ML20217N597

Notice of Violation & Proposed Imposition of Civil Penalty in Amount of $165,000.Violations Noted:Between Jan 1995 & 970120 Licensee Failed to Take Prompt & Effective Corrective Action for Significant Condition Adverse to Quality

ML20217N597
Person / Time
Site:	Pilgrim
Issue date:	04/27/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20217N541	List: IR 05000293/1997005 ML20217N597
References
EA-97-482, EA-97-525, NUDOCS 9805050381
Download: ML20217N597 (12)

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i ENCLOSURE l NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY l

Boston Edison Company Docket No, 50-293 Pilgrim Station License No. DPR-35 EAs 97-482: 97-525 During NRC inspections conducted between May 14,1997, and October 10,1997, for which i exit meetings were held on July 18,1997, August 28,1997, October 10,1997, and

November 19,1997, violations of NRC requirements were identified. In aiccordance with the

" General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the NRC proposes to impose a civil penalty 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 pensity are set forth below:

1. VIOLATIONS ASSESSED CIVIL PENALTIES A. VIOLATION ASSOCIATED WITH CONTAINMENT OVERPRESSURE 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Actions," requires, in l part, that measures shall be established to assure that conditions adverse to {

quality such as failures, malfunctions, deficiencies, deviations, defective l 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 is taken to preclude repetition.

10 CFR 50.59, " Changes, tests and everiments," permits the licensee, in part, to make changes to the facility as described in the safety analysis report without prior Commission approval provided the change does not involve an i 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. A proposed change shall be deemed to involve a USQ (i)  !

if the probability of occurrence or the consequences of an accident or i malfunction of equipment important to safaty previously evaluated in the safety I analysis report may be increased; or (ii) if a possibility for an accident or 1

malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.  ;

1 Appendix B to 10 CFR 50, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants", Criterion lil, " Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

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, Enclosure 2 10 CFR 50.71(e) requires, in part, that the final safety analysis report (FSAR) j be updated periodically to assure that the information included in the FSAR l- contains the latest material developed. Revisions to the FSAR shall be submitted annually or six months after each refueling outage provided the interval between submittats does not exceed 24 months. The revisions must reflect all changes up to a maximum of 6 months prior to the date of filing.

Contrary to the above, between January 1995 and January 20,1997, the  !

l licensee failed to take prompt and effective corrective action for a significant I condition adverse to quality, failed to assure that applicable regulatory i requirements and the design basis are correctly translated into specifications,  !

drawings, procedures, and instructions, and failed to perform an adequate l l written safety evaluation which provides the bases for the determination that i a design change did not involve a USQ. The deficiency involved ECCS net positive suction head (NPSH) calculations performed to support safety l l

evaluations for a design change to the drywell piping insulation. Specifically, I the calculations were changed to credit containment overpressure as a result of modification of insulation on recirculation loop piping located in the drywell in 1984. BECo Safety Evaluation (SE) No.1638, approved on August 31,  !

1984, was performed to support the design change. The design change i involved a USQ in that the probability of a malfunction of the ECCS pumps (i.e.,

residual heat removal (RHR) and core spray) was increased due to the l

potentially higher line pressure losses caused by the collection of insulation l debris on the pump suction strainers. The design change took credit for post-l accident containment overpressure to assure adequate ECCS pump NPSH.

l Crediting of containment overpressure was inconsistent with the plant design i basis as described in Section 14.5 of the Updated Final Safety Analysis Report (UFSAR). However, the licensee failed to recognize that crediting containment overpressure increased the probability of a malfunction of the ECCS pumps, and SE-1638 incorrectly concluded that the change did not involve a USQ.

Consequently, the change was made without NRC approval. This condition adverse to quality was not appropriately addressed until January 20,1997, i

when BECo requested NRC review and approval for including containment pressure as a component of NPSH margin in the Pilgrim licensing basis, despite prior opportunities to do so, namely:

1. in 1995, the service water system operational performance inspection (SWSOPI) self-assessment identified that the 1984 safety evaluation may have improperly credited containment overpressure in the NPSH calculations.

! 2. On March 25,1996, the licensee completed a new safety evaluation (SE-2971) which supported the previous replacement of all piping thermal insulation in the drywell and superseded SE-1638. SE-2971 also incorrectly concluded that the 1984 plant modification did not involve a USQ.

3. The report of an independent review of the containment overpressure issue performed by Yankee Atomic Electric Company, dated June 5, l 1996, concluded that containment overpressure was not credited in the i

Enclosure 3 Pilgrim licensing basis; however, prompt action was not taken to correct the deficiency.

