IR 05000244/1979019
| ML17249A776 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/22/1980 |
| From: | Kister H, Markowski R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17249A775 | List: |
| References | |
| 50-244-79-19, NUDOCS 8003190899 | |
| Download: ML17249A776 (22) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-244/79-19 Docket No.
50-244 License No.
DPR-18 Priority Licensee:
Rochester Gas and Electric Corporation 89 East Avenue Category Rochester, New York 14649 Facility Name:
R.
E. Ginna Nuclear Power Plant, Unit
Inspection at:
Ontario, New York Inspection conducted:
December 18-20, 1979
/
Inspectors:
/~. M.
R. S. Markowski, Reactor Inspector date signed date signed Approved by:
>stet, ref, Reactor Prospects Section No. 4, RO&NS Branch date si ned 8'co date signed Ins ection Summar:
Ins ection on December 18-20, 1979 Re ort No. 50-244/79-19 Areas Ins ecte
Routine, unannounce
~nspectson of p ant operations; inoffice an onsste review of Licensee Event Reports; review of interim corrective measures associated with a potential failure mechanism associated with the pressurizer power operated relief valve yoke and review of monthly operating reports.
The inspection involved 22 inspector-hours onsite by one NRC regional based inspector.
Results:
No items of noncompliance were identified.
Region I Form 12 (Rev. April 77)
~gg3190899
DETAILS 1.
Persons Contacted Mr. C. Anderson, guality Assurance Manager Mr. A. Curtis, NDE Coordinator Mr. D. Filkens, Supervisor of Chemistry and Health Physics
- Mr. J.
Noon, Assistant Superintendent Mr. C. Peck, Operations Engineer Mr. T. Schuler, gC Engineer
- Mr. B. Snow, Superintendent Mr. S. Spector, Maintenance Engineer The inspector also interviewed other licensee personnel including members of the operations, maintenance, instrumentation and controls,
, chemistry and.health physics and general office staffs.
- denotes those present at the exit interview.
2.
Plant 0 erations Review The inspector reviewed the following logs and records:
Station Chemistry Log, August - December 20, 1979; Bypass of Safety Function and Jumper Control Request No. 79-22 through 79-34; Station Holding Log; item nos.
79-2437, 79-2438, 79-2669, 79-2670, 79-2674, 79-2683 through 79-2687,.79-2698; 79-2748,. 79-2874 through 79-2889, 79-2898 and 79-2900; and, Reactor Coolant Leakage Surveillance Logs, November 1-6.
The above logs and records were reviewed to verify that:
the required chemical analyses were performed at the frequencies specified for the following:
reactor coolant gross activity, radio-chemical, E determination, Chloride and Flouride, oxygen and boron determinations; secondary coolant gross activity determi-nations; sodium hydroxide determinations; and, boron determinations associated with the boric acid storage tanks, accumulators, spent fuel pit and refueling water storage tanks;
the results of the above determinations were consistent with the
- appropriate Technical Specification Section 3 requirements or appropriate corrective action, and reporting requirements were complied with; the jumper. and hold log entries did not conflict with Technical Specifications; and, leakage monitoring was consistent with station procedures and technical specification requirements.
Acceptance criteria for the above review included inspector judgement, requirements of applicable Technical Specifications, and the following procedures:
A-46,. Bypass of Safety Function and Jumper Control, Revision 4; A-1401, Station Holding Rules, Revision 4; S-12.2, Operator Action in the Event of Indication of Significant Increase in Leakage, Revision 6; and, S-12;4, RCS Leakage Surveillance Records Instructions, Revision 8.
No items of noncompliance were identified.
3.
In Office Review of Licensee Event Re orts LERs The below listed LERs were reviewed in the NRC: I office to verify that details of the event were clearly reported including the accuracy of the description'f cause and adequacy of corrective action.
In addition, a determination was made of whether further information was required from the licensee, whether generic implications were involved and whether the event warranted on site followup.
