Source: https://www.nrc.gov/reading-rm/doc-collections/enforcement/actions/reactors/ea97070.html
Timestamp: 2020-05-27 03:02:36
Document Index: 763874052

Matched Legal Cases: ['art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 10']

NRC: EA-97-070; EA-97-117; EA-97-127; EA-97-256 - Three Mile Island 1 (GPU Nuclear Corp.)
Home > NRC Library > Document Collections > Enforcement Documents > Significant Enforcement Actions > Reactor Licensees > EA-97-070; EA-97-117; EA-97-127; EA-97-256
EA-97-070; EA-97-117; EA-97-127; EA-97-256 - Three Mile Island 1 (GPU Nuclear Corp.)
EAs 97-070
Middletown, PA 17057-0191
SUBJECT: NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES
- $210,000 (NRC Inspection Reports Nos. 50-289/96-201;97-01; 97-02; 97-03;
& 97-04)
Dear Mr. Langenbach:
This letter refers to the five referenced NRC inspections conducted between November 12, 1996, and May 15, 1997, at the Three Mile Island Nuclear Station in Middletown, Pennsylvania, the findings of which were discussed with you and members of your staff during several exit meetings, the last of which was held on May 28, 1997. These inspections included: (1) a design inspection conducted by the NRR Special Inspection Branch that focused on the capability of certain systems to safely perform functions required by their design basis, your adherence to the design and licensing basis, and the consistency of the plant configuration to the Final Safety Analysis Report (FSAR); (2) two routine inspections conducted by the Region I resident and region based staff during which the NRC reviewed your process for classification of plant components and the environmental qualification of the reactor building (RB) emergency cooling fans; and (3) two special inspections conducted by NRC Region I emergency preparedness staff to review your emergency response capabilities and performance during a full participation emergency exercise on March 5, 1997, as well as during a remedial exercise on May 13, 1997. The related inspection reports were sent to you previously.
On May 22, 1997, and July 25, 1997, predecisional enforcement conferences (conferences) were conducted with you and members of your staff, to discuss the violations, their causes, and your corrective actions. The apparent violations identified in NRC Inspection Report 97-03, issued on May 29, 1997, related to RB emergency cooling fans not being environmentally qualified, as well as your failure to address this condition in a timely manner, were discussed at the May 22, 1997, conference, even though the inspection report had not been issued at the time of that conference. On May 28, 1997, you informed Mr. P. Eselgroth of the NRC Region I office that you agreed that another conference was not needed to further discuss these environmental qualification issues.
Based on the information developed during the five referenced inspections, and the information provided during the two conferences, a number of violations of NRC requirements are being cited and are described in the enclosed Notice of Violation and Proposed Imposition of Civil Penalties (Notice). The most significant violations relate to several areas of plant performance, and consist of: (1) inadequate engineering design controls, including incorrect design inputs for certain design basis calculations, inadequate verifications to assure the adequacy of design, and inadequate safety evaluations prior to making design changes; (2) poor implementation of the process for classifying components, resulting in a number of nuclear safety related components being downgraded to a lower classification without an appropriate safety evaluation or other supporting engineering documentation; (3) failure to ensure the RB emergency cooling fans were environmentally qualified; (4) failure to take timely and appropriate corrective actions for conditions adverse to quality that existed at the facility, including conditions related to the Decay Heat Removal system, to the quality assurance findings regarding inappropriate equipment classification downgrades, and to the environmental qualification deficiency; and (5) inadequate implementation of the emergency preparedness program.
