ML20039D091
| ML20039D091 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/21/1981 |
| From: | Baynard P, Poole D FLORIDA POWER CORP. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20039D085 | List: |
| References | |
| 3F-1081-22, CS-81-259, NUDOCS 8112310306 | |
| Download: ML20039D091 (6) | |
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- 3F-1081-22 Mr. J. P. O'Reilly, Director Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission 101 Marietta St., Suite 3100 Atlanta, GA 30303
Subject:
Docket No. 50-302 License No. DPR-72 Ref.:
RII:TFS 50-302/81-15
Dear Mr. O'Reilly:
We cffer the following response to the violations listed in the referenced inspection report.
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A.
!0 CFR Part 50, Appendix B, Criterion V, requires activities
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th be accomplished in accordance with appropriate instructio s,
-., procedures, or drawings.
The licensee's Quality Program, as delineated in FSAR Section
, 1 7.6.7.1.E contains requirements and procedures to assure that
_the 18 criteria specified in 10 CFR 50, Appendix B are accomp-lished in accordance with procedures.
In addition this section requires strict adherence to these procedures.
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The following represent failures to implement appropriate in-structionc and procedures as required by Criterion V.
The instructions contained in Modification Approval Record 1.
(MAR) 78-1-1 implemented in September, 1978 and in MAR 78-1-1A n-implemented in March 1979, required relocation of the emer-
' 'gency feedwater pump turbine bearing oil sightglasses. The 3"
modifications required changes to the sightglass piping and adjustment of the sightglass level per arcached sketches so th'at normal level was indicated in the center of the sight-n gla2 pes.
In addition, MAR 78-1-1A required the scribing and leGling ef low oil level lines on the turbine casing and
,r ren<1 val.of all high/ low level marks from the sightglasses.
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J' Notice'of Violation Response I
.Rll;TFS ' 50-302/81-15 Page 2
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Contrary to.the above, as of August 13, 1981, the sightglass s
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levels'were not adjusted as specified in MARS 78-1-1 and E
78-1-1A.
Also, contrary to the above,.the turbine casing was not scribed and labeled with' lines indicating. low oil level, nor were the high/ low level marks removed from the sightglasses.
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A 1. Response:. Modification. Approval Record (MAR)~78-1-l' and 78-1-1A concerned relocation of emergency feedwater pump and~ turbine Pv bearing oil sightglasses. Florida Power Corporation. admits fail-ure to implement appropriate instructions in this case.
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The orig'inal MAR 78-1-1 was issued and the work completed under A3 ;
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Work Request #00441. During subsequent routine bearing' maintenance,
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th'e sightglasses were inadvertently returned'to their original con-3 q; figuration. MAR 78-1-1A was issued to restore the sightglass modi-fication, and work was completed under Work Request #07407. The completed construction work package for MAR 78-1-1A was then sent-to Nuclear Engineering for approval and closecut. During the course of their review, Nuclear Engineering added instructions which were not contained in original MAR 78-1-1 nor 78-1-1A, and returned the " updated" 78-1-1A to the plant site under cover of a transmittal letter marked " approved". The added instructions were not identified on the cover letter, and were, therefore, in-
, advertently overlooked when the package was received on site.
The correction of this deficiency has been implemented under Work-Request #26307. In addition, on August 25, 1980, Nuclear Engineer-ing issued Safety-Related Engineering Procedure (SREP) - 17,
" Preparation, Review, and Approval of Safety-Related Field Change Notices." This procedure will assist in ensuring that changes made to any design package for plant modification is contained and reviewed under separate cover from the original design package, to underscore the need for' approvals, quality control, and recog-nition of the chenged areas, to ensure adequate instructions are issued to the field. The-Field Change Notice also provides a ve-hicle for the field.to identify changes needed in the design due to physical interferences. In the case of the problems identified f.
with MARS, 78-1-1 and 78-1-1A, the FCN method would have reduced the potential to overlook the requirements to scribe oil levels on the turbine bearing casings added by Nuclear Engineering.
Further corrective action will be taken in the form of a Nuclear s
Plant Manager's Management Memorandum to all plant personnel to notify them of this violation, and as a reminder that safety-re-lated system configuration / equipment is not to be modified nor s
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' changed without an engineering approved Modification Approval Record (MAR) and associated construction work package.
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r Notice ~of. Violation Response-RII:TFS 50-302/81-15 Page 3 2.
The instructions contained in Modification Approval Record (MAR) 80-5-78C, implemented in August, 1980, required functional test-ing of the main' steam line radiation monitors.
I Contrary to the above, as of August 24, 1981, the test procedure was not completed as specified in the MAR.
A.2.-Response:.Although a review of the records verifies that testing vas
.not completed for more than twelve months following installation, we deny that this' example supports citation under 10CFR 50, Appendix B, 1
in that this modification was not classified safety related.
i Although this MAR was not safety related, discussion of this failure to ' properly implement procedure will be included in the aforementioned Management Memorandum.
3.
Procedure CP-114, Procedure for Preparation of Permanent'and Tem-porary Modifications, specifies in paragraph 7.1 that minor changes to modifications are accomplished by a Field Change Notice (FCN).
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Contrary to the above, as of August 24, 1981, a minor change was made to MAR 80-5-78C in that the data sheet to the test procedure j
was changed and no FCN was issued.
A.3. Response: This change was made without issuing a Field Change Notice.
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Although we have verified that a change was made to the " expected" value column of the. Test Procedure without an FCN, we deny that this j
example supports citation under 10CFR 50, Appendix B,-in that this modification was not classified safety-related.