The following LER's were reviewed:
LER 79-009/01T-O, Bor ic Acid Storage Tank "B" Low Boron Concen-tration; LER 79-010/01T-O, Boric Acid Storage Tanks A and B Low Boron Concentration; LER 79-011/01T-O, Pressurizer Power Operated Relief Valve Internals-Manufacturing Error;
--*
LER 79-18/03L-O,
"B" Emergency Generator Output Breaker to Bus
not Closing During Monthly Surveillance Test;
LER 79-023, Pressurizer Power Operated Relief Valve Nozzle Indica-tions; and,
-.-*
LER 79-24,. Boric Acid Storage Tanks A and B Low Boron Concentra-tion.
No items of noncompliance were identified.
.4. Onsite'Followu
'of'Licensee'Events For those LER's selected for onsite followup (denoted by an * in Paragraph 3}, the inspector verified by discussion and review of references listed below that:
the reporting requirements of the Technical Specification had been met; the corrective action as stated in the report was completed; the cause of the event had been determined; and, continued operation of the facility was conducted in accordance with Technical Specifications.
No items of noncompliance were identified.
The details of the review and any further actions which are required to be completed are detailed in the following subparagraphs.
a.
'"LERs 79-'009,'9-010'and 79-'24 On April 16 and 20, 1979 routine sampling of the Boric Acid Storage Tanks had indicated a boric acid concentration below 12%
(12-135 required by Technical Specifications).
Subsequent to this event, the licensee had determined that a
check valve in the boric acid supply line to the boric acid blender was leaking.
This check valve was replaced.
On December 17, 1979, a similar event occur red which indicated
~ that the check valve replacement had not been effective in preventing recurrence or some other cause existed for the reduction in boron concentration.
During this inspection, the inspector reviewed the below listed drawings and system operating procedures to determine if normal system configuration or operating practices had the potential for causing these dilution The documents reviewed were:
Flow Diagram, 33013-433, Chemical and Volume Control System, Revision 0; Flow Diagram, 33013-426, Chemical and Volume Control System, Sheet 81, Revision 2; W Drawing, 684J809, Boric Acid Tank Details, Section
through 3 of 3, Revision 5; S-3.1B, Pre-Operational Lineup of Boric Acid System, Revision 8;
S-3.1D, Boron Control Valve Lineup, Revision 2; S-ll, Batching Tank, Revision 6; S-ll.l, Boric Acid Storage Tank Transfer to Batch Tank, Revision 2.
As a result of this review, the inspector did not identify any procedural provisions which would have caused, if followed, a dilution of the Boric Acid Storage Tanks.
Further discussion with the licensee indicated that the occurrences could not be correlated with any major plant events i.e., startup and/or dilution of primary boron concentration during plant return to power.
The licensee stated that the most probable cause still appears to be relatable to leakage through the check.
valve and FCY110A which is normally open when de-energized.
The licensee has stated that an Engineering Work Request (EWR)
has been transmitted to corporate engineering recommending modification to the control circuitry to FCV110A to permit it to close subsequent to completion of normal boration/dilution operations.
The followup report associated with LER 79-24, dated December 28, 1979, has been submitted.
Subsequent review of this followup report by Rl, this item will be reinspected (79-19-01).
b.
LER 79-011 The inspector reviewed the following documents:
Purchase Order No.
N-EG-97539 dated June 5, 1979; Requisition No. 42573-29-1 and 42573-70-2 (receipt inspection documentation)',
RGSE Vendor Surveillance Report dated June 22, 1979; Copes Vulcan Inspection Reports dated June 13, 1979, May 25, 1979, June 5, 1979; Maintenance Procedure, M37.11, "Pressurizer Power Operated Relief Valves 430 and 431C Repair and Inspection,"
dated July 18, 1979 (PORV 430)
and July 17, 1979 (PORV 431);
Various correspondences between RG&E Engineering and Copes Vulcan and internal memoranda covering the period December
.1977
-. July 1979; and, Cor rective Action Request No. 1223, initiated May 18, 1979 and completed on July 26, 1979.