With respect to the violations related to inadequate design engineering control and implementation, which are set forth in Section I of the enclosed Notice, several significant concerns were identified. For example, the design bases regarding the switchover of suction for the decay heat removal system (DHRS) pumps from the borated water storage tank (BWST) to the reactor building sump, were not correctly translated into operating procedures. The procedures are used by operators for commencing manual operations to perform this switchover during a large break loss of coolant accident (LOCA). In the calculation of the BWST level setpoint specified in the abnormal transient procedures, nonconservative assumptions and input data were used. As a result, the calculated BWST level setpoint determined for the switchover phase may not have prevented air entrainment in the DHRS pumps and the reactor building spray (RBS) pumps, due to vortexing in the BWST. In turn, this could have resulted in air binding and/or cavitation of the pumps, causing them to be inoperable during the critical recirculation phase of a large break LOCA. Additionally, design control measures were inadequate for changes that were made to remove the sodium thiosulfate tank and revise the BWST low-low level alarm setpoint. Specifically, containment overpressure was credited in DHRS pump net positive suction head (NPSH) calculations, contrary to the design bases, and the safety evaluation that was performed to support the changes failed to identify that the changes involved an unreviewed safety question. In addition to the calculational errors and inadequate safety evaluation, adequate design control measures did not exist for verifying or checking the adequacy of design in several instances, as described in the enclosed Notice.
With respect to the violations related to improper downgrading of the classification of equipment, which are set forth in Section II of the enclosed Notice, the quality classification checklists (QCLs) for several components, including the nuclear river (NR) water discharge valve motor operator, decay river (DR) water strainer motor, and auxiliary building ventilation system (ABVS), were inappropriately revised to downgrade the components from nuclear safety related (NSR) to a non-safety related classification. As a result, the downgraded components were not subjected to the quality assurance (QA) program requirements needed to assure system operability for postulated accident conditions was maintained. More specifically, these included requirements for the maintenance, testing, calibration, receipt inspection and procurement of parts. In addition, the procedure for performing the component classification and downgrade processes, did not receive the required review and, in some instances, was not followed by your engineering personnel, as detailed in the enclosed Notice.
With respect to the violation related to a lack of environmental qualification of certain equipment, which is set forth in Section V.A of the enclosed Notice, your staff determined that between March 17, 1986 and March 24, 1997, the three reactor building emergency cooling fans were not environmentally qualified in that the application of heat shrink tubing left a small length of exposed conductor at the spark plug connector to the fan motors. As a result, there was not reasonable assurance that the fans would function as required during post LOCA reactor building atmospheric conditions. Inoperability of these fans is contrary to the technical specifications, and given the eleven year duration of this problem, represents a significant regulatory concern. The NRC commends your engineering staff who, during repairs of a motor failure of one of the fans, questioned the qualification of the heat shrink tubing application which was not sealed. However, while the deficiency on the fan in question was promptly corrected, the same deficiency which existed on the other two fans was not corrected for an additional 33 days. The failure to take timely corrective action for this condition adverse to quality represented a violation of 10 CFR 50, Appendix B, Criterion XVI as described in Section III of the enclosed Notice.
With respect to the remaining violations related to inadequate identification and correction of problems in Section III of the enclosed Notice, several of the concerns previously discussed herein should have been identified and corrected sooner. In addition to the failure to correct the environmental qualification of the two reactor building emergency cooling fans after a third fan was found to not be environmentally qualified as discussed above, you also failed to take prompt corrective action to include the DHRS pump vent valves in the environmental qualification program, a deficiency that was identified during the safety system functional inspection (SSFI) of the DHRS conducted in 1993. Additionally, during several internal audits between June 1, 1992 and March 2, 1997, your staff identified concerns that the documentation to support the quality classification of components was insufficient. The Independent Safety Review (ISR) of one of those audits specifically raised concerns regarding the quality classification of components being changed without a documented or approved basis, without any independent review, and without a written safety evaluation documenting the basis for the change. Given the longstanding nature of some of these issues, and the failure to correct them, these failures represent an additional significant regulatory concern.