In addition, between 1984 and June 1996, the licensee did not update Section 14.5 of the FSAR to reflect the design bases and methods for calculating the NPSH for the ECCS pumps as impacted by the modification to the ECCS pump strainers in 1984. Specifically, to support the modification, containment overpressure was credited in calculation of ECCS pump NPSH margin. FSAR Figure 14.5-10, "NPSH Availability for RHR and Core Spray System", was not revised to reflect the design bases and methods for calculating NPSH until June 1996. (01013)

This is a Severity Level ill violation (Supplement 1).

Civil Penalty $55,0GO.

B. VIOLATION ASSOCIATED WITH 480/120 VOLT TRANSFORMER REPLACEMENT 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Actions," 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 ronconformances are promptly identified and corrected. In the case of significrn conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action is taken to preclude repetition.

10 CFR 50.59, " Changes, tests and experiments," permits the licensee, in part, "o make changes to the facility as described in the safety analysis report without prior Commission approval provided the change does not involve an unreviewed safety question (USQ). The licensee shall maintain records of changes in the facility and these records must include a written safet'/

evaluation which provides the bases for the determination that the change doe 1 not involve a USO. A proposed change shall be deemed to involve a USQ i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increr. sed; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The Pilgrim Final Safety Analysis Report (FSAR), section 8.8.3, Safety Design Basis for the 120 Vac safeguard control subsystem describes that the 120 Vac safeguard control subsystem was designed and installed in accordance with IEEE-279 Standard. IEEE-279 Standard section 4.6, Channel Independence, required that the signals from both channels to be independent to accomplish decoupling of the effects of electric transients. Section 8.8.3.3 of the FSAR stated that the 120 Vac safeguard control subsystem was arranged so that no single component failure would prevent the system from providing power to the hydrogen / oxygen analyzer subsystem.

Enclosure 4 .j Contrary to the above, between April 1,1997 and October 10,1997, the licensee failed to take adequate corrective actions to preclude the recurrence I of a significant condition adverse to quality, and failed to perform an adequate written safety evaluation which provides the bases for the determination that a design change did not involve a USQ. The deficiency involved an unintended trip function of the microprocessors associated with two transformers (X55 and X56), which provide the 120 Vac power to the safeguard control subsystem.

The unintended trip function caused a common-mode malfunction (power loss) of both transformers due to a voltage transient on April 1,1997. Specifically, following the event on April 1,1997, the safety evaluation (SE 3091, dated April 10,1997) and engineering performed to support replacement of the microprocessors to eliminate the unintended trip function were inadequate in that the hardware and software of the microprocessors were not sufficiently evaluated to determine that the modification did not involve a USO. For example, voltage transients such as harmonic distortion or noise were not addressed, and the evaluation did not consider vendor configuration management, coding standards, or life cycle issues, all of which could have created a malfunction of a different type and, therefore, involved a USQ.

Consequently, the modifmation of the transformers in April 1997, a change that involved a USQ, was made without prior NRC approval. (02013)

This is a Severity Level ill violation (Supplement I).

Civil Penalty $55,000.

C. ADDITIONAL VIOLATIONS ASSOCIATED WITH INADEQUATE CORRECTIVE ACTIONS FOR KNOWN TECHNICAL ISSUES 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Actions," 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 is taken to preclude repetition.

l 10 CFR 50.59, " Changes, tests and experiments," permits the licensee, in part, i to make changes to the facility as described in the safety analysis report without prior Commission approval provided the change does not involve an unreviewed safety question (USO). 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 USO. A proposed change shall be deemed to involve a USO (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

Enclosure 5 I

Appendix B to 10 CFR Part 50, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants", Criterion lil, " Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

1. VIOLATION ASSOCIATED WITH SSW DESIGN INLET TEMPERATURE 1

Contrary to the above, between July 1994 and August 28,1997, the licensee failed to promptly identify and correct discrepancies associated with operation of the plant with salt service water (SSW) system inlet temperatures higher than the design temperature specified in the UFSAR, a condition adverse to quality. Specifically, in July 1994 and August 1995, SSW system inlet temperature exceeded the design temperature of 65'F, as used in the accident analysis and as described in UFSAR Section 14.5.3. Additionally, plant procedures were not consistent with the UFSAR with respect to the SSW design inlet temperature. Procedure 2.2.32, " Salt Service Water System," contained no guidance regarding the temperature limit and procedure 2.2.30, " Reactor Building Closed Cooling Water System," specified a design inlet SSW temperature of 75'F, representing a failure to properly translate the SSW design temperature limit into procedures. This condition adverse to quality was not promptly identified and corrected, despite prior opportunities to do so, namely:

a. In July 1994, the licensee recognized that the elevated SSW temperature was a nonconforming condition and performed I evaluations that concluded that the reactor building closed cooling water (RBCCW) system was operable with SSW inlet temperatures up to 75'F. However, the licensee failed to identify that operation of the plant with SSW system inlet temperature

j. above 65'F was a condition outside of the design basis of the l plant.

