EA-97-017 - Cooper (Nebraska Public Power District)
G. R. Horn, Senior Vice President
of Energy Supply
Nebraska Public Power District
1414 15th Street
Columbus, Nebraska 68601
SUBJECT: NOTICE OF VIOLATION AND EXERCISE OF ENFORCEMENT DISCRETION (NRC Inspection Report Nos. 50-298/96-24 and 96-31)
Dear Mr. Horn:
This refers to inspections conducted at the Nebraska Public Power District's (NPPD) Cooper Nuclear Station which were documented in the subject NRC inspection reports. These inspections identified several apparent violations related to the accuracy of Cooper's Updated Safety Analysis Report (USAR) and the adequacy of certain analyses performed under the provisions of 10 CFR 50.59. The results of the inspections and our concerns were discussed with Mr. Phil Graham and others of your staff on February 19, 1997. The inspection reports were issued on February 25, 1997, and February 28, 1997, respectively. A predecisional enforcement conference was held in Arlington, Texas on April 15, 1997, to discuss the apparent violations. The issues discussed during the predecisional enforcement conference also included a review of NPPD's January 24, 1997, letter (reference the NRC's letter to you dated February 28, 1997), and a review of whether Cooper's LER 96-14-01 identified a loss of control of the license basis (reference NRC's letter to you dated April 8, 1997). The NRC issued a meeting summary related to the topics discussed at the predecisional enforcement conference in a letter to you dated May 14, 1997.
Based on the information developed during the inspections and the information that you provided during the conference, as well as communications with your staff on specific issues following the conference, the NRC has determined that violations of NRC requirements occurred. These violations are cited in Enclosure 1, the Notice of Violation (Notice), and the circumstances surrounding them were described in detail in the subject inspection reports and above referenced letters. The first violation contains eight examples of a failure to update the USAR as required by 10 CFR 50.71(e). The second violation contains three examples of a failure to perform adequate written safety evaluations in accordance with 10 CFR 50.59.
During the conference, your staff contested five of the examples that the NRC had identified as apparent violations. The NRC has reviewed NPPD's reasons for contesting these five examples, and has decided in three cases not to include the examples in the Notice of Violation. Our basis for this decision and our decision that violations did occur in the remaining two contested examples, are discussed in detail in Enclosure 2.
In 1996, your staff determined that Cooper's USAR had many inaccuracies. However, your staff delayed in addressing the problem such that by the time of our inspection in October 1996, our inspectors identified further examples of USAR discrepancies in addition to those which your staff had identified and had not yet corrected. None of the issues impacted current operability, and the safety significance for these violations is low. Nonetheless, in the aggregate, these two violations are of regulatory significance because they represent a programmatic failure to meet the requirements of 10 CFR 50.71(e) and 10 CFR 50.59. This shows a significant lack of attention to detail on your part and has resulted in our current regulatory concern about the accuracy of the USAR. Therefore, these violations are classified in the aggregate in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 as a Severity Level III problem.
In accordance with the Enforcement Policy, a base civil penalty in the amount of $50,000 is considered for a Severity Level III problem. Because your facility has been the subject of escalated enforcement actions within the last 2 years1, 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. The NRC has determined that credit is not warranted for identification because (1) although your staff identified the general problem of USAR inaccuracies in May 1996, the issues were not addressed in a timely manner, and (2) most of the specific issues that were the subject of the predecisional enforcement conference were identified by the NRC. However, the NRC determined that credit is warranted for Corrective Action. This determination is based on the numerous and comprehensive corrective actions you have undertaken to address the underlying cause of the problem. Some of the corrective actions include: the USAR has been changed or clarified for the identified items; an audit has been conducted on-site to verify the accuracy of the USAR; Cooper management has increased the focus on using Problem Identification Reports (PIRs) to identify USAR inconsistencies; training has begun on conducting evaluations required by 10 CFR 50.59; additional management oversight of the evaluations will be provided; Cooper's procedures for performing the evaluations have been upgraded; Cooper has developed an action plan to review potential unauthorized modifications that have resulted from past maintenance activities; and Cooper has developed a USAR rebaselining project, which includes developing a database of design and licensing basis documentation that will facilitate performance of more comprehensive 10 CFR 50.59 evaluations.