With respect to the violations related to emergency preparedness, which are set forth in Section IV of the enclosed Notice, the NRC observed, during the full-participation emergency exercise on March 5, 1997, that your Emergency Director failed to classify a general emergency when such a declaration was warranted due to the simulated loss of the three fission product barriers. In addition, your staff, in responding to the exercise scenario, did not assess the need for protective action recommendations (PARs) for residents beyond the 10-mile emergency planning zone (EPZ) when plume dose projections appeared to indicate that protective action guidelines would be exceeded beyond that zone. This was a result of inadequate training and procedures that did not contain guidance for considering protective action recommendations beyond the 10-mile EPZ. Following the identification of these deficiencies, a Confirmatory Action Letter was issued on March 12, 1997, setting forth prompt corrective actions needed to address the exercise weaknesses, including the performance of a remedial exercise. The training and procedural weaknesses associated with the violations identified during the exercise, as well as the deficiencies associated with your dose assessment activities and qualification of ERO personnel which were identified as Severity Level IV violations in Inspection Report 97-04 issued on June 27, 1997, indicated that management oversight and involvement in EP was insufficient. Therefore, while citations are not normally made for violations involving emergency preparedness that occur during exercises, in this case, enforcement action is appropriate because of the seriousness of the weaknesses in your emergency preparedness program revealed by the exercise. Further, the NRC was concerned that your critique of the exercise did not identify two of the four exercise weaknesses, including the failure to assess PARs beyond the 10-mile EPZ. The NRC notes that the remedial exercise was successfully conducted on May 13, 1997.
Five separate Severity Level III problems or violations are being cited for the specific violations set forth in Sections I - V.A of the enclosed Notice. The violations in each section have been classified either individually, or in the aggregate, at Severity Level III in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600. Collectively, these violations demonstrate the need for, and importance of, management actively overseeing the implementation of important program activities and assuring that personnel are self-critical and aggressively pursue identification and correction of problems. For many of the issues described herein, adequate recognition and resolution of these violations did not occur in a timely manner.
In accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000 is considered for each of the five Severity Level III violations or problems(1). Since Three Mile Island has been the subject of escalated enforcement actions within the last 2 years,(2) the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy for each of the Severity Level III violations/problems. With the exception of the environmental qualification (EQ) violation in Section V.A, credit for identification is not warranted because all but one of the violations in Sections I, II, III, and IV were identified by the NRC. Credit for corrective actions is warranted for all five violations/problems because, in general, your actions were considered prompt and comprehensive. Your initial response to the problems identified by the inspectors with respect to the equipment downgrades tended to underestimate the scope, depth, and significance of the problems; however, once you recognized the significance of the problem, you took comprehensive corrective actions to address the problems at both Three Mile Island and Oyster Creek.
Therefore, to emphasize the importance of timely identification and comprehensive correction of problems and in recognition of your previous escalated enforcement actions, I have been authorized, after consultation with the Director, Office of Enforcement, to issue the enclosed Notice of Violation and Proposed Imposition of Civil Penalties (Notice) in the total amount of $210,000 ($50,000 each for the violations in Sections I and II, and $55,000 each for the violations in Sections III and IV). No civil penalty is proposed for the violation in Section V.A in recognition of your identification and correction of the problem with the 'A' reactor building emergency cooling fan.
Overall, the penalties described above reflect the NRC concern about current station performance. The violations have brought to light weaknesses in operations and management oversight that need attention, and reflect a philosophy that has not led to aggressive identification and correction of problems in the areas cited. The fact that many of these issues were identified by the NRC, despite prior opportunities to identify and correct them, exacerbates the seriousness of these issues.
Several other violations were also identified during the design inspection and are described in Section V of the Notice. These violations are classified individually at Severity Level IV. Additionally, it was determined that the failure to update the FSAR to correct the discrepancies identified during the design inspection constituted violations of minor significance. These violations are being treated as Non-Cited Violations (NCVs) consistent with Section IV of the Enforcement Policy.
D. Smith, PDMS Manager
J. Wetmore, Manager, TMI Regulatory Affairs
Three Mile Island Nuclear Station EAs97-070; 97-117; 97-127; 97-256
During NRC inspections conducted between November 12, 1996, and May 28, 1997, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the Nuclear Regulatory Commission 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:
I. VIOLATIONS ASSOCIATED WITH DESIGN ENGINEERING
A. 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable design bases are correctly translated into specifications, drawings, procedures, and instructions.
