ML19346A407
| ML19346A407 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/11/1981 |
| From: | Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19346A408 | List: |
| References | |
| NUDOCS 8106190283 | |
| Download: ML19346A407 (6) | |
Text
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S~ U" SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830 Sacramento, California 95813; (916) 452-3211 P,JR 81-391 June 11, 1981 p,
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DIRECTOR OFFICE OF INSPECTION AND ENFORCEMErlT
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U S NUCLEAR REGULATORY COMMISSION
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WASHINGTON DC 20555 OPERATINC PLANT STATUS REPORT DOCKET NO.
50-312 Enc!osed is the fla y Operating Plant Status 0.eport for Rancho Seco Unit One, f\\
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R.
J. Rodrigue Manager Nucle' Operations Enclosure cc:
If,E Washington (9)
Region V MIPC (2)
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SUMMARY
OF PLANT OPERATIONS 5-1 to 5-3 Continued plant heatup.
5-3 Began deboration to criticality.
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-0715 Reactor critical 9 10 amps on intermediate range.
5-4
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0140 Tripped reactor.
0240 Began RCS cooldown due to leaking sample line valve.
5-8 0230 Began RCS heatup.
1456 RCS @ 2255 psig for inservice hydro test.
2050 Reactor critical, groups 1-79100% WD, group 8 9 34%.
2111 Reactor @ 10" amps.
5-9 0909 Reactor trip due to low FW header pressure.
0940 Reactor critical.
1405 Began rolling turbine and raising reactor power to 15t.
5-10 0048 Closed OCB's.
Began raising power to approximately 25%.
0822 Raising power to 40%.
1034 Transferred site loads.
1140 Reactor @ 40%.
5-12 0815 Raising power to'67% 9 3% per hour 1545 Reactor @ 60%, holding.
5-14 1615 Raising power to 70% @ 3% per hour.
1915 Reactor @ 67%--began 5-hour hold.
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SUMMARY
OF PLANT OPERATIONS (Continued) 5-15 0031 Raising power to 75% ? 3% per hour.
0258 Reactor at 75%--holding.
2315.
Raising power to 92% ? 3% per hour.
5-1A 0220 Began 12-hour hold ? 37% power.
1700 Began increas i ng power to 100% 9 3% per hour.
2235 High radiation alarm on B" main steam line (tube 10ak).
2305 Decreasing power ? IMW per minute.
2331 increasing shutdown rate to 10MW per minute.
5-17 0053 Opened OCB's and tripped turbine.
0055 Tripped reactor.
5-17 to 5-31 Reactor in cold shutdown.
PERSONNEL CHANGES REQUIRING REPORTING No personnel chances that requ i re reporting in accordance with Technical Specifications Figure 6.9.2 were made in May, 1981.
SUMMARY
OF CHANGES IN ACCORDANCE WITH 10 CFR 50.59(b) 1.
Rewired the diesel generator auxiliary alarm contacts so that additional conditions will cause illumination of an annunciator window in the control room.
This modification is the result of a District commitment in response to a March, 1977 NRC letter.
This change will not increase the probabili y t
of occurrence, or the consequences of an accident or mal function of equip-ment important to safety previously evaluated in the FSAR, nor will it create the possibility for an accident or mal function of a type other than those previously evaluated in the FSAR or reduce the margin of safety as defined in the basis for any Technical Specification.
2.
A safety analysis was conducted as a result of an acid brush being dropped intu the pressurizer.
It has been determined that the brush will not be dislodged from its location in the pressurizer and, therefore, will not cause any damage to the D.C S.
SUMMARY
OF CHANGES IN ACCORDANCE 91TH 10 CFR 50.59(b) (Continued) y
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3 A safety analysis was performed to justify the operation-of Rancho Seco Cycle V at the licensed ' rated power of 2772 MW.
All accidents analyzed t
in the Rancho Seco FSAR have been reviewed for Cycle V cperation and, in cases where Cycle V characteristics were conservative compared to Cycle IV, no new analyses were performed.
The Technical Specifications have been reviewed and modi fied where required for Cycle V operation.
It is concluded that Rancho Seco Cycle V can be operated safety at its licensed core power level of 2772 MW.
g 4.
A safety analysis was conducted to establish the acceptability of operating Cycle V with three fuel assemblies which have damaged spacer grids.
Since damaged spacer grids are confined to single cells, at any level, the affected pin still has support within reasonable distances.
The internal structure of the grid assembly remains fully capable of supporting design lateral loads and geometry.
The condition of the fuel assemblies with torn spacer grids has been evaluated and found to be acceptable for continued in-reactor use.
They will not affect the safety of the reactor.
The fuel assemblies with damaged spacer grids will maintain a positive argin with respect to SEISMIC /LOCA loadings.
5 Completed a modification that changes the control power source to the starter coil controlling the NEMA size 3 MCC starter in the MCC S2B1, breaker 52-2B108, which powers and controls the emergency control room air conditioner and heating unit. This nodi fication will improve the probability of operation of the control room heating and ai r conditioning when needed.
There are no new failure modes for the equipment as a result of this change due to the power being supplied from the 120 VAC vital power source.
6.
Implemented a modification which limits the position of the reactor building purge valves so that allowable stresses will not be exceeded and adequate closure torques will be available.
This modification is the resui of a District commitment in response to an October, 1979 NRC letter. As a result of this change, the purge vaives can be partially opened during plant operation.
7 Incorporated a change which removed the letdown or!fice manual bypass valve for one cycle.
PLS-036, the letdown orifice manuel bypass, was removed and the line cap 7ed.
The original valve was removed due to gross erosion damage to the valve internals, caused by throttled operation.
The surrounding piping has radioactive crud captured in many = pots.
Capping the piping instead of welding in a replacement valve will sign!ficantly reduce the radiction exposure requi red to make this system operable.
This modification does not affect the safe operation of the plant since the orifice and remotely operated control valve provide the needed letdown flow.
This change does not involve an unreviewed safet f question or a change to the Technical Specifications.
7
. MAJOR ITEMS OF SAFETY-RELATED MAINTENANCE 1.
Repaired leaking OTSG tubes.
2.
' Secured loose jam nuts on main steam line snubber.
3 Replaced defective heater in diesel generator.
4.
Cas ibrated and checited faulty RCP speed indicator.
5 Rebuilt accumulator and valve block on RCP snubber.
6.
Replace leaking gasket on pressurizer spray valve; pressurizer level transmi tter, and second-of f isolat ion valve.
7 Replaced defective components in control for decay heat cooler bypass valve.
8.
Reset relief pressure on pressurizer code safety.
9 Replaced defective valve in nuclear service cooling water system.
10.
Replaced defective motor operator on pressurizer liquid sample isolation line.
Following the leak located visually in OTSG "B" Tube 77-17, eddy current examination was initiated in these categories:
1 19 tubes with previous OD indications were examined full length.
2.
370 tubes on three rows adjacent to both sides of the open lane were examined in the upper 12 feet, i.e. above the 13th tube support plate.
3 51 other randomly selected tubes that had previously been inspected were examined above the 13th TSP.
4.
Clusters of six peripheral tubes adjacent to the eight internal Auxillary Feedwater header lugs. were examined above the 13 TSP.
These were selected because of the recent Davis-Besse tube leak.
As a result of the examination, the following lane tubes were removed f rom service:
% THRU-WALL INDICATION 77-17 100 75-18 60-90 75-19 20-30 75-27 20-30 All four were explosively plugged at the lower tube sheet and stabilized by manually welding rods from the upper tube sheet extending down below the 14th TSP.
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