the Notice.
With respect to the violations in Section I of the Notice, in accordancewith the Enforcement Policy, a base civil penalty in the amount of $55,000is considered for a Severity Level III problem. Since Vermont Yankee hasbeen the subject of escalated enforcement actions within the last twoyears,(1) the NRC considered whether creditwas warranted for Identification and Corrective Actionin accordance with the civil penalty assessment process in Section VI.B.2of the Enforcement Policy. No credit is warranted for identification because,while you identified the discrepancy between the FSAR and the TS limitfor torus temperature, the NRC identified your failure to promptly evaluateand correct the discrepancy and your failure to report the condition outsideof the design basis. Credit is warranted for corrective actions becauseyour actions, once the violations were identified, were considered promptand comprehensive. Those actions, as described at the conference, include:1) reconstitution of the containment design basis; 2) development of amore formal document design change process; 3) improvements to your correctiveaction process, including establishing a single event report process,reduction of the backlog, and development of a trending program; and 4)revision of the reportability procedure to implement interim expectationspending further review of your administrative controls and guidelinesfor considering event complexity.
Therefore, to emphasize the importance of maintaining your facility inaccordance with its design, and promptly correcting conditions that arecontrary to the design, and in recognition of your previous escalatedenforcement actions, I have been authorized, after consultation with theDirector, Office of Enforcement, to issue the enclosed Notice of Violationand Proposed Imposition of Civil Penalty in the base amount of $55,000.
During the inspection, the NRC identified three additional apparent violationsinvolving failure to take timely corrective actions. With respect to thefirst of these apparent violations, after further review, the NRC hasconcluded that the failure to enter TS Limiting Conditions for Operation(LCOs) when performing testing does not constitute a condition adverseto quality in accordance with 10 CFR Part 50, Appendix B. Therefore, thefailure to take prompt corrective action to address this issue did notconstitute a violation of NRC requirements. However, the NRC considersthe failure to enter LCOs during testing an administrative weakness. Bynot tracking LCO entries during surveillance testing, the control roomstaff loses an opportunity to better control planned maintenance activitiesand to better respond to emergent system operability concerns.
With respect to the remaining two apparent violations, the NRC has concludedthat your failure to request revisions to the service water subsystemand battery charger Technical Specifications in a timely manner did notconstitute violations, because, in both cases, you had put interim administrativecontrols in place to compensate for the nonconservative TSs. However,the NRC is concerned that the revision of the nonconservative TSs andcorrective action to address entry into LCOs for testing were delayeddue to deferral of your Improved Technical Specification (ITS) program.As stated in our regulations, TSs are derived from the analyses and evaluationincluded in the safety analysis report. By maintaining your TSs current,you help maintain our mutual confidence in your safety analysis and yourability to operate the plant in accordance with that analysis. We acknowledgeyour commitment to submit your application for ITS soon.
You are required to respond to this letter and should follow the instructionsspecified in the enclosed Notice when preparing your response. The NRCwill use your response, in part, to determine whether further enforcementaction is necessary to ensure compliance with regulatory requirements.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copyof this letter, its enclosure, and your response will be placed in theNRC Public Document Room (PDR).
William L. Axelson for
Enclosure: Notice of Violation and Proposed Imposition of
R. McCullough, Operating Experience Coordinator - Vermont Yankee
G. Sen, Licensing Manager, Vermont Yankee Nuclear Power Corporation
D. Rapaport, Director, Vermont Public Interest Research Group, Inc.
D. Tefft, Administrator, Bureau of Radiological Health, State of New Hampshire
Chief, Safety Unit, Office of the Attorney General, Commonwealth of Massachusetts
D. Lewis, Esquire
G. Bisbee, Esquire
J. Block, Esquire
T. Rapone, Massachusetts Executive Office of Public Safety
State of New Hampshire, SLO Designee
State of Vermont, SLO Designee
Commonwealth of Massachusetts, SLO Designee
D. Katz, Citizens Awareness Network (CAN)
Vermont Yankee Nuclear Power Station Docket No. 50-271
During an NRC inspection conducted between September 29, 1997, and November20, 1997, for which exits meetings were held on October 3, 1997, and November20, 1997, violations of NRC requirements were identified. In accordancewith the "General Statement of Policy and Procedure for NRC EnforcementActions," NUREG-1600, the NRC proposes to impose a civil penalty pursuantto Section 2.34 of the Atomic Energy Act of 1954, as amended (Act), 42U.S.C. 2282, and 10 CFR 2.205. The particular violations and associatedcivil penalty are set forth below:
I. VIOLATIONS ASSOCIATED WITH TORUS TEMPERATURE
A. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for structures, systems, and components, are correctly translated into specifications, drawings, procedures and instructions and that measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations.
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 90F to 100F) 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 90F was assumed in these analyses. (01013)(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, such as failures, malfunctions, deficiencies , deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude 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 100F for maximum torus normal operating temperature (in place as of June 6, 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.
- The nonconformance between the FSAR and the TS was not entered into the licensee's corrective action process until November 2, 1995.
- Administrative controls to limit normal torus temperature to 90F, 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 90F 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 90F 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 90F. Since no instances were identified, the licensee incorrectly concluded 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 90F was assumed, the 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 90F for greater than 24 hours 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 basis of the plant in that an initial torus temperature of 90F 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 temperature of 90F was assumed. (01033)
These violations in the aggregate constitute a Severity Level III problem(Supplement I).
Civil Penalty - $55,000.
A. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for structures, systems, and components, are correctly translated into specifications, drawings, procedures and instructions and that measures shall be established for the selection and review of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Criterion III also requires that measures 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)
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)
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 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)
4. Contrary to the above, on December 6, 1995, the licensee used incorrect design inputs in the calculation of NPSH margin for the RHR pumps in calculation VYC-808, Rev. 2. pecifically, 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)
5. Contrary to the above, prior to November 20, 1997, the licensee failed 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. (06014)
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 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 (l), which were superseded by a later revision before the licensee issued Calculation VYC-298, Rev. 10. The latest revisions of the drawings indicated some dc load changes. (07014)
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 equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude 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.
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 350 gpm. Specifically, in a letter dated November 13, 1986, the vendor notified the licensee that the minimum flow for the RHR pumps should be increased to 2700 gpm for continuous operation (more than 2 hours of operation in 24 hours) and to 2075 gpm for intermittent operation. However, the licensee failed to adequately correct this nonconformance, despite prior opportunities, 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 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 hours) had the same relationship to pump degradation as the length of a specific event (4 to 5 hours 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 adequately reflect the vendor recommendations, provided in 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 vendor recommendations, these instructions did not reflect the recommendation, provided in May 1997, that RHR pump operation should not be sustained at a flow rate of 350 gpm for more than 30 seconds during surveillance tests. (08014)
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 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 quality involving the service water (SW) and emergency diesel generator (EDG) support systems was appropriately evaluated and corrected. 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)
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 alone could have prevented the fulfillment of the safety 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 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 submitted a LER, a period in excess of 30 days. The failure to provide adequate instructions for RHR pump minimum flow requirements could have resulted in failure of the RHR pumps which would have prevented the fulfillment of a safety function of a system required to remove residual heat. (10014)
Pursuant to the provisions of 10 CFR 2.201, Vermont Yankee Nuclear PowerCorporation (Licensee) is hereby required to submit a written statementor explanation to the Director, Office of Enforcement, U.S. Nuclear RegulatoryCommission, within 30 days of the date of this Notice of Violationand Proposed Imposition of Civil Penalty (Notice). This reply should beclearly marked as a "Reply to a Notice of Violation" and should includefor each alleged violation: (1) admission or denial of the allegedviolation, (2) the reasons for the violation if admitted, and ifdenied, the reasons why, (3) the corrective steps that have been takenand the results achieved, (4) the corrective steps that will be takento avoid further violations, and (5) the date when full compliance willbe achieved. If an adequate reply is not received within the time specifiedin this Notice, an Order or a Demand for Information may be issued aswhy the license should not be modified, suspended, or revoked or why suchother action as may be proper should not be taken. Consideration may begiven to extending the response time for good cause shown. Under the authorityof Section 182 of the Act, 42 U.S.C. 2232, this response shall be submittedunder oath or affirmation.
Within the same time as provided for the response required above under10 CFR 2.201, the Licensee may pay the civil penalty by letteraddressed to the Director, Office of Enforcement, U.S. Nuclear RegulatoryCommission, with a check, draft, money order, or electronic transfer payableto the Treasurer of the United States in the amount of the civil penaltyproposed above, or the cumulative amount of the civil penalties if morethan one civil penalty is proposed, or may protest imposition of the civilpenalty in whole or in part, by a written answer addressed to the Director,Office of Enforcement, U.S. Nuclear Regulatory Commission. Should theLicensee fail to answer within the time specified, an order imposing thecivil penalty will be issued. Should the Licensee elect to file an answerin accordance with 10 CFR 2.205 protesting the civil penalty, in wholeor in part, such answer should be clearly marked as an "Answer to a Noticeof Violation" and may: (1) deny the violation(s) listed in this Notice,in whole or in part, (2) demonstrate extenuating circumstances, (3) showerror in this Notice, or (4) show other reasons why the penalty shouldnot be imposed. In addition to protesting the civil penalty in whole orin part, such answer may request remission or mitigation of the penalty.
In requesting mitigation of the proposed penalty, the factors addressedin Section VI.B.2 of the Enforcement Policy should be addressed. Any writtenanswer in accordance with 10 CFR 2.205 should be set forth separatelyfrom the statement or explanation in reply pursuant to 10 CFR 2.201, butmay incorporate parts of the 10 CFR 2.201 reply by specific reference(e.g., citing page and paragraph numbers) to avoid repetition. The attentionof the Licensee is directed to the other provisions of 10 CFR 2.205, regardingthe procedure for imposing a civil penalty.
Upon failure to pay any civil penalty due which subsequently has beendetermined 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 collected by civilaction pursuant to Section 234c of the Act, 42 U.S.C. 2282c.
The response noted above (Reply to Notice of Violation, letter with paymentof civil penalty, and Answer to a Notice of Violation) should be addressedto: J. Lieberman, Director, Office of Enforcement, U.S. Nuclear RegulatoryCommission, One White Flint North, 11555 Rockville Pike, Rockville, MD20852-2738, with a copy to the Regional Administrator, U.S. Nuclear RegulatoryCommission, Region I and a copy to the NRC Resident Inspector at the facilitythat is the subject of this Notice.
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 thePDR without redaction. If personal privacy or proprietary informationis necessary to provide an acceptable response, then please provide abracketed copy of your response that identifies the information that shouldbe protected and a redacted copy of your response that deletes such information.If you request withholding of such material, you must specificallyidentify the portions of your response that you seek to have withheldand provide in detail the bases for your claim of withholding (e.g., explainwhy the disclosure of information will create an unwarranted invasionof personal privacy or provide the information required by 10 CFR 2.790(b)to support a request for withholding confidential commercial or financialinformation). If safeguards information is necessary to provide an acceptableresponse, please provide the level of protection described in 10 CFR 73.21.
this 14th day of April, 1998
1. e.g., A Notice of Violation with a proposed $50,000 civil penalty was issued on August 23, 1996, for a Severity Level III violation involving failure to analyze ECCS equipment to be free from single failures (EA 96-210).
2. This 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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