1. Contrary to the above, on June 1, 1994, the design bases were not correctly translated into operating procedures regarding switchover of the decay heat removal system (DHRS) pumps from the borated water storage tank (BWST) to the reactor building sump. Nonconservative assumptions and input data, such as operator response times, valve stroke times, instrument errors for the BWST low-low level alarm setpoint, containment pressure, and input data for estimating vortex depth were used in calculation C-1101-212-5310-050, "TMI-1 BWST Vortex Determination." This calculation provided the BWST level setpoint specified in abnormal transient and alarm response procedures for commencing manual operations to switchover the suction of the DHRS pumps from the BWST to the reactor building sump during a large break loss of coolant accident (LOCA). The BWST level setpoint determined in the switchover phase design calculations may not have prevented air entrainment in the DHRS pumps and the reactor building spray (RBS) pumps, due to vortexing in the BWST. This could have resulted in air binding and/or cavitation of the DHRS pumps and the RBS pumps causing them to be inoperable during the critical recirculation phase of a large break LOCA. (01013)
2. Contrary to the above, as of January 10, 1997, design bases were not correctly translated into operating procedures in that the calculations that support the makeup tank pressure/level limit curves in these procedures were based on incorrect inputs. Specifically:
a. Calculation C-1101-211-5310-0047, Revision 0, did not include a section of the makeup pump suction piping and fittings, used nonconservative values for maximum high pressure injection flow rates, and did not consider the effects of vortexing.
b. Calculation C-1101-211-5360-003, Revision 1, did not consider a conservative high pressure injection flow when the reactor pressure is 0 psig during loss of coolant accidents.
The use of the inaccurate makeup tank pressure/limit curves during operation of the high pressure injection pumps after a high pressure injection line break could have caused the pumps to operate under degraded net positive suction head conditions. (01023)
B. 10 CFR Part 50, Appendix B, Criterion III, states, in part, that the design control measures shall provide for verifying or checking the adequacy of design.
1. Contrary to the above, as of January 10, 1997, field sketches contained in engineering evaluation request (EER) 88-070-E were not verified for use by engineering in an evaluation request for incorporating static head correction in the calibration of a BWST level switch. The sketches had no drawing numbers assigned, did not show plant elevations or survey marks, and had no documentation that they had been reviewed or approved. (01033)
2. Contrary to the above, between July 1988 and January 1996, safety-related calculations were performed in memoranda, Technical Data Requests (TDRs), and EERs that were not verified or approved as evidenced by the following examples each of which constitutes a separate violation:
a. Memorandum 5310-92-024, "DH-V-14A/B, DH-V-5A/B, BS-V-52A/B Valves and IST Program," dated March 1, 1994, incorporated a calculation to omit leak testing of check valves DH-V14A and DH-V14B as part of the inservice testing (IST) program. An incorrect value for reactor building (RB) pressure was used in the calculation. As a result, the analysis incorrectly concluded that these valves did not perform a safety function in the closed position. Consequently, testing of DH-V14A and B in the closed position was not included in the IST program, contrary to Technical Specification 4.2.2 which requires testing of safety-related valves in accordance with 10 CFR 50.55a and Section XI of the American Society of Mechanical Engineers and Pressure Vessel Code. (10143)
b. Memorandum 5310-92-366, "Evaluation of TMI-1 HPI SSFI Observation No. 211-10," dated December 22, 1992, incorporated a calculation to resolve a concern from GPUN's safety system functional inspection (SSFI) of the makeup and purification (MU&P) system regarding dead heading of makeup pumps under various combinations of operating pumps. (10153)
c. TDR No. 836, Revision 6, dated January 31, 1995, contained a safety-related analysis for "Evaluation for Loading of the Emergency Diesel Generator and Engineered Safeguards (ES) Buses." (10163)
d. TDR No. 995, Revision 3, "Voltage Drop Study for Degraded Grid Condition," dated January 18, 1996, contained safety-related electrical system design calculations. (10173)
e. EERs 88-060-E, "BWST Level Alarm Setpoint Change," dated July 29, 1988, and 88-070-E, "Calibration of DH-DPS-914," dated August 10, 1988, contained safety-related setpoint correction calculations for specific gravity. (10183)
These memoranda, TDRs, and EERs were not verified or approved in order to check the adequacy of the design.
