ML19326A576
| ML19326A576 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/21/1978 |
| From: | Jeffery Grant TOLEDO EDISON CO. |
| To: | Fiorelli G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML19326A575 | List: |
| References | |
| 1-13, NUDOCS 8002050609 | |
| Download: ML19326A576 (2) | |
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\\~s' R)LEDO License No. NPF-3 I!E)lb5(ION Docket No. 50-346 JAMES S. GAANT v.c........m o..., sow, (4191 259 *232 February 21, 1978 Serial No. 1-13 Mr. Gaston Fiorelli, Chief Reactor Operations Branch Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Fiorelli:
Toledo Edison acknowledges receipt of your January 27, 1978, letter and report enclosure 78-01 referencing an apparent deviation from Davis-Besse Nuclear Power Station commitments to the NRC, listed as an " Infraction" under the heading " Notice of Violations".
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Following a thorough examination of the item of concern, Toledo Edison s_,-
herein offers information regarding this item, including actions and the dates of corrective actions.
Infraction:
Technical Specification 6.8.1 requires that written procedures shall be established, implemented and maintained.
Contrary to the above:
1.
An unapproved copy of procedure AD 1839.01 was used for field implementation 2.
The actuation of the pressure relief valve RC-2A on September 24, 1977, was not logged as a cycle event as required by AD 1839.01 3.
The cycle and temperature limitations of procedure AD 1839.01 Revision I was not revised to be in agreement with the curren_t_
limits of procedure PT 5164.03, Revision 0, for the pressurizer cycle relief valves and electromagnetic relief valve.
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THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO OHIO 43552 l
FE5 3s IE78
License No. NPF-3 Docket No. 50-346 Serial No. 1-13 Page 2
Response
The corrective action taken, the results achieved, and the corrective action taken to avoid further noncompliance are as follows:
Item 1.
The unapproved copy of AD 1839.01 was replaced 4
immediately at the time of the inspection with an approved copy of AD 1839.01.
The Control Room Trip and Transient Log Book was placed on the distribution for AD 1839.01.
Item 2.
We are presently reviewing our logs and computer printouts to insure the transients between August 8, 1977, and the present have been recorded properly.
This review and the resulting update of the log will be completed on March 6, 1978.
The Operations Support Engineers have been informed of the importance of maintaining the Trip and Tran-sient Log Book.
Item 3.
Administrative Procedure, AD 1839.01, was revised and approved on January 30, 1978, to be in agreement with the limits of procedure PT 5164.03 for the pressurizer cycle relief valves and electromagnetic relief valve.
Personnel were informed of the correlation between the procedure AD 1839.01 and PT 5164.03.
Very truly yours, JSG/WHG/ daw O
s em mee UNITED STATES
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f'e, NUCLEAR REGULATORY COMMISSION
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REGION IH E
3 7ss moostvtLT moAo CLEN ELLYN, ILUNol5 60137 Docket No. 50-346 2 7 1978 Toledo Edison Company ATTN:
Mr. James S. Grant Vice President - Energy Supply Edison Plaza 300 Madison Avenue Toledo, OH 43652 Gentlemen:
Tnis refers to the inspection conducted by Mr. T. N. Tambling of this office on December 13-15, 1977, and January 11-13, 1978, of activities at Davis-Besse Nuclear Power Station, Unit I authorized by NRC Operating Licensee No. NPF-3 and to the discussion of our findings with Mr. T. Murray, Plant Superintendent and other members of your staff at the conclusion of the inspection.
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The enclosed copy of our inspection report identifies areas
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examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel.
During this inspection, certain of your activities appeared
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to be in noncompliance with NRC requirements, as described
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in the enclosed Appendix A.
This notice is sent to you pursuant to the provisions of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. Section 2.201 requires you to submit to this office within twenty daye of your receipt of this notice a written statement or explanation in reply, including for each item of noncompliance: (1) corrective action taken and the results achieved; (2) corrective action to be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter, the enclosures, and your response to this letter will be placed in the NRC's Public Document Room, GO.
-ecceep paPG goo /40793
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Toledo Edison Company,,37gg7g I
except as follows. If the enclosures contain information-that you or your contractors believe to be proprietary, you must apply in writing to this office, within twenty days of your receipt of this letter, to withhold such information
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from public disclosure. The application must include a full oe statement of the reasons for which the information is con-sidered proprietary, and should be prepared so that proprietary information identified in the application is contained in an enclosure to the application.
We will gladly discuss any questions you have concerning this inspection.