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b. Although the licensee had identified that a design basis change  ;

was needed to change the SSW inlet temperature from 65'F to 1 75'F in July 1994 and again during the SWSOPl in 1995, they did not take prompt effective corrective actions to revise the design basis. A written safety evaluation to support a change to the design basis was not performed until March 1996.

However, the safety evaluation that was performed in March i 1996, was inadequate in that it inappropriately credited post- 1 accident containment overpressure in the analysis. Use of containment overpressure was inconsistent with the plant i licensing basis, i

c. The licensee did not identify the procedural discrepancies until l the SWSOPl was performed in 1995 despite prior opportunity in 1994. Procedure 2.2.32 was revised on February 23,1995, to  !

include an administrative limit stating that the RBCCW system l

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Enclosure 6 remained operable with a SSW inlet temperature up to 75'F.

However, the design discrepancy was not corrected because the l 75'F limit had not been incorporated into the licensee's current UFSAR analysis of record for design basis events and the performance of emergency core cooling equipment. These analyses used an ultimate heat sink temperature of 65'F.

Additionally, no written safety evaluation was performed to determine that the procedure change did not involve a USQ.

d. Interim corrective actions taken by the licensee to incorporate the use of SSW temperature " rolling averages" into plant instructions were also ineffective because the needed design analysis to support this method of determining inlet temperature was not performed. Specifically, in August and September 1995, the licensee approved the use of " rolling averages" to justify SSW temperature excursions above the 65'F design limit. However, no formal detailed analysis was performed to verify that temperatures above 65'F for limited periods of time were bounded by the analyses for design basis events and ECCS performance. (03013)

2. VIOLATION ASSOCIATED WITH SALT SERVICE WATER SYSTEM  ;

SINGLE FAILURE VULNERABILITY l Contrary to the above, between January 1995, and July 18,1997, the licensee failed to identify a design deficiency in the salt service water system that rendered the system vulnerable to a single failure, a condition adverse to quality, despite prior opportunity to identify the deficiency. The single failure vulnerability involved a loss of DC power during a design basis event at specific tidal conditions and Cape Cod Bay water temperatures. Under certain conditions, only one SSW pump would remain operating with the SSW headers cross-connected and insufficient NPSH; therefore, the single DC power failure could prevent the SSW system from performing its safety function. Although the licensee completed a single failure analysis on April 27,1997, in response to identification of single failure vulnerabilities in the SSW and RBCCW systems during the SWSOPl in 1995, the licensee failed to identify the vulnerability. (03023)

3. VIOLATION ASSOCIATED WITH ISOLATION OF NONESSENTIAL RBCCW LOADS l Contrary to the above, between February 1995, and August 28,1997, the licensee failed to promptly identify and correct a deficiency associated with isolation of nonessential RBCCW loads during design oasis conditions. Specifically, the design basis was not correctly translated into procedures in that there was no guidance in procedure 2.2.19.5, "RHR Modes of Operation for Transients," for isolating nonessential RBCCW system heat loads during a design basis accident as described in Section 10.5.5.3 of the UFSAR and other design

Enclosure 7 analyses. This condition adverse to quality was not promptly identified and corrected, despite prior opportunities to do so, namely:

a. Although the licensee identified that there was no procedural guidance to isolate the nonessential loads during the SWSOPl in l 1995, they failed to recognize that the condition caused the plant l to be outside the design basis of the plant.

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b. Although procedure 2.2.19.5 was revised on July 18,1997, to isolate non-essential RBCCW system heat loads if suppression i pool temperature exceeded 130'F, this corrective action was l not effective because no design-basis analysis was performed for the 130*F limit to demonstrate that the core spray pump bearing cooler would receive adequate flow during all design basis conditions. Additionally, no written safety evaluation was performed to determine that the restriction on isolation of the nonessential loads did not involve a USQ. A 50.59 safety l evaluation was required because the description in the UFSAR did not restrict isolation of the nonessential loads based on suppression pool temperature. (03033)