C. 10 CFR 50.59 states, in part, that changes in the facility as described in the safety analysis report may be made without prior Commission approval, unless the change involves an unreviewed safety question. A change shall be deemed to involve an unreviewed safety question, in part, if the probability of malfunction of equipment important to safety previously evaluated in the safety evaluation report may be increased.
Contrary to the above, on February 20, 1990, a change that involved an unreviewed safety question was made without prior Commission approval. Specifically, the sodium thiosulfate tank was removed and the BWST low-low level alarm setpoint was changed which caused the calculated post-accident reactor building sump water level at the time of switchover to recirculation to decrease. With the reduced sump level, the available net positive suction head (NPSH) for the DHRS pumps was less than the required NPSH with indicated low pressure injection (LPI) flow of 3300 gpm unless credit was taken for containment overpressure. Updated Final Safety Analysis Report (UFSAR) Section 6.4.2 was revised to indicate that credit was taken for containment overpressure in determining that sufficient NPSH would be available for the maximum LPI flows. This was contrary to the original safety evaluation which did not consider containment overpressure in the NPSH evaluation. Without taking credit for containment overpressure the required NPSH might not be available during an accident and the probability of malfunction of the DHRS pumps may be increased. Therefore, an unreviewed safety question was involved. (01093)
These violations in Section I represent a Severity Level III problem (Supplement 1). Civil Penalty - $50,000.
II. VIOLATIONS ASSOCIATED WITH EQUIPMENT DOWNGRADES
A. 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 (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.
Contrary to the above, between September, 1992, and December, 1994, the licensee made changes to downgrade the quality classification of components and systems described in the UFSAR without performing a written safety evaluation for the majority of these changes to provide the basis for the determination that the changes did not involve a USQ. With regard to nuclear safety related component quality, section 12.3.1.2 of the UFSAR states, "Materials and parts utilized in the repair and maintenance of the nuclear-related portions of the unit will be of the same quality as, or better than, the original materials." The facility was changed by excluding the component quality assurance (QA) program requirements for the maintenance, testing, calibration, receipt inspection and procurement of parts for the following components, each of which constitutes an individual violation, without performing a written safety evaluation for any of the changes.
1. The quality classification checklists (QCLs) for the nuclear river (NR) water motor operator discharge valves NR-V-1A/1B&1C were revised on July 18, 1994, to downgrade the components from nuclear safety related (NSR) to a non-safety related classification. The valve operator is required to maintain NR system operability for postulated accident conditions as required by Technical Specification section 3.3.1.4 and as described in UFSAR section 9.6.2.3. (02013)
2. The QCLs for the decay river (DR) water strainer motors DR-S-1A&1B were revised on July 18, 1994, to downgrade the components from NSR to a non-safety related classification. The strainer is designed to automatically operate to maintain DR system operability as required by Technical Specification section 3.3.1.4 and as described in UFSAR section 9.6.2. (02023)
3. The QCLs for the auxiliary building ventilation system (ABVS) fans, fan motors, HEPA filters, flow transmitters and other components were revised on November 10 and December 22, 1994, to downgrade the components from regulatory required (RR) to "other", a non-safety related classification. The system is designed to maintain the Auxiliary Building at a negative pressure to preclude the release of radioactive material, and mitigate the consequences of the postulated waste gas tank rupture and maximum hypothetical accident as described in the UFSAR sections 9.8.3, 14.2.2.5, and 14.2.2.6. (02033)
4. The QCLs for the valve operator and positioner for make-up (MU) valve MU-V-17 were revised on March 18, 1993, to downgrade the valve components from NSR to RR. The QCLs for the valve operator and positioner for MU-V-17 were revised on November 9, 1994, to downgrade the valve components from RR to "other", a non-safety related classification. The QCL for the regulator for MU-V-17 was revised in September, 1992, to downgrade the valve component from NSR to RR. The QCL for the regulator for MU-V-17 was revised on February 1, 1994, to downgrade the valve component from RR to "other", a non-safety related classification. MU-V-17 provides a safety function to isolate the normal reactor coolant system (RCS) MU line during a postulated small break LOCA to ensure adequate core cooling as described in UFSAR section 6.1.3.1. (02043)
B. Technical Specification 6.5.1.1 requires, in part, that each procedure which affects nuclear safety, and substantive changes thereto, be prepared by a designated individual(s)/group knowledgeable in the area affected by the procedure. Each procedure, and substantive changes thereto, shall be reviewed for adequacy by an individual(s)/group other than the preparer, but who may be from the same organization as the individual who prepared the procedure or change.
