ML20071B056
| ML20071B056 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/10/1983 |
| From: | Hunter R INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20071B046 | List: |
| References | |
| AEP:NRC:0760, AEP:NRC:760, NUDOCS 8302250311 | |
| Download: ML20071B056 (4) | |
Text
'
i e'.
lNDIANA & MICHIGAN ELECTRIC COMPANY P. O. BO X 18 BOWLING GREEN ST ATION N EW YORK, N. Y.10004 January 10, 1983 AEP:NRC:0760 Donald C. Cook Nuclear Plant Unit Nos. 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 REPORTS No. 50-315/82-15 (DPRP);
50-316/82-15 (DPRP)
Mr. James G. Keppler U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
This letter is in response to Mr. R. L. Spessard's letter of December 10, 1982 which forwarded the subject Inspection Reports of the routine safety inspection conducted by your staff at the Cook Plant on August 1 through September 15, 1982. The Notice of Violation attached to Mr. Spessard's letter identified four violations. Our response for these follows.
ITEM 1
" Technical Specification 3.2.2.3.b states "With more than one charging pump OPERABLE or with a safety injection pump OPERABLE when the temperature of any RCS cold leg is less than or equal to 188 F',
unless the reactor vessel head is removed, remove the additional charging pump and the safety injection pump motor circuit breakers from the electrical power circuit within one hour."
Contrary to the above, on July 16, 1982, the licensee had power to both operable charging pumps on Unit 1 from 0542 to 0659, a period of one hour and seventeen minutes while the RCS cold leg temperatures were less than 188 F, the reactor vessel head was installed, and the safety injection pumps were inoperable. This is a Severity Level IV Violation."
8302250311 830218 PDR ADOCK 05000315 0
Mr. James G. Kappler 2
AEP:NRC 0760 RESPONSE TO ITEM 1 There was a misunderstanding of the newly issued Technical Specification requirements in Amendment 53 to the Unit 1' operating license. 'The operators believed they were doing the surveillance test more conservatively in that only one charging pump was ever capable of injecting water into the Reactor Coolant System.' During the surveillance test the following conditions existed: the East
- Centrifugal Charging pump discharge valve was shut and the pump aligned for recirculation, rendering it incapable of being operable; the RCS was at half loop, the reactor head was on, the PORV's were blocked open, and a steam generator manway cover was removed. These conditions were conservative with: respect to the intent of the Technical Specification in that the available RCS surge volume and the openings in the RCS would have been more than sufficient to prevent overpressurization if both CCP's were functioning.
The misunderstanding came about because this is apparently the only Technical Specification which states specifically how the equipment was to be made inoperable..Either method used will result in a Condition Report, whether both pumps are racked in or out at the same time, as i
there is also a Technical Specification which requires one to be operable.
On August 13, 1982, AEPSC sent the Plant Manager a letter clarifying the method of performing the surveillance testing of the charging pumps which was agreed upon by AEPSC and the NRC. The Operations Department revised the surveillance test procedures for both Units to reflect the language of the technical specification. Two Operating Memo (82-75 and 82-86) were issued explaining the testing procedures and the reasons for the controls.
ITEM 2 "10 CFR 50, Appendix B, Criterion XI, states in part "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..."
The licensee commits in Section 1.7 of their FSAR, Amendment 82, dated November 1978, to ANSI N18.7-1976. Paragraph 5.2.19.3 of ANSI N18.7-1976 states in part " Tests shall be performed following plant modification... to confirm that the modifications... reasonably produce expected results and that the change does not reduce safety of operations."
Contrary to the above, modifications were made to the Unit 1 Control Room Emergency Ventilation System between August 30 and September 10, 1982, which affected the operability of both trains of the ESF Function of that system. Post-modification operability testing was not conducted on either train until after the unit entered an operational mode in which both trains of the system were required to be operable and exceeded that requirements of the Action statements of Technical Specification 3.7.5.1.