Sincarely, 1
Caston Fiore111, Chief i
Reactor Operations and Nuclear Support Branch i
Enclosuret IE Inspection i
Rpt No. 50-346/78-01 I
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1 T. D. Murray, Station Superintendent l
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Appendix A NOTICE OF VIOLATION Toledo Edison Company Docket No. 50-346 Based on the inspections conducted on December 13-15, 1977, and January 11-13, 1978, it appears that certain of your activities were in noncompliance with NRC requirements, as noted below.
The following item is an infraction.
T2chnical Specification 6.8.1 requires that written procedures shall be established, implemented and maintained.
Contrary to the above:
1.
An unapproved copy of procedure AD 1839.01 was used for field implementation.
2.
The actuation of pressurizer relief valve RC-2A on September 24, 1977, was not logged as a cycle event as required by AD 1839.01.
3.
The cycle and temperature limitations of procedure AD 1839.01, Revision I was not revised to be in agree-ment with the current limits of procedure PT 5164.03, Revision 0 for the pressurizer cycle relief valves and electromatic relief valve.
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m U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III Report No. 50-346/78-01 Docket No. 50-346 License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, OH 43652 e
Facility name: Davis-Besse Unit 1 Inspection at: Davis-Besse Site, Oak Harbor, OH Inspection conducted: December 13-15, 1977 and January 11-13, 1978 T. &. t cwAG Inspector:
T. N. Tambling i 2]l]T g
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Approved by:
R. C. Knop, Chie I
7 !78 Reactor Projects Section 1 Inspection Summary Inspection on December 13-15, 1977 and January 11-13, 1978 (Report No.
50-346/78-01)
Areas In.pected:
Review of licensee event reports, review of design, design changes and modifications, witnessing of a transient test of 40%
power and followup on licensee's evaluation on Auxiliary Feedwater Pump
. problems. The inspection involved 56.5 inspector-hours onsite by one NRC inspector.
Results: Of the four areas inspected, no items of noncompliance were found in three areas; one apparent item of noncompliance was found in one area (infraction - failure to properly implement and maintain a procedure - Para. 3).
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m DETAILS 1.
Personnel Contacted
- T.
Murray, Station Superintendent B. Beyer, Maintenance Engineer
- L.
Stalter, Technical Engineer L. Grime, Reliability Engineer
- J. Buck, Operations Quality Assurance Engineer
- W.
Green, Assistant to Station Superintendent
- T. Hart, Quality Assurance Engineer
- W. Schultz, Power Engineering and Construction C. Daft, Quality Control Supervisor The inspector also talked with and interviewed other licensee employees, including members of the technical and engineering staff, operations staff and radiation protection.
- Denotes those attending exit interview.
2.
Review of Nonroutine Events Reported by the Licensee The inspector reviewed licensee actions with respect to the follow-
/'~'N ing listed nonroutine event reports to verify that the events
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were reviewed and evaluated by the licensee as required by Techni-cal Specifications, that corrective action was taken by the licensee, and that safety limits, limiting safety system settings, and limiting conditions for operation were not exceeded.
The inspector examined selected Station Review Board minu6's, the licensee investigation reports, logs, and records, and inspected equipment and interviewed selected personnel.
Inadvertant boren dilution due to improper adjusted control valve (NP-33-77-02)
Loss of Decay Heat flow during surveillance test of SFAS (NP-32-77-05)
Momentary loss of core flow in Mode 4 due to valve switching sequencing problem (NP-32-77-06)
Failure of Decay Heat check valve DH-77 to close when Core Flood Tank was unisolated (NP-32-77-08)
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Both Auxiliary Feedwater Systems inoperable (NP-32-77-11) s Electrolyte level above maximum level indication mark on 125 volt batteries (NP-32-77-13)
Loss of Makeup Pumps due to blown control circuit fuse (NP-32-77-14)
Cooling water to Auxiliary Feedwater Pump bearing improperly aligned (NP-32-77-17)
Auxiliary Feedwater Pump turbine governor valve vibrated closed (NP-32-77-18 and Part 21 report, L. E. Roe to J. G. Keppler, RIII dated November 16, 1977)
Inadvertant SFAS actuation in Mode 5 coupled with loss of 120 VAC V2 bus (NP-33-77-06)
Failure of two control room emergency dampers to meet response time during testing (NP-33-77-09)
Loss of SEAS channel 1 due to loss of Y1 distribution panel (NP-33-77-14)
Containment internal pressure not being monitored (NP-33-77-16)
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N, JI Emergency Ventilation System 1-1 inoperable (NP-33-77-34)
Electrolyte level in station batteries high (NP-33-77-29)
Containment personnel air lock interlock inoperable (NP-33-77-35)
Steam generator level indication outside differential tolerance (NP-33-77-44)
Decay Heat isolation valves DH-11 and DH-12 declared inoperable (NP-33-77-50)
The inspector noted that the licensee had identified and corrected four items related to these events.