The above documents indicated the following:
surveillance performed by the licensee, prior to shipment of the replacement plug and cage assembly, indicated by review of drawing M-132476 Revision 5 that the template had been changed and dimensional checks had been performed; procedure M37.11 specified valve adjustment and testing criteria consistent with the vendor's recommendations; the testing subsequent to replacement of the plug assemblies met the acceptance criteria specified in procedure M37,.4; and, plug assemblies ordered to previous revisions of drawings M-132476 and maintained as spares were removed from the gA stock.
The inspector had no further questions on the corrective action taken for this ite LER 79-18 The inspector reviewed the Official Record and the A52.4 form dated September 13, 1979 and verified that the A Diesel Generator was run continuously during the period the B Diesel Generator Bus 16 breaker was inoperable and the breaker was tested satisfactorily subsequent to repair.
LER 79-23 On December 10, 1979, Region I was informed by the licensee that linear indications (cracks)
near the safe end of the Pressurizer Power Operated Relief Valves'ozzle were detected.
During the course of the week of December 10, 1979, Region I was kept informed of the ongoing evaluation of the cracks detected.
No further cracking was detected on the other nozzles on the pressurizer head.
During this inspection, the inspector reviewed results of the chemical analysis of the insulation material and the certificates of conformance of the replacement insulation.
The cause of cracking has been prelimarily determined-by the licensee to be stress corrosion cracking.
The chemical analysis of the in place insulation cover material, when analyzed in accordance with Regulatory Guide 1.36, indicated a combined CL/FL content which fell within the unacceptable region of Figure 1 of the Regulatory Guide.
The fill material was found to be acceptable.
Based on a review of the certificates of conformance associated with insulating material maintained in stock on site and the amount of work performed in the pressurizer head area in the recent past, the licensee has determined that the most probable source of chemical contaminants did not result from improperly specified insulation material but from contamination of the insulating material cover during maintenance activity.
A followup report associated with this LER was submitted, dated December 21, 1979.
Subsequent to further review of this followup report, by the NRC, this item will be reinspected (79-19-02).
5.
Pressurizer Power 0 crated Relief Valve Yoke Material; Unresolved Item 79-SP-0 a.
References Letter:
V. Stello-to Rochester Gas
& Electric Co., Potential Failure of Copes Vulcan Pressurizer Relief Valve Creating Small Break LOCA, September 18, 1979; Letter:
K. Amish to V. Stello, Information Requested Concerning Pressurizer Power Operated Relief Valve Yoke Material, October 5, 1979; O-l.l, Plant Heatup from Cold Shutdown to Hot Shutdown, Revision 38; 0-1.1D, Pre-heatup Plant Requirement Checklist, Revision 9; 0-1.2,, Plant from Shutdown to Steady Load, Revision 49; 0-2.2, Plant Shutdown from Hot Shutdown to Cold Condition, Revision 38; 0-2.5, Plant Shutdown. from Hot Shutdown to Cold Shutdown When Condensor Steam Dump is Unavailable, Revision 13; 0-6.2, Main Control Board System Status Verification, Revision 8;
0-7, Alignment and Operation of the Reactor Vessel Overpressure Protection System, Revision 5; S-2.2, Pressurizer Pressure and Spray Control, Revision 9; D-12, Pressurizer High Pressure, Revision 0 (at setpoint of 2385 psig);
F-2, Pressurizer High Pressure, Revision 1 (at setpoint of 2310 psig);
F-26, Pressurizer Pressure High Channel Alert; E1.4, Steam Generator Tube Rupture; Revision 6;
E-15.2, Mal-function of Pressurizer Heaters or Spray Valves, Revision 3; IE Report No. 50-244/79-13, Paragraph 6,
November 1, 1979.
