ML20150E492
| ML20150E492 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/08/1988 |
| From: | Bird R BOSTON EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 88-106, NUDOCS 8807150169 | |
| Download: ML20150E492 (4) | |
Text
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h eosem asm Pilgiim Nuclear Power station Rocky Hill Road Plymouth, Massachusetts 02360 Ralph G. Bird July 8, 1988 senior vice President - Nuclear BECo Ltr. #88-106 U.S. Nuclear' Regulatory Commission Attn: Document Control Desk Hashi.ngton, D.C. 20555 Docket No. 50-293 License No. OPR-35
Subject:
NRC Inspection Report 50-293/88-12
Dear Sir:
Attached is Boston Edison Company's response to the Notice of Violation contained ir the subject inspection report.
Please do not hesitate to contact me directly if you have any questions.
M
.G.
rd CS/bl
Attachment:
Response to Vio.ation 88-12-02 cc: Mr. Hilliam Russell Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Rd.
King of Prussia, PA 19406 Sr. Resident Inspector - Pilgrim Station I
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7 8807150169 880700 ADOCKOSOg93 PDR 0
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ATTACHMENT RESPONSE TO NOTICE OF VIOLATION (88-12-02)
NRC NOTICE OF VIOLATION "As a result of the inspection conducted on March 6, 1988 to April 17, 1988, and in accordance with the NRC Enforcement Policy (10 CFR 2, Appendix C), the following violation was identified:
10 CFR 50 Appendix B, Criterion III, Design Control, and Boston Edison Company Quality Assurance Manual (BEQAH) Volume II Section 3, Design Control, require that measures be established for the control of design activities to assure that appropriate quality standards be specified and design reviews be performed.
Further, BEQAM Section 3.3.2.8 requires that methods for verifying design changes, such as design reviews and qualification testing are pronerly chosen and followed; the most adverse design conditions specified for test programs are used to verify the adequacy of designs.
Contrary to the above on January 19, 1988, it was determined that the Field Revision Notice 85-07-62 to the Plant Design Change (PDC) 85-07 for installation of new reactor water level gauges had not been properly reviewed and released in that the configuration drawings were incorrect.
This resulted in incorrect installation of the gauges.
The Field Revision Notice (FRN) 85-07-62 was released on December 12, 1986 and the implementation of the FRN 85-07-62 was completed on April 22, 1987.
In addition, the design verification testing for the installed reactor water level gauges, Temporary Procedure 87-66, Pre-Operational Test of the New Barton Indicating Units LI 263-59A and LI 263-598, completed on June 10, 1987, did not meet the requirements of the BEQAM, Section 3.3.2.8 in l
that the testing failed to verify the design adequacy."
i Boston Edison Comoany Resconse Discussion:
As indicated in Inspection Report 87-57, the installation error was identified by the licensee during an investigation into the response of the reactor vessel level instruments during calibration.
The investigation was originally begun to determine the cause of an inadvertent scram signal and is documented in Licensee Event Report 88-02.
The procedure for verifying design adequacy (TP 86-188 Recalibration Test of l
Proportional Amplifiers PA 640-3A & B and Various Reactor Hater Level Transmitters) did confirm that level transmitters affected by PDC 85-07 l
tracked vessel level properly, but failed to confirm that the local indicators (LI 263-59A&B) tracked level properly.
In light of the two issues described l
in the violation (i.e., design adequacy and adequacy of design verificaticn testing) the segments of the violation response are segregated to address both issues.
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Cause:
The cause of the drawing configuration discrepancy is that the vendor did not communicate that the internals of the indicator had been reversed to satisfy the BECo instrument specification and that the drawing supplied did not accurately reflect the orientation and location of the high and low pressure connections.
The drawing showed the high pressure connection on the left side of the indicator when facing the indicator.
The indicator came with the high pressure connection on the right side of the indicator.
The indicator connections were not labelled, thus, there was no way to tell at receipt inspection or installation that the connections were physically reversed.
The design review performed for this modification was evaluated considering this event and found to be adequate and not a contributor to this event. All design documents accurately reflected the configuration shown on the vendor drawing.
After the problem was detected, Boston Edison contacted the vendor, who confirmed in a letter dated January 25, 1988 that the Differential Pressure Unit (Model 199) was mounted on the indicator opposite of what the drawing shows, which should be low pressure connections on left and high pressure connections on right.
Boston Edison investigated this problem using the corrective action process as documented by Potential Condition Adverse to Quality (PCAQ) N00 88-013.
Based on the root cause determination, it is considered to be an isolated event.
The cause of the failure to identify the problem during design verification testing was procedure inadequacy.
The two local level indicators were erroneously excluded from TP 86-188. When reactor vessel level was varied to demonstrate that the new reactor water level transmitters tracked actual level, the local indicators were omitted.
During calibrations the local level indicators a'te isolated from the reactor vessel so indicated level is not verified to track vessel level.
It should be noted that part of PDC 85-07, Revision 1 was a requirement to ensure that the affected instruments properly sensed reactor vessel water level.
Had this verification testing been accomplished for the local indicators, the drawing discrepancy would have been identified.
Corrective steps which have been taken to address the design adequacy issue are:
Written confirmation was received from the vendor that the high and low pressure input to the indicators should be opposite of that shown on the vendor's generic drawing.
Field Revision Notice 85-07-126 was then issued to modify the tubing connections accordingly.
The input connections to the indicators were physically reversed and tested to confirm that the indicators were oriented properly. A channel check was performed and will continue to be periodically i
performed as part of routine station Procedure 2.1.15 "Daily Surveillance Log (Tech. Spec. and Regulatory Agencies)".
ParT 2 of 3
.... s A corrective step which has been taken to address the adequacy of design verification testing issue is:
A sample of test procedure results for modifications installed this outage were forwarded to the Nuclear Engineering design engineers for review.
The purpose of this independent review is to confirm the adequacy of the test procedures to verify the design intent of the modification.
This review will be documented by PCAQ N00 88-012 and is scheduled for completion prior to startup.
Corrective steps which will be taken to avoid further violations:
Since the drawing discrepancy has been determined to be attributable to one supplier, and that we have informed the supplier about the problem, no further corrective steps are necessary.
The corrective steps which will be taken to avoid further violations relative to the adequacy of design verification testing issue are:
Completion of the PCAQ N0D 88-012 Corrective Action prior to start-up, including implementation of any additional requirements identified during the evaluation of the PCAQ.
Date when full compliance will be achieved:
Full compliance was achieved on March 24, 1988 when the level indicators uere properly oriented.
Safety Consecuences:
The 2 local level indicators are not used to actuate any automatic safety systems and therefore the improper installation would not affect the plant transient response to hny design basis accident. There are 6 other safety related level instruments which are used with the Emergency Core Cooling Systems and 4 additional level 1.9struments used for the Reactor Protection and Primary Containment Isolation Systems. These ten (10) instruments indicate in the Cable Spreading Room and four (4) also indicate in the Control Room. Other safety and non-safety related level instruments also indicate vessel level in the Control Room.
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In the case of a need to shutdown the reactor from outside the Control Room, there are level indicators available on the Anticipated Transient Without Scram (ATHS) panels 2277 and 2278 adjacent to the rack containing LI 263-59 A and D.
The additional local indicators would have mitigated l
the consequences of the unavailability of LI 263-59 A and B.
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In addition, the improper installation was identified prior to the I
indicators needing to be operable for the remote shutdown function.
The performance of routine surveillance and start-up testing would have identified this problem prior to completion of the Power Ascension Test
- Program, i
The public health and safety was therefore not affected by this violation.
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