4. VIOLATION ASSOCIATED WITH RHR SYSTEM DESIGN FLOW RATES Contrary to the above, prior to August 28,1997, the licensee failed to promptly identify and correct a design deficiency associated with translation of the RHR design flow rate into plant procedures. The RHR l design flow rate of 5100 gpm used in design basis containment heat l transfer and pressure / temperature response calculations was not adequately translated into procedures. The RHR flow range of 4800 to 5100 gpm specified in Operating Procedure (OP) 2.2.19.5, "RHR Modes i of Operation for Transients", was not supported by calculations that considered the effects of instrument accuracy on post-accident containment response or RHR heat exchanger integrity. After the deficiency was identified by the NRC in July 1997, the licensee failed to identify the significance of the deficiency and failed to take effective corrective actions to resolve the problem. Specifically, Operating Procedure 2.2.19.5 was revised on July 18,1997, to throttle RHR flow not to exceed 5600 gpm; however, this corrective action was not effective because the specification of the higher RHR flow rate was not

! adequately supported with the required calculations and analyses and was not representative of the design basis. Additionally, no written safety evaluation was performed to determine that the higher system flow rate did not involve a USO. Specifically, no evaluation was performed to ensure that a flow rate of 5600 gpm would not have an adverse effect on the RHR heat exchangers. (03043) i

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5. VIOLATION ASSOCIATED WITH EDG LOADING CALCULATIONS Contrary to the above, between January 1995 and August 28,1997, the licensee failed to promptly identify and correct design deficiencies ,

associated with emergency diesel generator (EDG) loading calculations i and procedures. Specifically, calculation PS-79, " Diesel Generator I Loading", did not include the power drawn during current limit operation for the 250 vDC battery charger and did not address the effect on generator load by motor driven pump frequency variation. The limit ,

specified in the precautions of procedure 2.2.8, " Standby AC Power System", for EDG 2OOO-hour rating did not account for accuracy of the j kilowatt meter. Additionally, the diesel generator loads documented in '

design basis calculation PS-79 were not properly translated into the diesel generator loading information specified in procedure 2.2.8.

l Although diesel generator loading was assessed during the SWSOPl in 1995, the licensee did not identify the deficiencies in calculation PS-79 and procedure 2.2.8. The licensee had identified the inconsistencies  ;

between design-basis information contained in calculation PS-79, and i diesel generator loading information in procedure 2.2.8 prior to the SWSOPl in 1995 and had initiated a tracking item to revise procedure 2.2.8 during the SWSOPl. However, the licensee failed to take prompt corrective action to resolve the discrepancies. Although both calculation  !

PS-79 and procedure 2.2.8 had been revisec'since the SWSOPI, as of ,

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August 28, 1997, the calculation and the procedure were still  !

l inconsistent. (03053)

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6. VIOLATION ASSOCIATED WITH EDQ AMBIENT TEMPERATURE DESIGN i LIMIT Contrary to the above, between January 1995, and August 28,1997,

! the licensee failed to promptly identify and effectively correct a design discrepancy associated with the EDG design maximum outdoor ambient temperature limit of 88'F specified in UFSAR Section 10.9.3.9 and Table 10.9-2. During certain summer periods since initial plant startup, ambient temperatures exceeded 88'F; however, the licensee failed to l ensure that the design basis limitation for operation of the EDGs was l translated into specifications. The licensee failed to promptly identify l- and correct this condition adverse to quality, despite prior opportunities l to do so, namely:

a. During the SWSOPl in 1995, the licensee identified that the maximum ambient temperature for operation of the EDGs had been exceeded in the past. However, they failed to recognize that the condition caused the plant to be outside the design basis of the plant.

b. No limits were placed on EDG loading when operating above f ambient temperatures of 88 'F until June 20,1997.

Enclosure 9

c. The safety evaluations performed to support the change to a 100% water mixture in June 1997 (SE-3102) and the change back to a 50/50 glycol mixture in August (SE-3114) were not comprehensive and were based on preliminary input that was not properly validated. The safety evaluations did not address the effects of higher air temperature on key engine performance characteristics such as fuel consumption rate or the overall impact on accelerated engine wear and possible engine power de-rating. Although the testing performed to validate the analysis upon which these evaluations was based did not achieve the expected results, the EDGs were still considered operable.

(03063)

7. VIOLATION ASSOCIATED WITH ENVIRONMENTAL QUALIFICATION RELATED TO DRYWELL TEMPERATURE PROFILE 10 CFR 50.49 (e) requires, in part, that the electric equipment environmental qualification program must include and be based on the time-dependent temperature and pressure at the location of the electric equipment important to safety. The time-dependent temperature and pressure must be established for the most severe design basis accident l  !

during or following which this equipment is required to remain functional.