No other items of noncompliance or deviations were identified.
During the exit interview the inspector requested supplemental reports for MP-32-77-18, NP-33-77-06 and NP-33-77-14 to cover the final corrective actions, concurred with the licensee's interpretation,
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\\s_s' the post-LOCA flow path through DH11 and 12 (NP-33-77-50), discussed interpretations regarding surveillance testing and noted improvement in the closing out of recent DVR's associated with non-routine reports.
The fo11'owing licensee event reports were reviewed and closed,out on the bases of an inoffice review and evaluation:
Loss of Shield Building integrity (NP-?'-77-32)
Containment isolation valve CV-5074 inoperable (NP-33-77-38)
Main Steamline hydraulic snubber SR17 and SRll inoperable (NP-33-77-70)
Makeup Pump 1-2 removed from service to allow maintenance (NP 77-73)
Decay Heat Valve Pit opened to perform wiring change (NP-33-77-77)
RPS channel 1 accidently de-energized (NP-33 i.-78)
Gove-eave closed on AFP 1-2 (NP-33-77-80)
EVS train 1-2 delta pressure controller malfunction (NP-33-77-94)
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Diesel generator trip on overspeed (NP-33-77-96)
Control Room isolation damper would not completely close (NP-33-77-95)
Failure of AF 3872, Stop valve between AFT l-2 and Steam Generator (NP-33-77-83) 3.
Reduction of Discharge Cycle Limits on Pressurizer Relief Valves, 4
LER NP-32-77-12 The inspector reviewed the licensee's report dated August 8, 1977 to determine whether the corrective action was being implemented.
Within this review on December 14, 1977 the inspector found that a log of operational transients was baing maintained in the control room.
- However, a.
An unapproved copy of AD 1839.01, Documentation of Allowable Operating Transient Cycles, was being used in the field for implementation.
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The actuation of the pressurizer relief valve RC-2A on September 24, 1977, had not been logged.
c.
The current revision (Revision 1) of AD 1839.01 limitation for the number of cycles and temperatures limits on the pressurizer relief valves was not in agreement with PT 5164.03, (Revision 0) Pressurizer Power Relief Valve Periodic Test.
The failure to use an approved copy of AD 1839.01 for field imple-mentation, to log the transient cycle on September 24, 1977, and to revise AD 1839.01 to reflect current limitations is considered to be an item of noncompliance with the requirements of Section 6.8.1 of the Technical Specification.
The licensee immediately replaced the unapproved copy of AD 1839.01 with an approved copy.
In the exit interview the licensee ccmmitted to revise AD 1839.01 to reflect the correct limits, review the transient log to insure that items are being recorded properly, to record the temperature at the time of pressurizer relief valves actuations, and to evaluate how they will assess pressurizer relief valve actuatiens at different temperatures.
4.
Containment Isolation Valve RC 240A Inoperable, LER NP-33-77-40 The inspector reviewed the licensee's report dated August 26, 1977, gss
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to determine the status of the corrective action. Although the initial corrective action had been completed, DVR 74-1 had not been closed out and the investigation by the architect / engineer was not available for review by the inspector.
This item remains unresolved pending the close out of the DVR and the investigation.
5.
Circular 77-13, Reactor Safety Signals Negated During Testing The inspector reviewed with the licensee his conclusion and actions on Circular 77-13.
The licensee had concluded that their proce-dures and management control systems were adequate to prevent a similar occurrence, however, the referenced incident was being used as a basis of a special training secession for operating and plant personnel to demonstrate the possible consequences of not adhering to procedures and controls.
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6.
Design, Design Changes and Modifications The inspector reviewed implementing procedures and selected Facility Change Request forms to determine whether changes to the facility were made in accordance with 10 CFR 50.59, the Technical Specifications and established QA/QC and administrative controls. The review included the following implementing procedures.
Quality Assurance Procedure 2030, Design Control.
Administrative Procedure 1845.00, Changes, Test and Experiments, Rev. O.
Power Engineering Instruction DB1, 320, Design Changes, Tests, and Experiments, Rev. 3.
Power Engineering Instruction DB1-334, Safety Review / Evaluation /
Accident Analysis, Rev. 1.
Power Engineering Instruction DB1-351, Work Package, Rev. 1.
At the exit interview the inspector discussed several areas in AD 1845.00 that should be clarified. These were:
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Section 2.3 - The processing of non-nuclear safety related design
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Section 4.2 - The definition of Nuclear Safety Related should be expanded to include other chapters of the FSAR.
