ML20070E086
| ML20070E086 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/06/1994 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD940627, NUDOCS 9407140058 | |
| Download: ML20070E086 (6) | |
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COOPER NUCLEAR STATION P.O. BOX 98, BRoWNVILLE, NEBRASKA 68321 Nebraska Public Power District
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M cu NSD940627 July 6, 1994 U.
S.
NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555
Subject:
Reply to Notice of Violation NRC Inspection Report No. 50-298/94-04 Cooper Nuclear Station, NRC Docket 50-298, DPR-46 Gentlemen:
The Nebraska Public Power District (District) hereby submits its response to the Notice of Violation (NOV) transmitted with NR~C Inspection Report No.
50-298/94-04.
This inspection report documents the results of the NRC Service Water System Operational Performance Inspection, conducted by your office on March 28 through April 1, and April 11-15, for Cooper Nuclear Station (CNS). The NRC identified two violations, each containing two examples, during the inspection.
In accordance with 10 CFR 2.201, an explanation of the violations and corrective actions taken and planned in response to each violation are presented below.
Additionally, the subject inspection report contained an unresolved item (298/9404-03) concerning the performance testing of the Reactor Equipment Cooling System heat exchangers. The District is evaluating this issue and will have all information finalized and available for review by August 31, 1994.
STATEMENT OF VIOLATION A.
10 CFR 5 0. 71 (e ) states in part that each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report to assure that the information included in the final safety analysis report contains the latest material developed.
The updated safety analysis report (USAR) shall be revised to include the effects of all changes made in the facility or procedures as described in the final safety analysis report and all safety evaluations performed by the licensee in support of conclusions that changes did not involve an unreviewed safety question.
USAR Appendix G Figure G-6-1 shows the residual heat removal service water system as an essential safety system auxiliary to shutdown cooling. USAR Section 8.2.5 describes the residual heat removal service water booster system as maintaining the service water side of the residual heat removal heat exchangers at a higher pressure than the residual heat removal system side to prevent out-leakage of radioactive water into the service water systg USAR Section 8.2.6 descri.bes that when the res; ',ual heat removal 9407140050 940706 kI
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U.S. Nuclear Regulatory Commission July 6, 1994-Page 2 of 6 system is in the shutdown cooling mode, that the service water booster pumps are started.
Contrary to the above, the District failed to revise the USAR, for CNS, to include the effects of:
1.
A safety evaluation performed by the licensee on June 6, 1990, in support of conclusions that the change-in the desicin basis temperature from 85aF to 900F did not involve an unreviewed safety question.
Although USAR Section 5.3 mentioned that additional evaluation had shown that adequate nec positive suction head existed with 90*F service water temperature, the USAR continued to present analysis results based on 85'F, including:
(1).USAR Figure VI-5-15 which presented the minimum containment pressure to assure adequate net positive suction head for the core sprey cooling pumps; (2) the maximum suppression pool temperature of 192*F; and, (3) USAR Drawings IV-8-1, VI-4-2, VI-4-3, and X-8-2=.
2.
A safety evaluation performed by the licensee on April 7, 1984, in support of conclusions that the change to not operate the service water booster pumps in shutdown cooling, and consequently not maintain service water system higher than residual heat removal system pressure, did not involve an unreviewed safety question.
This is a Severity Level IV violation (298/9404-01) (Supplement 1).
q REASON FOR VIOLATION 1.
The original analyses performed to support changing the licensing basis river temperature from 85'F to 90*F did not include all the loss of coolant accident (LOCA) cases (A through E) in Chapter XIV of the UMR, but only considered the limiting case E.
This is because the analysis was performed to support a 1989 Justification for Continue $ Operation (JCO) as opposed to supporting a design change.
While the JCO was not intended to be temporary in nature, it was intended to address only the most limiting scenario' (Case E).
This analysis was essentially a sensitivity study to ensure that, with the change from 850F to 90'F river temperature, the accident analysis was still bounding.
Upon conclusion that the accident analysis was still bounding, the statement was added to the USAR that'the plant was evaluated for 90*F river temperature.
This course was chosen because the entire accident analysis was not revised.
