IR 05000312/1980016
| ML19318B267 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/20/1980 |
| From: | Canter H, Faulkenberry B, Obrien J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML19318B263 | List: |
| References | |
| 50-312-80-16, NUDOCS 8006250184 | |
| Download: ML19318B267 (7) | |
Text
MAY J 01980
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U. S. NUCLEAR REGULATORY COM!ilSSION
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OFFICE OF INSPECTION AND ENFORCEMENT Report No. 50-312/80-0T/6 Docket No. 50-312 License No. DPR-54 Safeguards Group Licensee:
Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 94813 Facility Name:
Rancho Seco Generating Station Inspection at: Herald, California (Rancho Seco Site)
Inspection conducted:.
April 1-30, 1980 Inspectors: [h,v4 kw~
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Harvey L. Genter Senior Resident Inspection
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O ChJ John P. O',Bri n, Unit Resident Inspection
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ate Signed Ac.
'V X NC (April 28-30 only) John C.arlson, R,ea r Inspection Date Signed f
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f fC (April 28-30 only) Gerald Zwetzig Reactor Inspection
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Approved By:Ml c/ b2..
.c/2e/fs B. H. Faulkenberry, Chief,_ Reactor Operations and Date Signed Nuclear Support Branch
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Suninary:
Inspection between April 1 and 30,1980 (Report No. 50-312/80-16)
Areas inspected: Routine inspections of long-term shutdown activities; followup on a Regional Reauest item; and, independent inspection effort.
The inspection involved 106 inspector hours by the Senior Resident Inspector and 54 inspector hours by other NRC Inspectors.
Results: Of the three areas inspected, no items of noncompliance or deviations were identified.
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DETAILS 1.
Persons Contacted
- R. Rodgriguez, Manager, fluclear Operations P. Oubre', Plant Superintendent D. Blachly, Mechanical Engineer
- N. Brock, Electrical /I&C Maintenance Supervisor
- Q. Coleman, Quality Assurance Engineer
- R. Colombo, Technical Assistant G. Coward, Maintenance Supervisor
- W. Ford, Operating Supervisor H. Heckart, Engineering Technician J. Jewett, Scolor Quality Assurance Engineer
- J. McColligan, Mechanical Engineering Supervisor
- R. Medina, Quality Assurance Engineer R. Miller, Chemistry /Rhdiological Supervisor
- L. Schwieger; quality Assurance Director
- J. Sullivan, Quality Assurance Supervisor D. Whitney, fluclear Engineer B. Wichert, Mechanical Engineer The inspectors also talked with and interviewed several other licensee employees, including members of the engineering, maintenance, operations, and quality assurance (QA) organizations.
- Denotes those attending the Exit Interview on April 30, 1980.
2.
Long-Term Shutdown Activities The Rancho Seco plant has been shut-down since January 12, 1980, for Cycle 4 refueling.and plant modifications.
Daily, as appropriate, the inspectors observed control room instrumentation, manning, and procedural compliance by operators. Logs and operating records were -
examined and the cleanliness of radiation controlled area access points was observed.
Several surveillance tests were observed including portions of flon-fluclear Instrumentation (flNI) Tests and Diesel Generator Sequencing
Tests. The inspector queried various licensed operators to determine their knowledge of recent changes to procedures, facility configuration and plant conditions.
Tours of accessible areas were taken to make independent assessments of equipment conditions, plant conditions, radiological controls, security, safety and adherence to regulatory requirements.
Plant housekeeping conditioils were observed, especially in the areas of potential ignition sources.
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Physical security implementation was examined.
Isolation zones were
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noted to be clear, vehicles were found to be properly controlled, and vital area access controls appeared appropriate for the sample of areas observed by the inspector.
Compensatory measures appeared to be employed as appropriate including fire door alarm responses.
Maintenance on major plant compenents was performed during the month of April. The inspectors witnessed portions of the maintenance performed on Reactor Coolant Pumps, Control Rod Drives, Diasel Generators, and the Seismic Recording System.
No items of noncompliance or deviations were identified.
3.
Followup co Regional Request It.em On April 14, 1980, the Commission issued a Confirmatory Order concerning Rancho Seco's commitments to implement certain actions prior to restart from the current refueling outage. The Order was issued as a result of the experience gained from the Crystal River 3 incident of February 26, 1980, wherein a non-nuclear instrumentation power loss resulted in a series of unexpected events. The NRC staff and the licensee developed three commitments which were the subject of the Confirmatory Order:
(1) Actions which will allow the operator to cope with various combinations of loss of instrumentation and control functions.
(a) Equipment and control systems to give clear indications of functions which are lost or are unreliable.
(b) Procedures and training to assure positive and safe manual response by the operator in the event that instruments are unavailable.
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(2) Determination of the effects of various combinations of loss of instrumentation and control functions by design review analysis and verifications by test.
