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I TENNESSEE VALLEY AUTHORITY CHATTANOOGA, TENNESSEE 37401 40 ANNIVERSARY October 23, 1973 RfArMaks's'5 Mr. Norman C. Moseley, Director Directorate of Regulatory Operations U.S. Atomic Energy Commission Region II - Suite 818 230 Pea::htree Street, NW.
Atlanta, Georgia 30303
Dear Mr. Moseley:
This is in response to Directorate of Regulatory Operations letter RO:II:WSL 50-259/73-13, dated October 1, 1973, which stated that certain activities under our license appeared to be in violation of AEC requirements.
Specific citations and our responses follow:
1.A.1.a.--Contrary to paragraph 2.2.B of the Technical Specifications, TVA had not verified the correctness of the relief valve limiting safety system setpoints before the reactor was r, tarted up.
The violation occurred because the relief valves, after their setpoints were verified, were returned to the vendor for modi-fication and testing just before startup. Plant procedures now require documented verification that necessary retesting has been completed and that setpoints are correct.
- 1. A.l.b.-Contrary to Criterion XV of Appendix B to 10 CFR 50, a nonconforming head gasket was installed on the drywell by TVA.
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Silicone "O" rings were' riginally specified and purchased for plant usage. After the decision was made to use drywell head gaskets of a Nordel material, the silicone "0" rings continued to be used whenever preoperational activities required the dry-well head to be in place. For this reason, the silicone "O" rings had not been identified as nonconforming material although they were given segregated storage from the Nordel "O" rings.
Although the plant was aware that the Nordel "O" rings were superior, it was not recognized that the silicone "0" rings were completely forbidden. The Browns Ferry silicone "0" rings were l
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, Mr. Norman C. Moseley, Director October 23, 1973 identified to be retained for emergency usage. Although the silicone "0" ring was inadvertently issued to the plant crafts-men and installed, the mistake was found before any operation and reviewed by both the Plant Operations Review Committee and the Safety Review Board who agreed that limited operation was permissible using the silicone "O" rings. A review has been made of all drywell gaskets to ensure that only those of Nordel material are installed; and, to prevent future violations, all silicone gaskets purchased for drywell usage have been destroyed.
i 1.A.1.c.--Contrary to paragraph 6.3.A. of the Tech-nical Specifications and TVA Standard Practice BFA-25, the procedure for the operating hydrotest (BFGOI 100-7) 1 did not require verification of the removal of jumpers used in the test.
- 1. A.2.c.--Contrary to paragraph 6.3. A. of the Tech-nical Specifications, the temperature of the reactor vessel was not monitored as required by the procedure (BFGOI 100-7) during the operating hydro test.
This instruction has been revised to include verification of e;ch procedural step and all prerequisities such as jumper instal-lation and removal, manually recording vessel temperatures, and recording inspection results.
1.A.1.d.--Contrary to paragraphs 3.5.F. and 4.5.F.
of the Technical Specifications, the reactor pressure was not immediately reduced to less than 122 psig or was the NPCIS immediately demonstrated to be operable upon failure of the reactor core isolation cooling system (RCICS).
1.A.2.g.--Contrary to paragraph 6.7.2.A. of the Tech-nical Specifications, the Director of Regulatory Operations was not notified that the required flow was not met in tests of the RCICS and that reactor pressure was not immediately reduced below 122 peig and that the EPCIS was not immediately demonstrated to be operable.
At the time of the RCICS testing during the startup program on September 8,1973, rod movement and power level changes were being minimized to allow water cleanup in the reactor vessel. A conduc-tivity problem was being experienced that seemed to worsen with
, Mr. Norman C. Moseley, Director October 23, 1973 increased steam flow and improve with reactor water cleanup at constant power levels. At 10:40 p.m. and at 11:00 p.m. on September 8, and again at 3:45 a.m. on September 9, with reactor steam pressure at 150 psig, the RCICS delivered 560-gpm flow through the test line against a 260-psig discharge head. This was 55 psi higher than the discharge head against which the pump would be required to operate if needed to deliver flow to the reactor vessel. The cause of the discrepancy was recognized to be an undersized flow limiting orifice in the full flow test line; and, by means of pump curves, it was determined that the system was more than capable of delivering the required flow under actual operating conditions.
It was the intention at this time to obtain sufficient data to permit design evaluation for resizing of the test line orifice. Between these test runs, the RCICS instrumentation was recalibrated to eliminate any uncertainty about test results because of instrument inaccuracy. The results of this testing program were under constant review by the Startup Coordinating Committee whose membership includes members of the Plant Operations Review Committee. On Sunday morning, September 9, the Startup Coordination Committee made the decision to reduce reactor pressure to less than 122 psig. Pressure was raised temporarily to 150 psig for one additional RCICS test, after which the test line orifice was removed to permit using a throttling valve to vary discharge pressure so that the characteristics of the RCICS under varying discharge pressure conditions could be ascertained. The RCICS was then tested successfully using the test throttle valve after again raising reactor pressure to 150 psi.
The orifice has since been resized, reinstalled, and ratested.
It is our position that the RCICS was never inoperable, since the sys-tem was at all times capable of performing its design function, l.A.2.a.--Contrary to Criterion VI of Appendix B to 10 CFR 50, the control room copy of the Technical Specifications was not up to date.