Technical Specification 6.5.1.12 requires, in part, that individuals responsible for reviews performed in accordance with Technical Specification (TS) 6.5.1.1 render determinations in writing with regard to whether or not a change to a procedure which affects nuclear safety, as noted in 6.5.1.1, constitutes a USQ.
Contrary to the above, on June 16, 1993, Technical Functions Division procedure EP-011, "Methodology For Preparing The Quality Classification List," Revision 4, a procedure which affects nuclear safety, was revised and no safety evaluation was performed to determine that the change did not involve a USQ. (02053)
C. 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" requires that activities affecting quality be prescribed by documented instructions, or procedures of a type appropriate to the circumstances and be accomplished in accordance with these instructions, or procedures.
Technical Functions Division procedure EP-011, Rev. 4, "Methodology for Preparing the Quality Classification List (QCL)," provides instructions for performance of the component classification, downgrade, and upgrade processes, an activity affecting quality.
1. Section 2.2 of EP-011 states: "This procedure is applicable to hardware only and not activities, as detailed in paragraphs 2.1 and 2.2 of the Operational Quality Assurance (OQA) Plan." The OQA plan defines, in part, activities as maintenance, calibration and testing.
Contrary to the above, between the period of December 1990 through February 1996, safety related QCL activities were not accomplished in accordance with section 2.2 of EP-011 in that several QCL checklists were revised and contained changes to component activities such as maintenance, calibration, and testing. (02063)
2. Section 5.1 of EP-011 requires that QCL worksheets and checklists be prepared in accordance with Exhibit 2 of EP-011. Exhibit 2 of EP-011 requires that the criteria in Exhibit 3 of EP-011 be used to determine the functional class (classification) of components.
Contrary to the above, between the period of December 1990 through February 1996, safety related QCL activities were not accomplished in accordance with section 5.1 and Exhibit 2 of EP-011 in that the criteria in Exhibit 3 were not used to determine the functional class of some NSR and RR components. As a result, components were improperly downgraded. (02073)
These violations in Section II represent a Severity Level III problem (Supplement 1). Civil Penalty - $50,000.
III. VIOLATIONS ASSOCIATED WITH INADEQUATE CORRECTIVE ACTIONS
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. 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.
A. Contrary to the above, as of December 2, 1996, a condition adverse to quality, identified by GPUN during the safety system functional inspection (SSFI) of the decay heat removal system in 1992, had not been promptly corrected. Specifically, SSFI Observation 212-42 identified that the DHRS pump vent valves were not included in the environmental qualification (EQ) program. As of October 20, 1992, GPUN had determined that the DHRS pump vent valves had a safety function, and therefore, should have been included in the EQ program. However, as of December 2, 1996, these valves were not included in the EQ program. (03013)
B. Contrary to the above, between June 1, 1992, and March 2, 1997, identified deficiencies in the supporting documentation for the safety classification of components were not promptly corrected. Specifically:
- In audit report O-COM-91-13, finding 1, dated June 1, 1992, the licensee identified that procedure EP-011 or the Quality Classification List worksheet did not always provide sufficient details to support the classification of items without recourse to the originator. This significant condition adverse to quality was not corrected until May 25, 1993.
- In audit report O-COM-95-09, finding 12, dated December 21, 1995, the licensee identified that EP-011 did not define how to prepare a revision to a Quality Classification List Checklist and did not require that the bases for revisions be documented so those other than the Quality Classification List engineer who prepared the change could understand the basis of the change. This significant condition adverse to quality was not corrected as of March 2, 1997.
