EA-96-412 - Clinton (Illinois Power Company)

June 9, 1997

EAs 96-412, 97-001, 97-002, and 97-060

Mr. John G. Cook
Senior Vice President
Illinois Power Company
500 S. 27th Street
Decatur, Illinois 62525

SUBJECT: NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES - $450,000 (NRC Routine Inspection Report 50-461/96009(DRP), Special Inspection Report 50-461/96010(DRP), Operational Safety Team Inspection Report 50-461/96011(DRS), Inspection Report 50-461/96012(DRS), Special Inspection Report 50-461/96014(DRP), and Office of Investigations Report 3-96-047)

Dear Mr. Cook:

This refers to five inspections conducted from July 30, 1996, through January 23, 1997, at the Clinton Power Station (CPS). The inspections included evaluations and assessments of the 1) September 5, 1996, reactor recirculation pump seal failure, 2) operations and engineering activities supporting operations, 3) Division III emergency diesel generator inoperability, 4) feedwater containment isolation check valves' inoperability, and 5) radiation protection program. A management meeting was held on September 23, 1996, to discuss Illinois Power Company's assessment of the September 5, 1996, reactor recirculation pump seal failure event, and a public exit meeting was held on October 4, 1996, to present the NRC's inspection findings. Exits for the other inspections were conducted on November 21, 1996, December 12, 1996, and January 23, 1997. In addition, the NRC Office of Investigations conducted an investigation of the circumstances of the September 5, 1996, event. Because of the seriousness of the issues emanating from the inspections, predecisional enforcement conferences were conducted on February 4, March 4, and March 20, 1997.

Based on the information developed during the inspections and the information CPS provided during the predecisional enforcement conferences, the NRC has determined that a number of significant violations of NRC requirements occurred from September 5, 1996, through January 23, 1997. The violations demonstrate a lack of conservative decision-making, pervasive procedural adherence problems, inappropriate procedures, and a lack of rigor in conducting routine plant activities. These violations have occurred throughout various site organizations. Enclosure 1 contains a Notice of Violation and Proposed Imposition of Civil Penalties (Notice), that describes the violations.

The violations in the Notice are grouped in five sections as follows:

Section A of the Notice, "Reactor Recirculation Pump Seal Failure," describes two significant procedure violations during an activity undertaken on September 5, 1996, to allow continued plant operation by isolating the "B" reactor coolant loop, thereby stopping, or limiting, the suspected leaking pump shaft seal package. The procedural violations directly resulted in reactor coolant leakage considerably exceeding the allowed Technical Specification Limit and requiring a plant shutdown. CPS site supervision and management inappropriately emphasized attaining a single loop configuration to allow continued power operation. This resulted in a focus to operate the plant rather than the more conservative decision to shut down the unit. The deliberate actions of members of the shift crew on September 5, 1996, resulted in very significant violations. Based on a review of the evidence obtained from the Office of Investigations, NRC inspections, and enforcement conferences, CPS's actions demonstrated careless disregard for procedural requirements.

Section B of the Notice, "Failure to Follow Procedures," involves seventeen violations involving operations and radiation protection procedures which collectively demonstrates a breakdown in the control of activities involving the adherence to procedures. The violations demonstrated that CPS had established an environment that condoned procedure compliance through accomplishment of the user's interpretation of the procedures' intent without regard for the actual procedural steps.

Section C of the Notice, "Inoperable Emergency Diesel Generator," involves two violations associated with the Division III Emergency Diesel Generator (EDG). The licensee failed to maintain design control with this EDG and failed to identify and correct a condition adverse to quality. As a result, the EDG was inoperable for more than a year.

Section D of the Notice, "Failure to Perform Safety Evaluations," involves seven violations for a failure to perform safety evaluations required by 10 CFR 50.59. Safety evaluations were not performed to justify operating the plant differently than described in the updated safety analysis report (USAR); to justify the acceptability of performing tests on operable equipment; or to justify continued operation when as-found plant conditions were different than the USAR description. The examples represented a significant lack of attention toward the process for performing 10 CFR 50.59 evaluations.