This is a Severity Level IV violation."
l
Mr. Jamas G. Keppler, 3
AEP:NRC 0760 RESPONSE T0 ITEM 2 Refer to LER 1-82-077.-01X-1 for response to this iten. LER 1-82-077-01X-1 has been reviewed by this office and is being sent to you under separate cover.
ITEM 3
" Technical Specification 6.8.1 states " Written procedure shall be established, implemented and maintained covering the activities referenced below:.. b. Refueling Operations... Cycle VI-VII Refuelirg Procedure FP,AEP-R6 Section 9.2.3, Manipulator Crane, Precaution 6 states in part; In the event that a large or unexplained. change in load appears, the operator should immediately stop the equipment and evaluate the situation. Step C.13 states " Check that.the Fuel Transfer System container... is in position and ready to receive fuel..."
Contrary to the above, for Unit 1, at about 1400 on August 4, 1982, the Manipulator Crane Operator failed to immediately stop and evaluate the situation when he observed an unexplained load change on the Dillon Load Cell while lowering a fuel assembly into the fuel transfer system container and failed to check that the fuel transfer system container was in position to receive fuel. This is a Severity Level IV Violation".
RESPONSE TO ITEM 3 Several control methods have been implemented to ensure adequate adninistrative controls over the refueling operations. These include:
1.
Plant Manager Standing Order #43, " Duties of the Licensed Senior Reactor Operator assigned to Fuel Handling Duties", was rewritten and issued 10-27-82.
This PMSO outlines the duties and-responsibilities of the SRO-CA from the start of Refueling Contractor activities on site until the reactor head is replaced following refueling.
2.
Four operators have been designated as permanent SRO-CA's.
In order to qualify as an SRO-CA,-the operators undergo special training to ensure they have sufficient knowledge of the refueling activities to properly supervise the operation. These special classes were held in October, 1982, in preparation of the Unit 2 refueling outage. These classes will be conducted prior to each Units refueling outage by the D. C. Cook Training Department until personnel are trained sufficiently.
3.
Meetings will be held between the Westinghouse refueling crew and the Cook personnel involved in the refueliug af ter the crew arrives on site but before the refueling is started to discuss the planned activities and attempt to head off problems before they occur.
(
4.
The Westinghouse refueling procedure has been revised to require i
proper communication discipline at all times during the fuel shuffle.
3
Mr. Jam:a G. Kcppler 4
AEP:NRC:0760
. +
ITEM 4
" Technical Specification 3.9.2 states in part that "As a minimum two source range neutron f' ux monitors shall be operating, each with i
continuous visual indication in the control room and one with audible indication in the containment.. with the requirement of the above specification not satisfied, immediately suspend all operations involving core alterations..." Contrary to the above on August 16, 1982, at 1130 the inspector observed fuel being loaded into the Unit I core without audio count rate indication in containment.
This was confirmed by personnel in containment including the SRO-CA, who immediately suspended core alterations until a distinguishable audit count rate signal was re-established in containnent. This is a Severity Level V Violation".
RESPONSE TO ITEM 4 A source assembly was moved to a core location further from the source range detector N-31.
The audio multiplier for the detector was set at 1000 CPS, which was too high for the new source location. This resulted in too long a time interval between sounds. The audio count rate was lowered to 100 CPS restoring a more frequent audible indication. The multiplier was not set in the most advantageous position af ter the source was moved, but the detector was operable and would increase the frequency of audible sounds with any increase of core activity.
The Refueling Procedure now includes a requirement to maintain the audible count at a distinguishable level.
A complete discussion as to the _cause and corrective action taken is contained in Licensee Event Report No. 315/82-074/03L-0.
This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Very truly yours,.
_ my R. S. Hunter Vice President RSIVsag Attachment cc: John E. Dolan - Columbus M. P. Alexich R. W. Jurgensen W. G. Smith, Jr. - Bridgman R. C. Callen G. Charnoff NRC Resident Inspector at Cook Plant - Bridgman 2