The normal civil penalty assessment process in this case would result in a $50,000 civil penalty being proposed. However, after consultation with the Director, Office of Enforcement, I have been authorized not to propose a civil penalty in this case in accordance with the provisions provided in Section VII.B.6 of the NRC's Enforcement Policy. This decision is based on: (1) consideration of the generally low safety significance of the violations; (2) the comprehensiveness of NPPD's corrective actions; (3) the fact that the inspections were occurring at about the same time the NRC's Enforcement Policy was revised to place additional emphasis on USAR accuracy problems; and (4) our recognition that communications with the NRC may have inadvertently contributed to delaying your USAR upgrade program2. However, significant violations in the future could result in a civil penalty.
You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response will be placed in the NRC Public Document Room (PDR).
Sincerely, org signed by James E. Dyer for Ellis W. Merschoff Regional Administrator
Docket No. 50-298
License No. DPR-46
Enclosures: As stated
John R. McPhail, General Counsel
Nebraska Public Power District
P.O. Box 499
Columbus, Nebraska 68602-0499
P. D. Graham, Vice President of
Nebraska Public Power District
P.O. Box 98
Brownville, Nebraska 68321
B. L. Houston, Nuclear Licensing
and Safety Manager
Nebraska Public Power District
P.O. Box 98
Brownville, Nebraska 68321
R. J. Singer, Manager-Nuclear
907 Walnut Street
P.O. Box 657
Des Moines, Iowa 50303
Mr. Ron Stoddard
Lincoln Electric System
11th and O Streets
Lincoln, Nebraska 68508
Randolph Wood, Director
Nebraska Department of Environmental
P.O. Box 98922
Lincoln, Nebraska 68509-8922
Nemaha County Board of Commissioners
Nemaha County Courthouse
1824 N Street
Auburn, Nebraska 68305
Cheryl Rogers, LLRW Program Manager
Environmental Protection Section
Nebraska Department of Health
301 Centennial Mall, South
P.O. Box 95007
Lincoln, Nebraska 68509-5007
Dr. Mark B. Horton, M.S.P.H.
Nebraska Department of Health
P.O. Box 950070
Lincoln, Nebraska 68509-5007
R. A. Kucera, Department Director
of Intergovernmental Cooperation
Department of Natural Resources
P.O. Box 176
Jefferson City, Missouri 65102
Kansas Radiation Control Program Director
Nebraska Public Power District Docket No. 50-298 Cooper Nuclear Station License No. DPR-46 EA 97-017
During NRC inspections conducted on October 7 through February 19, 1997, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:
A. 10 CFR 50.71(e) requires that the Updated Safety Analysis Report be updated periodically to assure that the information in the Updated Safety Analysis Report contains the latest material developed. 10 CFR 50.71(e)(4) requires that revisions be filed annually or 6 months after each refueling outage provided the interval between successive updates 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, the licensee did not update the USAR within the required timeframe for the following examples, each of which constitutes a separate violation.
- As of November 1, 1996, Updated Final Safety Analysis Report Section XII-18.104.22.168.2, "[Seismic Analysis] Piping," and Updated Safety Analysis Report, Appendix C, "Structural Loading Criteria," Section 22.214.171.124, "Piping Seismic Analysis," was not updated to accurately reflect the seismic analysis practices at the time. Since initial construction of the facility, these sections of the Updated Safety Analysis Report (and the Final Safety Analysis Report), have described, in detail, the procedure for dynamically analyzing Class-I seismic piping systems without restricting the requirement for dynamic analysis to large bore piping. However, as of November 1, 1996 (and since initial construction), the dynamic seismic analysis described in the Updated Safety Analysis Report was not performed for 2-inch and smaller piping systems.