Section E of the Notice, "Ineffective Corrective Actions to Resolve Inoperable Containment Penetrations," involves two violations for inoperable feedwater primary containment isolation valves caused by inadequate corrective action for a longstanding equipment problem and a failure to perform adequate testing to assure Technical Specification surveillance requirements were met. This resulted in the serious degradation of two containment penetrations such that they may not have been able to function when required.

Individually and collectively, the violations are serious. These violations included a pervasive problem regarding procedural quality and adherence which you acknowledged during the predecisional enforcement conference, weaknesses in the conduct of operations, and weaknesses in engineering support to operations. CPS management failed to ensure that its economic expectations had been put in proper focus with safe, conservative facility operation. In the aggregate, given the depth and breadth of the violations, it is evident that existing management control systems were not used effectively to ensure early detection and timely resolution of conditions adverse to safe plant operation. For example, a number of procedures, written to ensure that systems were operated and tested consistent with the facility's design bases, were inappropriately changed or not followed. Site-wide procedures were usually looked upon as guidance rather than as requirements to be followed. Tests were performed on operating systems without procedures and without required reviews to ensure that unreviewed safety questions did not exist. Beyond the operations oriented-violations, significant deficiencies existed in the engineering program with regards to safety evaluations and operability assessments. Other inspection activities identified a significant issue involving the Division III EDG being inoperable for almost a year due to an inadequate calibration procedure. In addition, the feedwater outboard containment isolation valves' ability to fully perform their intended safety function was not determined over the past operating cycle due to inadequate test configurations. Lastly, several radiation protection violations were caused by procedure compliance problems which were similar to those identified in operations and engineering. These violations occurred well after the September 5, 1996, reactor recirculation pump seal failure event indicating CPS was having difficulty implementing effective corrective actions for procedure compliance problems. In that regard, it is of particular concern that the corrective actions discussed at the March 20, 1997, predecisional radiation protection enforcement conference were narrowly focused on the individual items. Your staff failed to discuss the issues as an integrated theme from a radiation protection perspective and as they related to the broader procedure compliance problems discussed at the two previous enforcement conferences. This indicated that CPS was not looking at the identified problems in an integrated fashion.

In summary, the inspection findings demonstrated 1) a lack of conservative decision-making, 2) pervasive procedural adherence problems as indicated by individuals believing it was acceptable to not follow existing procedures in order to accomplish work activities, and 3) poor-quality procedures. Extensive NRC intervention was required to ensure CPS recognized and understood the problems' scope and magnitude. This included management's lack of recognition of its failure to make the conservative decision to shut down the unit on September 5, 1996, and CPS's lack of understanding of the associated procedure violations.

Due to the safety significance of the violations discussed above, their relationship to operational safety, and in accordance with the NRC Enforcement policy, NUREG-1600, the violations in Section A of the Notice have been categorized in the aggregate as a Severity Level II problem. The violations in Sections B - E have each been categorized as Severity Level III problems.

Significant civil penalties are warranted to emphasize to you and to other reactor licensees the importance of strong management oversight and direction from both the site and utility in maintaining a clear focus on operational safety; the need for plant personnel to challenge and investigate discrepancies; the need to adequately plan safety-significant activities; the need to take timely and effective corrective actions; and the need for a strong self-assessment program. Enforcement discretion is being exercised as provided by Section VII.A. of the Enforcement Policy to assess civil penalties of $200,000, (the maximum statutory amount) for the violations in Section A of the Notice, and civil penalties of $100,000 for the violations in Section B. In accordance with Section VI.B.2 of the Policy, civil penalties of $50,000 for the violations in each of Sections C - E are being assessed. The assessments are more fully described in Enclosure 2.

Accordingly, I have been authorized, after consultation with the Director, Office of Enforcement, and the Deputy Executive Director for Regulatory Effectiveness, to issue the enclosed Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $450,000. But for the extended shutdown of CPS and the substantial corrective actions taken during this shutdown period to improve performance, a larger civil penalty would have been proposed.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements. I further note that my staff continues to review activities at CPS, and further enforcement actions may be taken if additional violations are identified.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosure(s), and your response will be placed in the NRC Public Document Room (PDR).