The plant was advised by telephone on August 31, 1973, that the requested technical specification revision had been verbally approved by the Directorate of Reactor Licensing.
This information was disseminated to plant employees with a r.eed to know, but actual revisions were not entered into the plant copies of the technical specifications since we had expected to receive formal notification from AEC regarding the changes.
In the future, the plant superin-tendent will be notified promptly by telephone of approved changes; and he will cause the changes to be entered in the official plant I
. Mr. Norman C. Moseley, Director October 23, 1973 copies. More formally documented changes will be made before the end of the third working day after the date of AEC approval.
l.A.2.b.--Contrary to paragraph 6.2.A.7. of the Technical Specifications, the Safety Review Board (SRB) minutes did not indicate that the startup program had been adequately reviewed, as required by the SRB Charter.
The Safety Review Board is scheduled to meet at Browns Ferry on October 24 and 25, 1973, to review both the unit 1 startup test program and the units 2 and 3 preoperational test program.
Results of this review will be documented in the minutes of that meeting.
Although the SRB has not formally reviewed the startup test pro-gram since December 1970, some members of the SRB, through their normal line functions, have been intimately aware of the progress of the program.
all SRB members are regularly informed of all reportable abnormal occurrences and irregularities. The SRB is developing and implementing a surveillance plan identifying its major areas of responsibility.
The purpose of this plan is to aid in preplanning and scheduling the SRB's review activities to coincide more closely with TVA's overall program activities.
The Safety Review Board expects to complete its review and make any appropriate recommendations by January 1, 1974.
l.A.2.d.--Contrary to paragraph 6.3.A. of the Tech-nical Specifications, the Master Hot Functional Test Instruction (MHFTI) did not have the proper approvals to perform the test prior to beginning.
The Browns Ferry Operational Quality Assurance Manual, the appli-cable plant Standard Practice, and the Master Hot Functional Test Instruction specify that the hot functional test program begins with the beginning of nuclear heatup and extends through the com-plation of testing at acout the 10-percent reactor power level.
Included in the Master Hot Functional Test Instruction are a number of items intended to be signed off as completed (prerequisites) before the commencement of nuclear heatup. We believe the AEC inspector mistakenly took the itemized preraquisites to be an integral part of the hot functional test program rather than sign-off items.
In reviewing the master hot functional test program as a result of this alleged violation, we can see that these items were not clearly designated as prerequisites to the hot functional test program l
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4 Mr. Norman C. Moseley, Director October 23, 1973 although we would have been in more serious violation had the hot functional test program been permitted to start before these items were completed. The Master Hot Functional Test Instruction was signed authorizing test performance before the initial criti-cality following reactor head installation (beginning of nuclear heatup). To eliminate any possibility of future misinterpretation on the part of the regulatory inspector, the Master Hot Functional Test Instruction will be revised before its usage on Browns Ferry unit 2 to indicate clearly that these items are prerequisites which i
must be completed before the initial approach to criticality after the reactor vessel head installation.
l.A.2.e.--Contrary to paragraph 6.6.A. of the Tech-nical Specifications, several items subsequently reported as abnormal occurrences were not entered j
in the shift engineer's journal.
Paragraph 6.6.A of the technical specifications requires that records and/or logs shall be kept of all abnormal occurrences in a manner convenient for review.
It does not require that these items be kept entered in the chift engineer's journal. Records have been kept of all abnormal occurrences as required by the technical specifications and we now keep a copy of all abnormal occurrence notifications in the shift engineer's office.
1.A.2.f.--Contrary to paragraph 6.2.B.4. of the Tech-nical Specifications, the Plant Operations Review Committee (PORC) did not investigate an unexpected 35% increase in the "A" recirculation pump speed.
Each normal workday morning, the plant Startup Coordination Committee meets to discuss outstanding problems and work in pro-gress and to plan, in. detail,_ activities for the next several days. Members of the Plant Operations Review Committee partici-i pate in the Startup Coordination Committee meeting along with other plant supervisors and vendors' representatives.
The speed drift of "A" recirculation pump which occurred on September 4, 1973, was not formally investigated by the Plant Operations Review Committee because the cause of the drift was understood and corrective action was in progress.
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Mr. Norman C. Moseley, Director October 23, 1973 The activity list of the Startup Coordination Committee meeting indicates that problems with recirculation pump speed control were a discussion item on August 27, 28, 29, 31, and September 4, 5, 6, and 7.
During this period, maintenance and adjustment work were in progress and PORC members vare thoroughly aware of the recirculation pump speed control problems. To document the fact that members of the PORC participate in the daily discussion and planning sessions, brief minutes of the Startup Coordination Committee meetings are now being kept showing significant items discussed and PORC members present.
[:jh general, we have, for each alleged violation, studied the c circumstances surrounding the occurrence; and where it would
'estrengthen our quality assurance program, we have implemented
,, programmatic changes. We have also reviewed Regulatory Operations
. ' Inspection Report No. 50-259/73-13 for proprietary material and have found none.
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Very truly yours, e.s
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/ J. E. Gilleland Assistant to the Manager of Power l
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Mr. R. B. Beers TVA Project General Electric Company Atomic Power Equipment Department San Jose, California 95125 1