- As of November 1, 1996, Updated Safety Analysis Report, Section III-9.3, "[Standby Liquid Control System] Description," was not updated to accurately reflect the expected room temperatures for the standby liquid control system and the controls in place to ensure safe operation with a room temperature of 50·F. Updated Safety Analysis Report, Section X-10.3.2 "[Heating, Ventilation and Air Conditioning Systems] Station Heating System," states that winter design temperatures for the system are given in Table X-10-1. Table X-10-1, "[Heating, Ventilation and Air Conditioning Systems] Station Heating System Design Temperatures
(Winter)," states that the normal minimum indoor temperature for the reactor building is 50·F. The equipment containing the solution is installed in a room in the reactor building. However, as of November 1, 1996 (and since initial construction), Updated Safety Analysis Report, Section III-9.3, "[Standby Liquid Control System] Description," stated that, "The equipment containing the solution is installed in a room in which the air temperature is to be maintained within the range of 65·F to 100·F."
- As of November 1, 1996, Updated Safety Analysis Report, Section III-9.4, "[Standby Liquid Control System] Safety Evaluation," was not updated to correctly describe the safety basis for the standby liquid control system relief valves. Design Change 86-34A, "SLC/ATWS Modifications," Revision 0, dated March 4, 1988, changed the safety basis for the standby liquid control system relief valve settings from:
assuring injection into the reactor above the normal [emphasis added] pressure of approximately 1030 psig in the bottom of the reactor,
assuring injection into the reactor above the anticipated transient without scram reactor pressure conditions, which would equal the reactor safety/relief valves' setpoints plus the accumulation at the maximum anticipated transient without scram steam flow, (i.e., approximately 1100 psig plus the static head in the reactor vessel).
Specifically, Section III-9.4 continued to state that "The SLC system and pumps have sufficient pressure margin, up to the allowed system relief valve setting range of 1450 to 1680 psig, to assure solution injection into the reactor above the normal [emphasis added] pressure of approximately 1030 psig in the bottom of the reactor."
- As of November 1, 1996, Updated Safety Analysis Report, Section III-9.3, "[Standby Liquid Control] Description," was not updated to be consistent with Technical Specification Figure 3.4.2. The Updated Safety Analysis Report states that, at the minimum room temperature of 65·F, the maximum permitted solution concentration is 12.5 weight percent. Section III-9.3 also states that a concentration of 11.5 percent corresponds to an adjusted saturation temperature of 61·F. The Updated Safety Analysis Report adjusted saturation temperature includes a 10·F margin over saturation, which corresponds to the definition for the Technical Specification minimum allowable temperature. However, Technical Specification Figure 3.4.2, "Percent Sodium Pentaborate by Weight of Solution versus Temperature," indicates that at 65·F, the maximum permitted concentration was 12.1 percent. At 11.5 percent concentration, the minimum allowable temperature was 62·F.
- As of November 1, 1996, Updated Safety Analysis Report Section IV-9.3, "[Reactor Water Cleanup System] Description," was not updated to clearly indicate the effect of a modification on the reactor water cleanup system isolation valves' control logic. Further, neither Section IV-9.3 nor Section III-9.3, "[Standby Liquid Control System] Description," were updated to indicate that following the modification it was always necessary to operate both SLC trains to close both motor operated valves and maintain comparable defense against a single failure of the reactor water cleanup isolation valves.
At the time of the inspection, Updated Safety Analysis Report, Section IV-9.3, "[Reactor Water Cleanup System] Description," stated that, "In the inlet piping to the cleanup recirculation pumps, two motor operated isolation valves, one on either side of the primary containment, are automatically closed by... standby liquid control system actuation." Design Change 86-34A, "SLC/ATWS Modifications," Revision 0, March 4, 1988, changed the reactor water cleanup system isolation valves' control logic such that initiation of one train of the SLC system no longer closed both motor operated valves. In order to achieve comparable defense against a single failure of the reactor water cleanup isolation valve, the licensee implemented administrative procedures which require the operators to always operate both trains of the standby liquid control system.
- In 1994, during the surveillance test validation program status review, the licensee identified at least two discrepancies in the Updated Safety Analysis Report, which were not corrected in the July 22, 1996 update to the Updated Safety Analysis Report.
(a) Updated Safety Analysis Report, Table VII-3-1, "Pipeline Penetrating Containment," Note 4 incorrectly stated that the control rod drive system solenoid valves open during a reactor scram. On reactor SCRAM the solenoid valves remain closed and the air-operated SCRAM valves open to insert the control rods and to exhaust water to the SCRAM discharge volume.