Sincerely, Original Signed By A. Bill Beach Regional Administrator

Docket No. 50-461
License No. NPF-62

Enclosures:
1. Notice of Violation and Proposed Imposition of Civil Penalties
2. Civil Penalty Assessment

cc w/encls:
W. D. Romberg, Assistant
Vice President
P. Yocum, Plant Manager
Clinton Power Station
R. Phares, Manager-Nuclear Assessment
J. Sipek, Director - Licensing
Nathan Schloss, Economist
Office of the Attorney General
G. Stramback, Regulatory Licensing
Services Project Manager
General Electric Company
Chairman, DeWitt County Board
State Liaison Officer
Chairman, Illinois Commerce Commission

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NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES

Illinois Power Company Docket No. 50-461 Clinton Power Station License No. NPF-62 EA Nos: 96-412, 97-001, 97-002, and 97-060

During five NRC inspections conducted from July 30, 1996, to January 23, 1997, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the NRC proposes to impose civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act), 42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalties are set forth below:

A. Reactor Recirculation Pump Seal Failure

Clinton Power Station (CPS) Technical Specification 5.4.1.a requires, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Revision 2, Appendix A, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," states, in part, that procedure adherence (Section 1.d) and recirculation system (Section 4.a) are typical safety-related activities which should be covered by written procedures.

CPS 1005.14 (Rev. 4), "Formatting of Procedures and Documents," a procedure required by Section 1.d of RG 1.33, at Step 8.1.11.4 states, in part, that if a specific order of performing the procedure is required, an asterisk (*) at the beginning of the section to annotate that the steps are to be performed in the sequence that they are written.

CPS 3302.01 (Rev. 18), "Reactor Recirculation," a procedure required by Section 4.a of RG 1.33, specified that Section 8.2.4 was required to be performed in sequence as indicated by the "*" next to the section heading.

Section 8.2.4 of CPS 3302.01 specified, in part, the sequence to isolate an idle reactor coolant loop as follows:

• Step 8.2.4.4: Cool the idle loop to < 250F

• Step 8.2.4.5: Shut 1B33-FO75B, "Pump B Seal Stage Shutoff Valve"

• Step 8.2.4.6: Shut 1C11-FO26B, "CRD Supp Isol to RR Pump B"

1. Contrary to the above, on September 5, 1996, the operators failed to perform the steps in the sequence specified in Section 8.2.4 as demonstrated by their failure to wait until the idle reactor coolant loop had cooled to < 250F as specified in Step 8.2.4.4 before performing step 8.2.4.5 and shutting 1B33-FO75B.

2. Contrary to the above, on September 5, 1996, the operators failed to perform the steps in the sequence specified in Section 8.2.4 as demonstrated by their failure to wait until the idle reactor coolant loop had cooled to < 250F before performing Step 8.2.4.6 and shutting 1C11-FO26B.

This is a Severity Level II problem (Supplement I). Civil Penalty - $200,000.

B. Failure to Follow Procedures

1. Clinton Power Station (CPS) Technical Specification 5.4.1.a requires, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Revision 2, Appendix A, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," states, in part, that the following are typical safety-related activities that should be covered by written procedures: shift and relief turnover (Section 1.g); log entries (Section 1.h); changing loads (Section 2.f); fuel storage cooling system (Section 4.k); control room heating and ventilation system (Section 4.s); and loss of coolant (including leak-rate determination (Section 6.a)).

a. CPS 3317.01 (Rev. 16), "Fuel Pool Cooling Cleanup," a procedure required by Section 4.k of RG 1.33, at Step 8.1.2.14, required that valve 1FC004A or 1FC004B on the idle spent fuel pool loop be closed.