Subsequent to this decision in 1989, the Design Criteria Document (DCD) was prepared for the SW System as part of the Design Basis Reconstitution effort.
The Design Basis Group 'vas aware of the ongoing work for Design Change (DC)91-144, "RHR Hx Tube Plugging Increase" and associated changes to the USAR.
Under this DC, the accident analyses were being completely revised at a 90*F river water temperature.
In an effort to ensure that the 90*F limit was stated as the design basis, the'DCD was revised to clearly identify the river temperature limit as 90*F.
However, approval of DC 91-144
+
U.S. Nuclear Regulatory Commission July 6, 1994 Page 3 of 6 i
did not occur until June 1, 1994.
The Design Basis Group (DB) did not identify the USAR as a discrepancy because: 1) the DB Group was aware of the situation involving the JCO in 1989 and 2) the DB Group j
was aware of the upcoming change to the USAR as a result of DC 91-144.
2.
The District failed to recognize that operating in the Shutdown Cooling Mode, without operating the Service Water Booster pumps, was a change to the Updated Safety Analysis Report (USAR).
The 10 CFR 50.59 analysis for Special Procedure 84-002 did correctly evaluate-the safety aspects of the proposed change in station operation.
However, when Procedure 2.2.70 was subsequently revised to add the i
new operating mode, a change to the USAR was not made.
CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED 1.
License Change Request (LCR)94-033 for DC 91-144 was submitted to Nuclear Licensing on April 11, 1994.
DC 91-144 was SORC approved June 1,
1994.
The LCR, in combination with the design change, resulted in changes to the USAR that reflected the 90 F SW value for all of the LOCA cases.
Drawing Change Notices (DCNs) were issued for each of the drawings illustrated in Figures IV-8-1,
.VI-4-2, VI-4-3, and X-8-2.
Even Lnough the drawings illustrated in these USAR figures have been updated by DCNs, these USAR figures will not be updated until the July 1995 update.
USAR updates for DC'91-144 will be consistent and give SW maximum river temperature as 90 F.
2.
The District plans to initiate an LCR, along with an updated 10 CFR 50.59 engineering analysis, and revise the USAR to adequately describe the windmilling operation mode of the RHR SW Booster Pumps.
In addition, the District will revise the SW DBD to reflect the operation of the RHR SW Booster Pumps in the windmilling mode.
CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS 1
The root cause of the problem which led to confusion in the USAR as to the maximum river temperature, and the failure to updated several USAR figures, was the use of a safety analysis in the Justification for Continued Operation (JCO) to change the design basis of the plant.
A design change should have been performed in conjunction with the JCO to ensure that the appropriate design documents (calculations, drawings, procedures, etc.)
were evaluated and updated.
A revision will be made to plant procedures to prohibit the use of Operability Evaluations alone to change the design basis of the plant. A design change, or other appropriate documentation, will be required to accompany the JCO,' if it affects the plant design and licensing
- basis, 2.
The currently existing Procedure 0.4,
" Procedure Change Process",
has provisions to ensure that procedure changes that af fect the USAR are reflected in the USAR.
Since the described failure to update
- ... U.S. Nuclear Regulatory Commission July 6, 1994 Page 4 of 6 the USAR occurred, major procedure upgrades have been made to this procedure which specifically address the steps to be followed for procedure changes which affect the USAR.
The District also recognizes a weakness in the Design Basis Reconstitution Program identified by the NRC Inspectors during this inspection.
That is, the Validation and Verification (V & V) process focuses almost entirely on plant design, and does not include a thorough evaluation of the plant operation.
The District intehds to strengthen the V & V process by increasing the scope and depth of the review of plant operating procedures.
This will improve the District's capability to identify any disconnect between the design basis and-plant operation.
DL WHEN FULL COMPLIANCE WILL BE ACHIEVED 1.
The District will complete revisions to plant procedures to prohibit the use of Operability Evaluations to perform changes to the design basis of the plant by August 31, 1994.
2.
The updated 10 CFR 50.59 engineering analysis to reflect the windmilling operation of the SW Booster Pumps, will be approved by August 31, 1994.
The DCD for the SW System will be revised by August 31, 1994.
