ML20024H192
| ML20024H192 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/17/1991 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 1938, TAC-74247, NUDOCS 9105300065 | |
| Download: ML20024H192 (5) | |
Text
o CENTERIOR ENERGY Donald C. Shelton 300 Maison Avenue Vice Piesident Nuclear Toledo. Oil 43t.5? 0001 Davis Besse (419)249 2300 Docket Number 50-346 License Number NPP-3 Serial Number 1938 May 17, 1991 United States Nuclear Regulatory Commission Document Control Desk Vashington, D. C.
20555
Subject:
Deferral of Several 7RF0 Commitments Gentlemen:
This letter is to confirm the status of several modifications previously committed to be implemented during the upcoming seventh refueling outage (7RFO) for the Davis-Besse Nuclear Power Station Unit 1.
The revised schedule for implementation of these commitments was briefly discussed at the April 5, 1991 Quarterly Management meeting.
The duration of the 7RF0 had been previously established as 70 days.
The critical path for 7RF0 was the replacement of the remaining two essential inverters.
Since the original scheduling of the refueling outage it was decided that the installation of the remaining two inverters was no longer necessary.
Notification of this change was made in Serial 1899, dated March 21, 1991.
As a result of this change, the total time to complete the remaining modifications became critical path for the outage.
It also became apparent that the duration of the eighth refueling outage (8RFO) vas increasing as the scope of the critical path activity, Reactor Coolant Pump refurbishment, became better defined.
As a result of the increased duration of 8RF0, it was decided to review the work scheduled for 7RF0 to determine if it could be deferred to 8RF0, thus decreasing the duration of 7RFO.
Ia order to determine if any modifications could be deferred from 7RF0 a reassessment of all approved modifications was performed.
Each g
mod fication was evaluated relative to its impact on: 1) plant safety,
- 2) plant reliability, 3) resource availability and 4) impact on overall outage duration.
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Docket Number 50-346 1.icense Number NPF-3 i
Serial Number 1938 Page 2 As a result of the reassessment several modifications committed to the NRC to be completed during the outage were deferred. These modifications are in the areas of: B&V Ovners Group Safety &
Perf oniance Improvement Program (SPIP); Human Engineering Deficiencies (ilED) resolution; Makeup System improvements for feed and bleed; and N Bulletin 89-02, stress corrosion crackir g of bolting in certain Anchor Darling sving check valves.
Attachment one provides a brief summaty of the otiginal commitments and revised implementation schedule.
Toledo Edison has reviewed the deferral of these commitments both on an individual and collective basis and has concluded that their defertal vill not adversely impact the continued safe operation of the Davis-Besse Nuclear Power Station.
If you have and questions, please call R. V.
Schrauder, Manager -
Nuclear Licensing at (419) 249-2366.
Very truly yours, "j7 1 j$ h[:h/h JMM i
/
Attachment cci P. H. Byron, NRC Region III, DB-1 Senior Resident Inspector A. B. Davis, Regional Administrator, NRC Region III J. B. Hopkins, NRC/NRR DB-1 Senior Project Manager Utility Radiological Safety Board
Docket Number 50-346 License Number NPF-3 Serial Number 1938 Attachment Page 1 B&V Owner's Group Safety and Performance Improvement Program Items By letter dated January 24, 1986, the NRC Executive Director for operations informed the Chairman of the B&WOG that a number of recent events at B&V designed reactors should be re-examired.
B&V0G responded by committing to lead an effort to define concerns relative to reducing the frequency of reactor trips and the complexity of post trip response in B&V plants. This program was entitled Safety and Performance Improvement Program. The SPIP program developed 222 technical recommendations (TRs) that vere to be evaluated by each utility for applicability to their particular plant. Of the 184 recommendations applicable to Davis-Besse, 170 have been closed and 14 remain open.
The remaining 14 items vill be completed during the upcoming outage with the exception of the following items.
TR-159-0PS remote manual control of all post trip steam flow paths TR-178-ICS safe state on loss of power to ICS/NNI TR-ll4-PES installation of synchronized check relays on diesel generators TR-159-0PS includes installation of motor operated valves to provide control room isolation of various steam paths. This is an operational enhancement to allow selective isolation.
Currently this is accomplished by closing the main steam isolation valves.
TR-178-ICS provides for an automatic trip of the reactor on loss of power to Integrated Control System /Non-Nuclear Instrumentation (ICS/NNI) and activation of the Steam Feed Rupture Control System (SFRCS) along with modifying power to the atmospheric vent valves.
This function is now accomplished by manual operator actions directed by plant emergency procedures. The atmospheric vent valve power modification is also considered an operational enhancement to allov operation of these valves from the control room versus local manual handvheel in the event of a loss of ICS power.