Contrary to the above, between September 18 and 25, 1996, operators failed to close 1FC004A in the idle "A" train fuel pool cooling loop as required by CPS 3317.01.

b. CPS 3402.01 (Rev. 14), "Control Room HVAC," a procedure required by Section 4.s of RG 1.33, at Step 8.1.1.1.1, required the final position of OVC043B, "Moisture Separator Drain Valve," be open and OVC096B, "Loop Seal Fill Valve," be closed upon completion of filling the makeup air filter moisture separator loop seal.

Contrary to the above, on September 18, 1996, after filling the makeup air filter moisture separator loop seal, the licensee failed to open the moisture separator drain valve (OVC043B) and close the loop seal fill valve (OVC096B) as required by CPS 3402.10.

c. CPS 4001.01 (Rev. 7), "Reactor Coolant System Leakage," a procedure required by Section 6.a of RG 1.33, at Step 4.4, required the control room to notify radiation protection (RP) and request area samples and or AR/PR trending information to assist in detecting the location/source of the leak.

Contrary to the above, on September 5, 1996, RP was not notified of the need to assist in identifying the unidentified leakage.

d. CPS 3005.01 (Rev. 18), "Unit Power Changes," a procedure required by Section 2.f of RG 1.33, at Step 6.1.b, required the control room to notify the chemistry department, after a thermal power change of greater than 15% in one hour, to perform the applicable sections of CPS 9940.01, "Weekly Chemistry Surveillance Log". In this case, the applicable sections required a gaseous sample.

Contrary to the above, on September 6, 1996, thermal power was changed from 55% to 38%, an amount greater than 15%, within a one-hour period, and the control room failed to notify the chemistry department so it could take a gaseous sample.

e. CPS 1401.01 (Rev. 20) "Conduct of Operations," is a procedure required by sections 1.g and 1.h of RG 1.33.

(1) Section 8.4.3.13 of CPS 1401.01 required the Line Assistant Shift Supervisor (LASS) to inform the relief operator of, at a minimum, current plant status, operations in progress and work to be performed in the immediate future.

Contrary to the above, on September 17, 1996, the LASS failed to inform the relief operator of work to be performed in the immediate future which was going to affect fuel building differential pressure. Specifically, the relief operator was not informed that the work activity would result in a high differential pressure fuel building annunciator alarm in the control room. Consequently, an operator was unnecessarily dispatched to investigate the cause of the expected alarm.

(2) Section 8.3.3.1 of CPS 1401.01 required the shift supervisor to remain in a monitoring role during off normal operation unless he determines that the LASS is not able to deal with the situation.

Contrary to the above, on September 6, 1996, the shift supervisor failed to remain in a monitoring role and directed activities to place the unit in single loop operation without determining that the LASS was not able to deal with the situation.

(3) Section 8.4.4.10. e) and f) of CPS 1401.01 required that significant plant operating data, such as abnormal plant conditions and plant transients, be entered in the shift supervisor and main control room journals.

Contrary to the above, on September 6, 1996, no entry was made in the shift supervisor's journal for an abnormal condition, when suppression pool level exceeded the technical specification limit requiring entry into a limiting condition for an operation action statement.

(4) Section 8.1.6.2.1 of CPS 1401.01 required the Shift Technical Assistant (STA) to assist the shift supervisor in evaluating conditions for possible entry into an emergency classification condition.

Contrary to the above, on September 5, 1996, the STA failed to assist the shift supervisor in evaluating conditions for possible entry into an emergency classification condition.

2. CPS Technical Specification 5.2.2e, "Unit Staff," requires, in part, that administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions. Controls shall be included in the procedures such that individuals shall be reviewed monthly by the plant manager, or his designee, to ensure that excessive hours have not been assigned.

CPS 1001.10 (Rev. 6), "Control of Working Hours," Step 8.7, which implements the overtime control and review requirements of Technical Specification Section 5.2.2e, requires that individual overtime records shall be reviewed at least monthly by department management to ensure that excessive hours have not been assigned, and to ensure that overtime limits have not been exceeded without prior authorization.

Contrary to the above, during the period from April 1996 through August 1996, the required reviews of overtime usage by the Operations Department personnel were not performed.