Contrary to the above, between January 1996, and August 28,1997, the licensee failed to take corrective action to preclude recurrence of a

deficiency associated with the environmental qualification (EQ) accident

! temperature profile for electrical equipment in the drywell, a significant condition adverse to quality. The condition involved a computer modeling error for certain small break sizes and an incorrect assumption that resulted in higher average drywell temperature than the analysis of record for the containment temperature profile used for EQ of electrical equipment in the drywell. The modeling error caused the analysis of I l record since 1987 to be nonconservative due to differences in the predicted peak temperature and drywell temperatures from one hour to approximately 220 hours0.00255 days <br />0.0611 hours <br />3.637566e-4 weeks <br />8.371e-5 months <br /> after the event.

In January 1996, the licensee identified and corrected the errors in the i drywell temperature profile; however, they failed to take action to preclude recurrence of the deficiency. Specifically, the licensee determined that the cause of the error was failure to review the input values and assumptions used by the vendor in their analysis; however, as of August 28,1997, no change had been made to engineering procedures to preclude repetition of the condition.

From 1987 to Januay 1996, a condition existed in which design basis drywell accident temperature profiles would have exceeded equipment environmental qualification temperature limits during postulated main steam line break accidents, placing the plant outside its design basis.

Although the licensee identified the condition in January 1996, they i

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l Enclosure 10 failed to recognize that the condition caused the plant to be outside the I design basis of the plant. (03073) l  !

! These violations (l.C.1 - 7) represent a Severity Level 111 problem (Supplement 1).

l Civil Penalty - $55,000. )

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! II. VIOLATIONS NOT ASSESSED A CIVIL PENALTY 10 CFR 50.72, "Immediate notification requirements for operating nuclear power .

reactors," requires, in part, that the licensee shall notify the NRC as soon as practical l and in all cases within one hour of any event or condition during operation that results  :

in the nuclear power plant being in a condition that is outside the design basis of the plant.

10 CFR 50.73, " Licensee event report system," requires, in part, that the licensee shall submit a Licensee Event Report (LER) within 30 days after the discovery of any event or condition that resulted in the nuclear power plant being in a condition that was I

outside the design basis of the plant.

Contrary to the above, notifications and reports were not made within the required times as evidenced by the following examples, each of which represents a separate violation:  ;

A. As of August 28,1907, BECo had not reported in accordance with 10 CFR 50.72 and 10 CFR 50.73 that in July 1994 and August 1995, SSW system inlet temperature exceeded the design temperature of 65'F, as used in the accident analysis and as described in UFSAR Section 14.5.3., representing a condition  ;

outside the design basis of the plant. (04014)

This is a Severity Level IV violation (Supplement 1).

l B. On June 6,1997, a design deficiency was identified in the SSW system that rendered the system vulnerable to a single failure in the event of a loss of DC power during a design basis event at specific tidal conditions and Cape Cod Bay water temperatures. UFSAR Section 10.7.2 indicates that no single active failure can prevent the system from achieving its safety objective. This condition was not reported to the NRC in accordance with 10 CFR 50.73 until July 18,1997. (05014)

This is a Severity Level IV violation (Supplement I).

j C. As of August 28,1997, BECo had not reported in accordance with 10 CFR 50.72 and 10 CFR 50.73 that in the past, as identified during the SWSOPl in 1995, the ambient temperature for operation of the EDGs had exceeded the maximum of 88'F as specified in UFSAR Section 10.9.3.9 and Table 10.9-2 representing a condition outside the design basis of the plant. (06014)

This is a Seventy Level IV violation (Supplement 1).

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Enclosure 11 D. As of August 28,1997, BECo had not reported in accordance with 10 CFR 50.72 and 10 CFR 50.73 the identification of errors that resulted in higher average drywell temperature than the analysis for the containment temperature profile used for EQ of electrical equipment in the drywell as specified in the General Electric analysis of record, SUDDS/RF 87-917, "Drywell Temperature Analysis,"

dated September 2,1987, representing a condition outside the design basis of the plant. (07014)

This is a Seventy Level IV violation (Supplement 1).

Pursuant to the provisions of 10 CFR 2.201, Boston Edison Company (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, l U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation l and Proposed imposition of Civil Penalty (Notice). This reply should be clearly marked as a l

" 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 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 Commission, 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 l 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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Enclosure 12 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, rnay be 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: James Lieberman, Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, One White Flint North,11555  !

l 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 facility that is the subject of this Notice.

l Because your response will be placed in the NRC Public Document Room (PDR), to the extent i 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 l 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 E9.9.t 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 1

l 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.

i l Dated at King of Prussia, Pennsylvania this 27th day of April 1998 l

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