- In the Independent Safety Review (ISR) of audit report O-COM-95-09, initiated on May 16, 1996, the Independent Safety Reviewer identified the concern that changing the quality classification of components without documented or approved basis, independent review, and without a written safety evaluation documenting the basis of the change may be a violation of 10 CFR 50.59. This significant condition adverse to quality was not corrected as of March 2, 1997.
The corrective actions for the findings of O-COM-91-13 were ineffective in that they did not preclude repetition of the problem of insufficient documentation to support the QCL activities. (03023)
C. Contrary to the above, from March 21, 1997 until April 24, 1997, the licensee failed to take prompt and adequate corrective action for a condition adverse to quality. Specifically, on March 21, 1997, the licensee identified that reactor building emergency cooling fan, AH-E-1A was not environmentally qualified, in that the application of heat shrink tubing left a small length of exposed conductor at the spark plug connector to the fan motor. The licensee failed to conduct sufficient additional reviews to identify and resolve the similar condition for the other two reactor building emergency cooling fans, AH-E-1B and AH-E-1C, until April 24, 1997. As a result, reactor building emergency cooling fans, AH-E-1B and AH-E-1C, were inoperable contrary to TS 3.3. (03033)
These violations in Section III represent a Severity Level III problem (Supplement 1). Civil Penalty - $55,000.
IV. VIOLATIONS ASSOCIATED WITH EMERGENCY PREPAREDNESS
10 CFR 50.54(q) states, in part, that "a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b) and the requirements in Appendix E of this part."
A. 10 CFR Part 50, Appendix E, Section IV.B, "Assessment Actions," requires, in part, that the means to be used for determining the magnitude of and for continually assessing the impact of the release of radioactive materials be described, including the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety.
The GPU Nuclear Corporate Emergency Plan, which was developed in accordance with the provision of 10 CFR Part 50, Appendix E and 10 CFR 50.47, Section 5.1.3.1, "Direction and Coordination," states, in part, that the Emergency Director is vested with certain authority and responsibility that shall not be delegated to a subordinate, including classification of emergency events.
The licensee's Emergency Plan Implementing Procedure (EPIP) TMI-.01, "Emergency Classification and Basis," Item G4.2, requires the declaration of a general emergency for the loss of two of three fission product barriers with a potential loss of the third barrier.
Contrary to the above, during the full-participation exercise on March 5, 1997, the EPIP procedure TMI-.01 was not followed in that the Emergency Director failed to classify a general emergency when such a declaration was warranted due to the simulated loss of the three fission product barriers. (04013)
B. 10 CFR 50.47(b)(10) requires, in part, that a range of protective actions have been developed for the plume exposure pathway emergency planning zone (EPZ) for emergency workers and the public. Guidelines for the choice of protective actions during an emergency, consistent with Federal guidance, are developed and in place.
Section 2.0 of the GPU Nuclear Corporate Emergency Plan states, in part, that the Emergency Plan is consistent with the guidelines given in NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," dated November, 1980.
Section I.D.2 of NUREG-0654 states, in part, that for the worst possible accidents, protective actions would need to be taken outside the planning zones.
Contrary to the above, as of March 5, 1997, emergency response training was not adequate and procedures contained insufficient guidance for considering protective action recommendations (PARs) beyond the 10-mile EPZ. As a result, emergency response management did not communicate recommendations for PARs for residents beyond the 10-mile EPZ when plume dose projections appeared to indicate that protective action guidelines would be exceeded beyond that zone during the full participation exercise on March 5, 1997. (04023)
These violations in Section IV represent a Severity Level III problem (Supplement 1). Civil Penalty - $55,000.
V. VIOLATIONS NOT ASSESSED A CIVIL PENALTY
A. 10 CFR 50.49(f) requires each item of electrical equipment important to safety to be environmentally qualified by testing or by combination of testing and analysis.