STATEMENT OF VIOLATION B.
Criterion III, requires that measures shall-be established to assure that applicable regulatory requirements and the design basis, as defined in Section 50.2 and 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.
Contrary to the above:
1.
Established measures did not assure that the change in service water design basis temperature from 85"F to 90 F.was correctly' translated into Calculation 91-256, datea September 16, 1991; Isometric Drawing 2852-3, Revision 5; Procedure 13.15.1,
" Reactor Equipment Cooling Heat Exchanger Performance Analysis; " Procedure 13.17,
" Residual Heat Removal Heat Exchanger Performance Evaluation;" and Procedure 13.18, "D
Jacket Water and Lube Oil Heat Exchanger Performance Evaluation."
2.
Established measures did not assure that the design basis, as specified in the general design criteria document for internal flooding, to qualify fire protection system piping in the service water system pump room to Class I (seismic) standards, was correctly translated into specifications and drawings.
This is a Severity Level IV violation (298/9404-02) (Supplement 1)'.
o U.S. Muclear Regulatory Commission Jul3 0,
1994 PGge 5 of 6
- REASON FOR VIOLATION 1.
Generally, Justification for Continued Operation (JCOs), or similar documents, do not result in a permanent change to the facility.
However, in this particular situation, a change to the facility as described in the USAR, i. e, the maximum river temperature, resulted in a permanent change. The method for preparing JCOs (included in Procedure 0.27.1, Operating Evaluations), does not require the evaluation of the impact on the design input documents, therefore resulting in the above violation.
2.
The District was unable to find calculations which demonstrated that the Fire Protection line in the Intake Structure was qualified to Seismic Class IS standards, and drawings of the system incorrectly identified the system as IIS. The District had committed to qualify the FP lines to IS standards as a part of original plant licensing.
It was apparent from the supports on the lines that the piping was in all likelihood Class IS supported.
However, due in part to the fact that CNS was issued a construction permit prior to 10 CFR 50, Appendix B requirements, not all of the plant records of this period are available.
This is one of the main reasons for the voluntary District effort to reconstitute the design basis. As a part of the Design Basis Reconstitution effort, the commitment to qualify this FP line to IS standards was identified. However, as pointed out by the NRC inspectors during this audit, this commitment was not followed through in a V & V effort, as would be the case with a safety system.
The current program does not include V & V of the topical Design Criteria Document.
This weakness in the program prevented the District from correcting the lack of documented evidence necessary to prove that the FP line is Class IS.
CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED 1.
The District will perform a document search / upgrade to ensure the 90 F Service Water design temperature is properly incorporated into all applicable design input documents.
2.
The District has analyzed the fire protection piping in the Service Water Pump Room to the current Class I standards. The District has also included the possible effects of microbiologically inf1_enced corrosion (MIC) in the seismic analysis.
These activities were completed April 25, 1994.
The SQUG walkdown recommendation to restrain the air handling units in the Service Water Pump Room will be completed prior to startup from the 1995 refueling outage.
CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS 1.
The District will make procedure changes to insure that a JCO, or other similar document, which'affects the design and licensing basis of the plant, will be followed up with a design change, or other
U.S. Nuclear Regulatory Commission July 6, 1994 Page 6 of 6 appropriate documentation, to ensure that drawings and procedures are updated on a timely basis.
The DC process has the appropriate checklists and design engineering guidelines to ensure that design input documents are captured.
2.
The District will include V & Vs of topical Design Criteria Documents in the Design Basis Reconstitution Program.
This will preclude a non-safety system specific commitment from being
)
overlooked in the V & V process.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED 1.
The District expects to have completed the document search / upgrade and make the necessary procedural revisions by August 31, 1994.
r 2.
The Design Basis Program will be revised to include V & V of topical DCDs by October 1, 1994.
A V & V effort of the Internal Flooding DCD, the only topical document completed to date, will be completed c
by June 30, 1995.
Should you have any questions or require any additional information, please contact me.
Sin erely, W
Vi President - Nuclear GRH/WLS/dnm cc:
NRC Regional Administrator Region IV Arlington, TX NRC Resident Inspector Cooper Nuclear Station f
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