3. 10 CFR 50.54(m)(2)(iii) states that when a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the unit's technical specifications, such licensees shall have a person holding a senior operator license of the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be present at the controls at all times.

CPS 1001.05 (Rev. 8) "Authorities and Responsibilities of Reactor Operators for Safe Operation and Shutdown," which implements the requirements of 10 CFR 50.54(m)(2)(iii) at Section 2.1.2, defines the "A" reactor operator (RO) as the licensed RO present "at the controls" of a fueled nuclear power unit.

Contrary to the above, on September 18, 1996, with the reactor fueled, the "A" RO left the "at the controls" area for approximately 3 minutes without obtaining an appropriate relief.

4. Clinton Power Station Technical Specification 5.4.1.a requires that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978.

a. Section 7(e)(1) of Appendix A to RG 1.33, requires a radiation protection (RP) procedure for access control to radiation areas including a Radiation Work Permit system.

Station Procedure No. 1905.10 (Rev. 17), "Radiation Work Permit," implemented the requirement of Section 7(e)(1) of Appendix A to RG 1.33 and stated at step 6.2 that deviations from a Radiological Safety Work Plan (RSWP) are not permitted without the approval of the Supervisor-Radiological Operations.

RSWP 96-01 (Rev. 1), states:

• If an extended time window (12 hours minimum) exists in which no irradiated core components or fuel movements are to occur, the requirements of this RSWP may be relaxed (Section C(2));

• If the RSWP is temporarily suspended, the restricted area posting on the drywell 790' elevation and the notification posting restricting access to the drywell 767' elevation shall be removed and the dropped fuel bundle warning system shall be placed in standby (Section II(a)).

Contrary to the above, on November 14, 1996, workers were allowed to enter the 796' elevation of the drywell under the following deviations from RSWP 96-01 that had not been approved by the Supervisor-Radiological Operations:

i. Fuel movement was suspended for a maximum 8 hour period and not for the minimum 12 hour period specified in section C(2) of RSWP 96-01 prior to the entry.

ii. The restricted area posting on the drywell 790' elevation and the notification posting restricting access to the drywell 767' elevation were not removed, and the dropped fuel bundle warning system was not placed in standby prior to suspension of RSWP 96-01 as specified in section II(a) of RSWP 96-01.

b. Section 7(e)(7) of Appendix A to RG 1.33 requires a radiation protection (RP) procedure for Personnel Monitoring.

Station Procedure No. 1032.02 (Rev. 23), "Security Access Control," implemented Section 7(e)(7) of Appendix A to RG 1.33 and required at Step 8.8.2 that an individual remain in the immediate area and contact RP personnel if the individual alarms a radiation portal monitor twice.

Contrary to the above, on December 28, 1996, and on January 7, 1997, a records supervisor and auxiliary operator, respectively, exited the plant after twice alarming the gatehouse radiation portal monitor and without contacting RP personnel as specified at Step 8.8.2 of Station Procedure No. 1032.02.

c. Section 7(e)(1) of Appendix A to RG 1.33 requires a radiation protection (RP) procedure for access control to radiation areas.

Station Procedure 1024.02 (Rev. 4) "Radiological Work Control," implemented Section 7(e)(7) of Appendix A to RG 1.33 and required at Step 6.1.1 that workers adhere to established RP control requirements unless issued written or verbal guidance from RP personnel.

RP control requirements contained specific prohibitions against eating, drinking and smoking in the Radiological Controlled Area (RCA) were posted at various locations in the plant, communicated during Nuclear General Employee Training (NGET) and were listed on page 15 of the refueling outage (RF-6) handbook distributed to all personnel. Also, during NGET workers were instructed on the proper radiological controls which shall be used during ingress/egress to/from a contaminated area including removing protective clothing when exiting contaminated areas.

i. Contrary to the above, on November 22, 1996, the licensee identified that an unapproved sleeping/smoking area had been set up inside the Radiological Controlled Area (730' elevation of the radiological waste building), comprising of three sleeping places and used (freshly smoked) cigarette butts.

ii. Contrary to the above, on January 7, 1997, a worker exited a posted contaminated area prior to removing his protective clothing.

d. Section 7(b)(1) of Appendix A to RG 1.33 requires procedures for limiting the release of solid radioactive waste material such as spent resin and filter sludge to the environment.