10 CFR 50.49(j) requires that a record of the environmental qualification be maintained in an auditable form to permit verification that each item of electric equipment important to safety is qualified for its application and meets its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function.
Contrary to the above, from March 17, 1986, until March 24, 1997, the three reactor building emergency cooling fans were not environmentally qualified by testing and/or analysis for post loss of coolant accident (LOCA) reactor building atmospheric conditions. Specifically, the application of heat shrink tubing left a small length of exposed conductor at the spark plug connector to each of the fan motors. There was no documentation to support the qualification of the motors with the exposed conductors in the post-LOCA environment in which they would be required to function. As a result, the reactor building emergency cooling fans were inoperable contrary to TS 3.3. (05013)
B. 10 CFR Part 50, Appendix B, Criterion V, states, in part, that activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings.
Procedure SP-9000-44-001, "Instrument and Control Instrument Installation," Revision 0, Section 3.3, specifies the use of protective barriers where separation criteria cannot be met.
Procedure 1420-HT1, "Heat Trace Repair and Replacement," Revision 11 and the vendor drawing ET-30250, Revision 2, provide instructions for the installation of heat tracing.
Contrary to the above, as of December 2, 1996, the as-installed configuration of BWST level transmitter DH-LT-808 was not in accordance with the applicable specifications, procedures, and drawings. Specifically:
1. The protective barrier for level transmitter DH-LT-808 was not intact in that the cover plate was open and the fasteners for the cover were missing.
2. The heat tracing for level transmitter DH-LT-808 was not installed in accordance with procedure 1420-HT1 and the vendor drawing in that the heat tracing was not wrapped around the sensing lines or the transmitter.
Without the proper electrical separation between redundant components or proper heat tracing, the level transmitter could have been subject to a single failure or freezing. (06014)
C. 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
1. Contrary to the above, as of January 10, 1997, the test program failed to assure that the tests specified by IEEE Report No. NSG/TSC/SC4-1, "Proposed IEEE Criteria for Class 1E Electrical Systems for Nuclear Power Generating Stations." June 1969, were identified and performed. Specifically, molded case circuit breakers in safety-related motor control centers 1A ES ESF VENT and 1B ESF VENT had not been tested since they were installed in 1986 contrary to IEEE Report No. NSG/TSC/SC4-1 which specifies that tests shall be performed at scheduled intervals to demonstrate that components that are not exercised during normal operation are operable. UFSAR Section 8.1, "Design Basis," states that the electrical system design satisfies IEEE Report No. NSG/TSC/SC4-1. (07014)
2. Contrary to the above, as of December 2, 1996, the test program failed to assure that testing of safety related check valves required by TS 4.2.2 in accordance with 10 CFR 50.55a and Section XI of the American Society of Mechanical Engineers (ASME) and Pressure Vessel Code (The Code), was identified and performed. Specifically, the DHRS pump discharge check valves (DH-V16A and B) and the ball check valves located in the floor drains of the DHRS pump vaults were not tested in the closed position as specified by part 10 (OM-10) of ASME/ANSI ONa-1988 which is referenced by Section XI of the ASME Code. DHRS pump discharge check valves have a safety function for preventing back flow thus maintaining low pressure injection flow when the discharge header cross-connect valves are open. Ball check valves located in the floor drains of the DHRS pump vaults have a safety-related function in preventing flooding of the other DHRS vault if excessive water leakage occurs in one vault. (08014)
Pursuant to the provisions of 10 CFR 2.201, GPU Nuclear Corporation (Licensee) is hereby required to submit a written statement or explanation to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the receipt 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.
1. The violations in Sections I and II occurred prior to November 12, 1996, which is the date that the base amount for a Severity Level III violation or problem changed from $50,000 to $55,000. Therefore, the base penalty for the violations in Sections I and II is $50,000.
2.e.g., A Notice of Violation without a civil penalty was issued on March 26, 1996 for a repetitive violation of security requirements (EA 96-057), and a Notice of Violation without a civil penalty was issued on March 11, 1996 for a violation involving the failure to adequately control a modification to the reactor coolant system drain line piping (EA 95-238).