Introduction By letter dated September'8, 1979, the NRC requested that a
review be performed by Rochester Gas and Electric Co.
(RG8E) to determine if the yoke material utilized in the Pressurizer Power Operated Relief Valve (PORV) was cast iron.
IE Report No. 50-244/79-13 documented the initial on site review of this matter and the confirmation that operators were made aware of the potential failure mechanism of the subject yoke.
RG&E's response (dated October 5, 1979) stated that grey cast iron yokes were installed; provided justification for interim plant operation with the PORV block valves closed; and, coranitted to replacement of the yokes.
During the cour se of this inspection, the inspector reviewed the above referenced procedures and held discussions with licensee personnel to establish whether:
plant heatup procedures were revised to contain provisions for PORV block valve closure; plant cooldown procedures contain provisions for PORV block valve opening for low temperature pressure protection; emergency procedures that may require the use of PORVs pro-vide sufficient provisions for PORV operation; and, actions taken to date to secure the replacement cast steel yokes were consistent with RG8E's response letter dated October 5, 1979.
The results of this review are as follows:
(1)
Procedures 0-1.1, 0-1.1D, 0-1.2, 0-6.2, 0-7 and S2-2 did not specifically require block valve closure concurrent with plant heatup.
Discussions with licensee personnel indicated that the block valves had been maintained closed utilizing the equipment holding provisions of procedure A-1401.
The records of this action were not available for the inspector's
review prior to the end of this inspection on December 20, 1979.
On December 21, 1979, the Assistant Plant Superinten-dent notified NRC:Region I that the records had been located and appropriate procedures had been formally revised.
Pending review of the subject records and procedure revisions, this item remains unresolved (79-SP-01).
(2)
The inspector performed a review of Plant Operations Review Committee (PORC) minutes to determine if this matter had been formally reviewed and documented by the PORC.
The minutes reviewed covered the period September 18 - October 5, 1979, on a sampling basis.
l<ithin this time frame, documentary evidence of this review was not identified.
In independent discussions with the Plant Superintendent and other PORC members the inspector determined that the matter had been discussed by the PORC.
The Plant Superintendent stated a further review of minutes would be conducted.
Pending completion of this action and review by RI, this item is unresolved (79-19-03).
.
c.
Status of Re lacement Yokes Replacement cast steel yokes have been received.
However, during
- radiographic examination.at the site, cracking had been detected in the lower section of one of the yoke arms on both replacement yokes.
In conjunction with the NDE Coordinator, the inspector qualitatively reviewed the radiographs.
It was noted that the cracks originated in and were located within a uniform section of the lower yoke assembly, approximately equally distant from geometric transition areas and the pattern was a longitudinal crack (with some indication of voids along this axis) from which emanated transverse cracks.
The licensee, in conjunction with the supplier, is presently evaluating the indications and the need for repair or further replacement.
Pending further review of the licensee's disposition by the NRC, this item is unresolved (79-19-04).
6.
Nonthl Re ort Review An inoffice review of monthly reports of operating status information submitted in accordance with Tethnical Specification 6.9.1 was conducted.
- The reports were reviewed to determine if any information in the report
, should be classified as an abnormal occurrence and whether the required information was submitted.
The monthly reports of February through August, 1979 were reviewed.
No items of noncompliance were identified.
7.
Unresol ved'tems Items about which more information is required to determine acceptability are termed unresolved.
Paragraph 5 of this report details three unresolved items.
8. 'xit:Interview A management meeting with licensee representatives (denoted in Paragraph 1)
at the conclusion of the inspection on December 20, 1979.
Subsequent to the completion of the inspection, telephone discussions were conducted on December
and 26, 1979 between the Assistant Plant Superintendent and the Chief, Reactor Projects Section 4, and the Plant Superintendent and this inspector respectively.
During the meeting and the subsequent telephone discussions the scope and findings as detailed in this report were presented.