TR-ll4-PES provides protection to assure the diesel generators cannot be sychronized to the grid out of phase. The modification to implement this recommendation vill install sychronized check relays for the output breakers to ensure the breakers cannot be closed with the generators out of phase from the grid.
Emergency Diesel Generator (EDG) 1-1 vill be completed this outage while EDG l-2 vill be completed by the end of 8RFO. The new diesel generator schedul,d to be installed this outage, to satisfy 10 CFR 50.63 vill be installe/. vith the synchronized check relays in place.
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Docket Number 50-346 License Number NPP-3 Serial Number 1938 Attachment Page 2 Toledo Edison had previously notified the NRC that all remaining SPIP ltems vould be completed prior to restart from the 7RFO.
The recommendations remaining after 7RF0 vill be completed by the end of 8RFO.
Human Eng_ineering Discrepancy (HED) Resolution in August 1990, (Serial 1820), Toledo Edison submitted an addendum to the 1988 Detailed Control Room Design Reviev (DCRDR) Summary Report.
This addended report contained a section providing a schedule for the completion of outstanding HEDs.
Toledo Edison committed to complete the remaining items by the end of 7RFO.
Toledo Edison is deferring two of the remaining HEDs due to emergent work with greater safety significance for the facility.
The two items being deferred are HED numbers 4.1.020 and 9.8.044.
HED 4.1.020 is to install dual light indication on motor operated throttle valves. Of the sixteen valves requiring modification, five non-safety related valves remain to be modified.
One valve vill be i
vorked during 7RF0, while the remaining four valves vill be worked during cycle eight.
HED 9.8.044 is to modify indication lights on motor operated valves to represent actual travel limits.
Of the approximately 150 valves requiring modification, fifteen non-safety related valves remain to be worked. The work on these valves is currently. scheduled during 8RF0; however, should any of these valves require corrective maintenance this outage the indication lights vill also be modified.
Reactor Coolant System (RCS) Makeup System Upgrades for Feed and Bleed Toledo Edison made a long term commitment in a November 4, 1985, submittal (Serial 1207) to enhance the existing primary system feed and bleed capability at Davis-Besse. The original commitment, based upon a
. preliminary evaluation, involved installation of Reactor Coolant System (RCS) blowdown valves.
In subsequent submittals (Serial numbers 1382, 1526, and 1656 dared June 25, 1987, August 8, 1988 and May 5, 1989, respectively), the original blowdown valve approach was modified because detailed analysis could not fully support viability of the method.
As a result, in order to achieve an effective method of feed and bleed cooling capability, RCS makeup system flov enhancements and upgrade modifications were made during the fifth and sixth refueling outage.
Docket Number 50-346 ticense Number NPF-3 Serial Number 1938 Attachment i
Page 3 In letter Serial 1836, dated September 18, 1990, Toledo Edison informed the NRC of the status of the Feed and Bleed modifications after the si> ' h ref ueling outage. This letter stated that TE satisfied its commitment to enhance the feed and bleed capability with the exception of qualification / replacement of the AC powered makeup pump bearing oil pump motors. Toledo Edison committed to install new qualified motors during 7RF0.
As previously stated, when compared to the other modifications scheduled for this outage, this modification does not contribute to plant safety and reliability to the degtee that the other modifications do, therefore it was determined that this modification be def err ed until 8RFO.
NRC Bulletin No. 89-02, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350V Sving Check Valves or Valves of similar Design Toledo Edison responded to NRC Bulletin 89-02 en June 14, 1990 (Serial 1807) by stating that no Anchor Darling sving check valves are installed at Davis-Besse in safety related applications.
HovcVer, the review identified tvelve Velan valves of similar design. Toledo Edison teported that mod'.fications have been completed on eleven of the twelve identified valves.
Visual examinatian of eleven Velan valves during the sixth refueling outage (6RFO) irentified no cracking of the retaining bolts.
The remaining valve, Component Cooling Vater Valve CC-91, was not inspected during the 6RF0 as stated in Serial 1807 and was to be inspected during 7RFO.
Inspection of CC-91 requires CCV Train 2 drainage. This vould create a coriditic-in which CCV Train 1 would be available with Train 2 drained and fuel
.n the core. This is not a preferred plant condition as a fault in Train 1 would result in loss of all CCV cooling.
Inspection of CC-91 is not considered critical since it was disassembled and inspected during SRFO.
No evidence of bolting degradation was identified at that time.
Further, four similar valves of the same type as CC-91 were inspected and no evidence of bolting fallute was found.
Also, as part of the In-Service-Test (IST) Program, CC-91 vill be reverse flow tested during the 7RFO.
Based on the above, Cc-91 inspection is being rescheduled for BRF0 when the preferred plant condition, full reactor coce off-load, is scheduled.
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