Procedure STD-P-03-028 (Rev 1) , "Waste Sluicing Procedure" was written to implement Section 7(b)(1) of Appendix A to RG 1.33.

Contrary to the above, on January 7, 1997, procedure STD-P-03-028 was found to be inadequate because it did not describe the vent path for the waste evaporator tank used during the sludge sluicing and did not describe the actual sluicing wand used during the job. The result of following this inadequate procedure, was the spread of radioactive material and the contamination of several workers when they disconnected a pressurized sludge hose.

e. 10 CFR 20.1501 requires, in part, that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present.

10 CFR 20.1701 requires, in part, that the licensee shall use, to the extent practical, process or other engineering controls to control the concentrations of radioactive material in air.

Pursuant to 10 CFR 20.1003, survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

i. Contrary to the above, on January 7, 1997, the licensee's evaluation failed to adequately evaluate the radiological conditions and potential radiological hazards present, or use appropriate engineering controls to control the concentrations of radioactive material in air, prior to disconnecting a hose which had become clogged during the transfer of radioactive material.

ii. Contrary to the above, on January 3, 1997, the licensee's evaluation failed to adequately evaluate the radiological conditions and potential radiological hazards present, or use appropriate engineering controls, during the removal of mirror insulation from reactor water cleanup system piping.

This is a Severity Level III problem (Supplements I and IV). Civil Penalty - $100,000.

C. Inoperable Emergency Diesel Generator

1. Technical Specification Limiting Condition for Operation (LCO) 3.8.1 requires that three diesel generators be operable. The LCO is applicable during Modes 1, 2, and 3 of operation.

Technical Specification Surveillance Requirement (SR) 3.8.1.11.c.1 requires that once every eighteen months it be verified that on an actual or simulated loss of offsite power signal each emergency diesel generator energizes permanently connected loads in < 12 seconds. Licensee Procedure CPS 9080.23, "Diesel Generator 1C Integrated," was intended to satisfy this SR.

SR 3.0.1 states, in part, that "failure to meet a SR, whether such failure is experienced during performance of the SR or between performances of the SR, shall be failure to meet the LCO."

Contrary to the above, from September 26, 1995, until November 5, 1996, Diesel Generator 1C was inoperable in that SR 3.8.1.11.c.1 could not be satisfied. On September 26, 1995, the licensee miscalibrated relay K54X. The miscalibration directly caused, on November 2, 1996, the inability of Diesel Generator 1C to satisfy SR 3.8.11.c.1 in that it could not be demonstrated that permanently connected loads were energized in less than 12 seconds. As demonstrated through the performance of Procedure CPS 9080.23, loads were not energized until 20 seconds after receipt of the actuation signal.

2. 10 CFR Part 50, Appendix B, Criterion III, "Design Control," states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.

10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions" states, 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.

Contrary to the above, on August 1, 1995, the licensee failed to correctly translate the design basis for the closing time of Diesel Generator 1C output breaker when the preventative maintenance task evaluation request sheet PEMDGM025 to calibrate relay K54X, "Permissive Signal for Closure of the Division III EDG Output Breaker" was issued. Specifically, the licensee incorporated the wrong delay time in PEMDGM025 which directly caused the closure time of Diesel Generator 1C output breaker to be in noncompliance with Technical Specifications. On September 26, 1995, the licensee failed to identify and correct a condition adverse to quality when workers found a substantial discrepancy between the as-found set point of .55 seconds and as-left set point (specified in PEMDGM025) of 11.28 seconds for the K54X relay. The licensee's failure to properly translate design requirements into working instructions (PEMDGM025), and the failure to both identify as a nonconformance and take corrective actions for the substantial difference between the as found and as left setpoints for the K54X relay contributed to Diesel Generator 1C being inoperable from September 26, 1995, to November 5, 1996, as described in violation C.1.