10 CFR Part 50.59 addresses the FSAR and not only Chapter 15.
Also, ASME items are not addressed.
Section 3 - Section 6.5.1.6 of the Technical Specifications states that SRB will review tests, experiments, changes or modifications that affect nuclear safety.
Nuclear safety as used there has a much broader scope than implied in section 4.2.
Section 6.9 and 6.10 - These sections need to be clarified as to who is responsible for preparation of test procedures, the information required in the procedures and the format to be used.
Section 9.5 - This section should be clarified that a delegated representative can act for the Project Engineer to insure that t
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testing and acceptance criteria are met before returning the affected equipment to service.
The following selected Facility Change Requirements were reviewed.
FCR 77-221 AFPT Speed Control Relay Modification FCR 77-199 Modification of Wiring on AFPT Control Relays CW/ AUX and CCW/ AUX FCR 77-154 Pressure Doors 215, 601 and 602 g
FCR 77-417 CRDCS - Trip Breakers FCR 77-026 Proposed Technical Specifications change for DHII and DH12 FCR 77-34 Modification to DH12 Control Circuit FCR 77-47 Proposed Technical Specification Change to EVS FCR 77-072 Change EVS delta P Cell Range from 10 inches to 2 inches H O 2
FCR 77-073
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Proposed Technical Specification Change on Steam
%--l Generator Level Transmitters
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FCR 77-120 Separation of NI-1 and NI-2 Cable Runs FCR 77-152 Modification to EVS Blowout Panel Setpoints The inspector found these facility change requests in various states of completion with the master copies in circulation for review and approval.
In the exit interview the inspector discussed the general status of closing out completed FCR's and stated that the final review of FCR's would be unresolved pending an updating of the licensee's files. This was due in part that the licensee was engaged in an effort to close out outstanding FCR's.
7.
Licensee Internal Audits While reviewing the licensee's implementation of design changes and modification (Facility Change Requests, Paragraph 6), the 2 -
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inspector noted that the licensee had identified an item N' /
associated with the implementation of procedures QAP 2030 and AD 1845.
8.
Auxiliary Feedwater Systems As a result of an inoffice review of problems with the auxiliary feedwater systems, a telephone conversation on December 12, 1977, was held with representatives of the licensee to determine what special action they were taking to insure the reliability of the systems.
On December 13, 1977, the licensee stated they would:
Complete an engineering evaluation of the auxiliary feedwater pump control system to determine if the system is adequate.
Would increase the surveillance testing from monthly to weekly.
Would not proceed above 75% power if any more problems developed until both the licensee and the NRC could evaluate the problem (s).
On January 12, 1978, the licensee completed an engineering evaluation of the auxiliary feedwater system problems including a failure on December 28, 1977.
This evaluation and the associated system modi-fications were reviewed by the inspector.
The results of the evaluation and modifications were discussed in the exit interview.
9.
Unresolved Items
\\,,-l Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance or deviations.
Unresolved items disclosed during inspection is discussed in Paragraphs 4 and 6.
10.
Power Ascension Testing, TP 800.14 The inspector witnessed the performance of TP 800.14, Reactor / Turbine Trip performed December 15, 1977 from 40% power.
Overall performance was evaluated, including adherence to test procedure and meeting of acceptance criteria.
No items of noncompliance or deviations were identified.
11.
Exit Interviev The inspector met with licensee representatives (denoted in Para-graph 1) on December 15, 1977 and January 13, 1978.
The inspector summarized the scope and findings of the inspection.
The licencee m) t l
l representative made the following remarks in response to certain of the items discussed by the inspector.
Acknowledged the statement by the inspector with respect to the items of noncompliance (Paragraph 3).
Acknowledged the inspector's request for supplementary closeout reports to specified licensee event reports (Paragraph 2).
Acknowledged the inspector's statement that to closecut of LER NP-33-77-40 would remain unresolved pending the closeout of the investigation (Paragraph 4).
Reaffirmed their review and action on Circular 77-13 (Paragraph 5).
Stated that they would review AD 1845.00 with regards to the inspector's comments and would continue their efforts for the timely closeout of facility change requests.
They also acknowledged the inspectors statement that closeout of facility change requests would remain unresolved pending updating of the licensee files (Para-graph 6).
Stated that they would continue weekly surveillance testing on the auxiliary feedwater pump. This surveillance frequency would continue until confidence and reliability had been reestablished, but would continue for a minimmn of 8 weeks before a reduction to
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normal frequency (Paragraph 8).
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