4.
The instructions contained in Modification Approval Record (MAR) 80-02-72B required pipe support EFH-64 to be modified in accord-ance.with Gilbert Associates Inc., drawing 32064. This drawing required the installation of a 12" square base plate perpendicu-lar with the pipe run and a 12" x 19" base plate in-line with the pipe. Drawing 30264 required bolt spacing of 8" for base plate "D" and the installation of grout under the base plates.
i Contrary to the above, as.of February 27, 1981, installation records for pipe support EFH-64 showed that base plates were re-i versed. The records also showed that minimum bolt spacing was 7 3/8" for base plate "D" and that the base plates'were grouted without a documented procedure'for grouting. In addition, no record of an engineering evaluation and acceptance of these con-ditions was available.
A.4. Response: MAR 80-02-72B required piping support EFH-64 to be modi-fied in accordance with Gilbert Associates, Inc. drawing 32064. The "as built" configuration noted by the NRC was found to have the two i
4 Notice of Violation Response RII:TFS 50-302/81-15 Page 4 base plates reversed, bolt spacing reduced,and plates grouted without a documented grouting procedure.
Florida Power Corpora-tion admits failure to implement appropriate instructions in this case. This incident was a result of personnel error and insuf-ficient quality control of field work.
An immediate seismic evaluation was performed on EFH-64 which proved it to be acceptable.
The FCN will allow the field to make immediate changes to the drawings for "as built" conditions, to notify engineering so as to enable them to assess the impact of such changes, and to initiate seismic or other analyses necessary.
Discussion of the finding on MAR 80-02-72B will also be included in the aforementioned Management Memorandum. The above actions will be completed by 1 November 1981, and corrective action will be achieved.
B.
10 CFR 50.59 states that the holder of a license authorizing opera-tion of a production or utilization facility may make changes in the facility as described in the Safety Analysis Report without prior commission approval unless the proposed change involves ac unreviewed safety question. FSAR Figure 4-5 shows seven auxiliary feedwater inlet lines to each steam generator. FSAR paragraph 10.2.1.2 de-scribes the emergency feedwater system including emergency feed-water head requirements. FSAR paragraph 10.2.1.2'also indicates that emergency feedwater piping is designed to seismic category 1.
Contrary to the above, as of February 1981, Modification Approval Records (MARS) 5-62 and 79-5-62A, authorized modification of the Emergency Feedwater (Auxiliary Feedwater) system and plugging of one inlet line per steam generator, without commission approval, by performing a safety evaluation that consisted of verbal Babcock and Wilcox advice that sufficient emergency feedwater would be de-livered after the modification. No analysis was performed by the licensee.
In addition, the effect of the modification on the seis-mic analysis had not been 3etermined. The modification was completed on July 15, 1980.
B.
Response: Florida Power Corporation denies the alleged violation of Technical Specifications with the following clarifications:
-The NRC cited CR-3 for failure to perform safety evaluations for plant modifications as required by 10CFR 50.59, Paragraph 9.b.,
and additionally, 10CFR 50.59 states that the holder of a license authorizing operation of a production or utilization facility may make changes in the facility as described in the Safety Analysis Report, without prior commission approval, unless the proposed change involves an unreviewed safety question.
E Notice of Violation Resposne RII:TFS 50-302/81-15 Page 5
-After. reviewing the circumstances of the above modification, we deny the alleged violation on the basis that a safety evaluation
. was performed by Florida Fower Corporation which found that an un-reviewed safety question did not exist. Recognizing that this evalua-tion may not have addressed the hydraulic and seismic concerns in a manner.that is consistent with the quality and safety significance of the system affected, subsequent evaluations were made which verified that an unreviewed safety question did-not exist.
We further take exception to'the allegation that verbal advice from B&W is an' unacceptable basisLfor a safety evaluation concerning a modification in which B&W acted as our agent for both design and design verification. We do, however, acknowledge that'such informa-tion should have been formslly documented as telephone. notes.
As a result of discussions with the inspector at the time this find-ing was initiated, Florida Power Corporation immediately requested Gilbert Associates. Inc., to review the seismic effect of the sub-ject modification. Line stress and piping support loads were veri-fied to be acceptable.
Additionally, B&W was requested to verify that the minimum emergency feedwater flow capability to remove decay heat was not impaired.
B&W responded with written verification that emergency feedwater flow was not diminished below an acceptable level.
Notwithstanding our denial that an actual Technical Specification violation occurred, we have initiated additional corrective actions to Administrative Procedures governing plant modifications in recog-nition of our concern with the-potential safety effects of this event.
Specifically, our Nuclear Engineering Department's Procedures, SREP-1,
" Safety Identification and Design Input Requirements" and SREP-4,
" Design Verification" were revised and issued to place' increased em-phasis on design criteria impacted by safety-related modifications.
Additionally, it was discovered that the bases for CR-3 Standard Technical Specificacion.3.7.1.2 requires updating to eliminate a po-tential' source of confusion. As currently written, the second Para-graph of the bases reads:
"The electric driven emergency feedwater pump is capable of delivering a total feedwater flow of 740 gpm at a pressure of'1144 psig to the entrance of the steam generators.
Each steam driven _ emergency feedwater pump is capable of deliver-ing a total feedwater flow of 740 gpm at a pressure of 1144 psig to the entrance of the steam generators. This capacity is sufficient to' ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 280*F where the Decay Heat Re-moval System may be.placed into operation."
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