This is a Severity Level III problem (Supplement I). Civil Penalty - $50,000.

D. Failure to Perform Safety Evaluations

1. 10 CFR 50.59(a)(1), "Changes, Tests and Experiments," states, in part, that the holder of a license authorizing operation of a utilization facility may make changes to the facility as described in the safety analysis report and conduct tests or experiments not described in the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question.

10 CFR 50.59(b)(1) requires, in part, that the licensee shall maintain records of changes in the facility made pursuant to this section, to the extent that these changes constitute changes in the facility as described in the safety analysis report. The licensee shall also maintain records of tests carried out pursuant to paragraph (a) of 10 CFR 50.59. These records must include a written safety evaluation which provides the bases for the determination that the test does not involve an unreviewed safety question.

The applicable sections of the Updated Safety Analysis Report (USAR) include Figure 9.1-4, which shows the piping configuration for the spent fuel pool cooling system and provided the system configuration for an idle spent pool cooling loop; Section 5.4.7, which describes the design and functional basis of the Residual Heat Removal System (RHR); and Section 3.9.4, which describes the control rod drive system.

a. Contrary to the above, from 1989 until October 1996, the licensee had operated the fuel pool cooling and cleanup system, as prescribed in CPS 3317.01 (Rev. 16) at step 8.1.1.6 with a valve line up different from that shown on USAR Figure 9.1-4 and a written safety evaluation had not been performed to determine that the change to the system configuration specified in the USAR did not involve an unreviewed safety question. Specifically, the procedure required fuel pool cooling pump valve 1FC011A or B, for the idle loop, to be open, not closed as prescribed in Figure 9.1-4.

b. Contrary to the above, on August 1, 1996, the licensee performed a test that was not described in the safety analysis report, to verify that there was no negative impact on RHR system (Section 5.4.7 of the USAR) when cycled condensate to the containment was isolated. The test was performed without performing a written safety evaluation to determine that the test did not involve an unreviewed safety question.

c. Contrary to the above, on August 1, 1996, the licensee performed a test that was not described in the safety analysis report, to verify functionality of RHR (Section 5.4.7 of the USAR) water leg pump check valve 1E12F085A. The test was performed without performing a written safety evaluation to determine that the test did not involve an unreviewed safety question.

d. Contrary to the above, between August 2 and September 18, 1996, the licensee performed a weekly test that was not described in the safety analysis report, to verify the operability of RHR (Section 5.4.7 of the USAR) check valve 1E12F085A. The test was performed without performing a written safety evaluation to determine that the test did not involve an unreviewed safety question.

e. Contrary to the above, on May 3, 1995, with the reactor at power, the licensee performed a test that was not described in the safety analysis report, to determine if the control rod drive (CRD) pump's (Section 3.9.4 of the USAR) drop in CRD pressure was due to leaking valves or CRD pump degradation. The test was completed without performing a written safety evaluation to determine that the test did not involve an unreviewed safety question.

2. 10 CFR 50.59, "Changes, Tests and Experiments," 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 unreviewed safety question. 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 an unreviewed safety question.

10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," states, 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.

10 CFR 50.71(e), "Maintenance of Records, Making of Reports" requires, in part, that the licensee update the safety analysis report originally submitted as part of the application for the operating license to assure that the information included in the safety analysis report contains the latest material developed. The updated safety analysis report shall be revised to include the effects of, in part, all safety evaluations performed by the licensee in support of conclusions that changes did not involve an unreviewed safety question.

10 CFR 50.9(a), "Completeness and Accuracy of Information," requires, in part, that information provided to the NRC by a licensee or information required by regulation to be maintained by a licensee shall be complete and accurate in all material respects.