(b) Updated Safety Analysis Report, Section VII-4.5.44, "[Core Spray System Control and Instrumentation] Core Spray Valve Control," incorrectly stated that two pressure switches monitor system pressure (for the low pressure permissive). In addition, it indicates that either switch can initiate opening of the discharge valves for core spray. There actually are four pressure switches designed in a 1-out-of-2 twice logic and a minimum of two switches are required to actuate to initiate opening of the core spray valves.
- As of March 16, 1996, Updated Safety Analysis Report, Table V-2-2, "Penetration Schedule," pages V-2-9 to V-2-12, was not updated to correctly list all the penetrations; the quantity of lines in three penetrations; and line descriptions in five penetrations.
- As of May 4, 1996, Updated Safety Analysis Report, Table V-2-7, "Testable Primary Containment Isolation Valves," pages V-2-44 to V-2-46, did not list 23 penetrations (X20, X-30E and -30F, X-33E and -33F, X-35A through E, X-45D, and X-229A through L) and their associated valves.
B. 10 CFR 50.59(b)(1) requires that the licensee maintain records of changes in the facility and of changes in procedures made pursuant to this section, to the extent that these changes constitute changes in the facility as described in the safety analysis report or to the extent that they constitute changes in the procedures as described in the safety analysis report. Further, 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.
Contrary to the above, the licensee either did not perform the required written safety evaluation or performed an inadequate safety evaluation as shown in the following examples, each of which constitutes a separate violation.
- Updated Safety Analysis Report, Section X.8.2.8.C, "Common Mode Failure Analysis - Fire," required that combustibles not be located in the service water booster pump room area since both trains of the service water system were located in close proximity. However, Procedure 0.7.1, "Control of Combustibles," Revision 6, allowed up to 90 pounds of wood or 5 gallons of flammable liquid for this area of the plant. In addition, on December 2, 1996, combustible materials (rags, papers, and flammable chemicals) were located in the service water system booster pump room. The written safety evaluation for this change in procedure was inadequate, in that, a common mode failure analysis had not been performed to justify the presence of combustible materials in the service water booster pump area, and therefore, the safety evaluation did not provide a bases for the determination that the change does not involve an unreviewed safety question.
- Updated Safety Analysis Report, Section XII.126.96.36.199, "Intake Structure," states, in part, that in order to keep ice away from the intake structure during cold weather, an ice deflector is installed during the winter months. Although a portion of the ice deflector was installed on December 18, 1996, the deflector had not been fully installed at any time during the winter months. The failure to fully install the ice deflector by the winter months was a configuration changes that had not been evaluated, through a written safety evaluation, as a change to the facility.
- Updated Safety Analysis Report, Section IV.10.3, "Nuclear System Leakage Rate Limits - Description," states, in part, that each containment drywell sump has an alarm system and automatic starting sequence on rising water level. Both containment drywell sumps are equipped with a fill rate timer and alarm. This alarm can be set at or below the Technical Specification limits and would provide immediate indication when this preselected rate is reached or exceeded. However, the safety evaluation dated December 20, 1996, that addressed the failure of the automatic pump starting system, and the failure of the sump fill rate timer and high level alarm, was inadequate in that it did not address the lack of control room alarm. A separate written safety evaluation did not exist for this change to the facility.
These violations represent a Severity Level III problem (Supplement I) (50-298/96024-14).
Pursuant to the provisions of 10 CFR 2.201, Nebraska Public Power District is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region IV, and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. 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. Where good cause is shown, consideration will be given to extending the response time.
Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.
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 Arlington, Texas
this 25th day of June 1997
1.The most recent case involved a Severity Level III violation issued on December 20, 1996 (EA 96-488), with no civil penalty, for a failure to meet station blackout requirements.
2.NPPD planned to update the USAR to include all safety basis information and eliminate nonsafety basis information using 10 CFR 50.59. NRC concerns about the acceptability of removing nonsafety basis information (i.e., the basis for using 10 CFR 50.59 was being questioned) resulted in NPPD placing the project on hold pending resolution of these questions. During a May 9, 1996 meeting (reference NRC Memorandum dated May 20, 1996), the licensee informed the NRC that the overall USAR upgrade program was on hold. Additional guidance on the use of 10 CFR 50.59 to remove information from the USAR was not available at the time of the meeting.