The applicable sections of the Updated Safety Analysis Report (USAR) are Section 9.4.5.2, which describes the cathodic protection and table 8.3-13 which describes the delay time for equipment to sequence on the emergency diesel generators.

a. Contrary to the above, the description of the facility in the USAR was not accurate in all material respects in that the USAR did not match the facility, required safety evaluations were not performed, corrective action was not implemented when conditions adverse to quality were identified, and the USAR was not properly updated. Specifically, in August 1995, the licensee had identified a condition adverse to quality, in that the cathodic protection system was not adequate to protect buried piping as stated in the USAR Section 9.4.5.2. As of October 1996, the licensee had neither taken prompt corrective action nor performed a written safety evaluation to determine if an unreviewed safety question existed for the degraded cathodic protection system.

b. Contrary to the above, the description of the facility in the USAR was not accurate in all material respects in that the USAR did not match the facility, required safety evaluations were not performed, corrective action was not implemented when conditions adverse to quality were identified, and the USAR was not properly updated. Specifically, in 1993 the licensee had identified a condition adverse to quality, in that a discrepancy existed between the as-built condition of the control room chillers and the system as described in USAR table 8.3-13. The licensee had identified that the chillers may auto-start in about 2.5 minutes after an event while the USAR documented that they would start 20 minutes after an event. As of October 1996, the licensee had neither taken prompt corrective action nor performed a written safety evaluation to determine if an unreviewed safety question exists for the auto-restart of the control room chillers after loss of power.

This is a Severity Level III problem (Supplement I). Civil Penalty - $50,000.

E. Ineffective Corrective Actions to Resolve Inoperable Containment Penetrations

1. Clinton Power Station (CPS) Technical Specification 3.6.1.a requires that feedwater primary containment isolation valves be operable.

Technical Specification 3.6.1.3.8, the surveillance requirement for Technical Specification 3.6.1, requires verification that the combined leakage rate for all secondary containment bypass leakage paths is < .08L _a when pressurized to > P _a .

10 CFR Part 50 Appendix J Section III.C.2.(a) requires, in part, that valves be pressurized with air at a pressure of P _a .

CPS surveillance procedure 9861.02 (Rev. 26), "Local Leak Rate Testing Requirement and Type C (Air) Local Leak Rate Testing," which implemented Technical Specification 3.6.1.3.8 and 10 CFR Part 50, Appendix J, requires at step 5.16.1 that both sides of the valve seat shall be drained below the valve seating surfaces prior to performing air leak testing of containment isolation valves.

Contrary to the above, on April 2 and 10, 1995, the licensee did not drain water from the outboard feedwater primary containment isolation valves (1B21F032 B & A, respectively) to below the valves' seating surfaces prior to leak testing the valves. This resulted in the failure to ensure that the primary containment isolation valves were operable during operating cycle 6.

2. 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action" requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, and defective material and equipment, 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 actions taken to preclude repetition.

Contrary to the above, during refueling outages 1, 2, 3, 4, 5, and 6 that were performed between January 1991 and October 1996, the licensee failed to establish corrective actions to preclude repeated failures of the outboard feedwater containment isolation check valves to pass the as-found local leak rate air test performed during each refueling outage, thus resulting in a significant condition adverse to quality.

This is a Severity Level III problem (Supplement I).
Civil Penalty - $50,000.

Pursuant to the provisions of 10 CFR 2.201, Illinois Power Company 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 Notice of Violation and Proposed Imposition of Civil Penalties (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 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 to 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 penalties 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 penalties 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 penalties will be issued. Should the Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalties, in whole or in part, such answers should be clearly marked as an "Answer to a Notice of Violation" and may: (1) deny the violations 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 penalties should not be imposed. In addition to protesting the civil penalties in whole or in part, such answers may request remission or mitigation of the penalties.

In requesting mitigation of the proposed penalties, 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 civil penalties.

Upon failure to pay any civil penalties due which subsequently have been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalties, unless compromised, remitted, or mitigated, may 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 penalties, and Answer to a Notice of Violation) should be addressed to the Director, Office 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 III and a copy to the NRC Resident Inspector station at the Clinton facility.

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 Lisle, Illinois
this 9th day of June 1997