(3) Corrections of electrical deficiencies which may allow the Power Operated Relief Valve (PORV) and pressurizer spray valves to open on non-nuclear instrumention power failures, such as, the event which occurred at Crystal River, Unit 3, on February 26, 1980.
The following actions were taken by the licensee to satisfy the above listed commitments:
(1) The Bailey Meter Company (vendor for flNI's) supplied installation
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instructions on the proper method of NNI buffer card replacement.
(Improper buffer card replacement may have initiated the Crystal River 3 event). All buffer cards were inspected for proper installation and the Bailey instructions were added to applicable procedure _ _ _
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(2). The fluclear operations group developed rectangular foam rubber plugs to be used to insert in or cover the open backlighted pushbutton (BLPB) modules whenever the lamp bulb section of the module is lifted out. Also, the upper bulb fixture portion of the module _is removed to a remote location for bulb replacement.
(3) STP 610 was performed on the original NNI-Y power supply system to verify various design conditions which were used to supply information for the present modified NNI system.
(4) The majority of the 5 amp fuses in the NNI system were replaced with 0.75 amp fuses. This should allow for faster clearing of faults and prevent tripping of the 120 volt AC input circuit breakers.
In other words, protective device coordination should occur.
(5) ffMI instrument power supply modifications were performed. The new design fuses all circuits leaving the NNI panels and all internal AC circuits.
Enclosure 1 to this report is excerpted from a March-12, 1980 letter from W. C. Walbridge, General Manager, to H. R. Denton, Director of NRR. This figure is a one-line diagram of the 1980 NNI power supply modifications. The key change that was made to the previous system is that the power supply and monitor arrangement on the right portion of the drawing (NNI-Z) is now available upstream of S1 and S2 such that the instrument selector switches, indicating lamps and auxiliary relays are capable of l
receiving power from more than one source through an ABT (Automatic Bus Transfer Switch).
(6) New instrumentation has been installed with readout capability on the plant computer in the control room. The following is a list of this instrumentation. All listed instruments except the uncompensated pressurizer level are completely independent of both NNI-X and NNI-Y power.*
Instruments **
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Uncompensated pressurizer levels 0-320 inches water Wide range RCS pressure 0-2500 psig Wide range RCS Loop A T 50-650 degrees F c
Wide range RCS Loop B T 50-650 degrees F RCS Loop A T 520-620 degrees F h
RCS Loop B T 520-620 degrees F OTSGAStartDplevel 0-600 inches water OTSG B Startup level 0-600 inches water OTSG A pressure 0-1200 psig 0TSG B pressure 0-1200 psig Makeup Tank level 0-j00%6 Source range nuclear instruments 10 -10 cps
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- Available on the computer, on a spare panel in the computer room adjacent to the control room proper, and on the shutdown panel.
Incore thermocouple values with a range of 0-2000 degrees F. are also available on the computer in the control room.
- The uncompensated pressurizer level inputs, although not completely independent of both NNI-X and NNI-Y are designed so that there will always be two uncompensated pressurizer levels available for loss of NNI-X or NNI-Y. That is, out of four uncompensated pressurizer level inpiits on the control room computer, one is totally independent of NNI-X or Y, two are supplied by NNI-X and one is supplied by NNI-Y.
(7) Directions have been placed on the computer console and in casualty procedures directing operators to place the following points on 1.he six computer trend recorders.during loss of NNI events:
Uncompensated pressurizer level (independent of NNI-Y and X)
Wide range RCS Pressure Wide range RCS Loop A Tc Wide range RCS Loop B Tc OTSG A startup level OTSG B startup level In addition, group 15 on the computer has been set aside for the following printouts at one minute intervals (by procedure).
RCS Loop B Hot Leg Temp.
Incore thermocouples RCS SFAS (A) Wide Range Pressure Make-up Tank level 0TSG A&B Outlet Steam Pressure Source Range NI's Pressurizer level (uncompensated)
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These items print out on the line printer and function independently from the alarm printer, therefore an overload on the line printer due to alarms on the alarm printer is not possible.
(8). Casualty procedures C43, 44, and 45, were written to handle the loss of NNI-X, NNI-Y, and all NNI's respectively.
Licensed operators were trained in March 1980, on these changes to procedures and the modified plant NNI system.
(9) STP-616 was performed to prove independence of the above listed instrumentation on NNI power upsets.
(10) The PORV circuitry was modified to close the valve on NNI failure.
-(11) The pressurizer spray valves circuitry was modified such that an NNI power supply failure should not cause either of the two spray valves to open..
No items of noncompliance or deviations were identifie.-
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4.
Independent Inspection Effort Discussions were held between the Senior Resident Inspector and operations, security, and maintenance personnel in an attempt to better understand problems they may have which are related to nuclear safety. These discussions will continue as a standard practice.
On various occasions during the month of April, the Senior Resident Inspector attended outage status meetings. These meetings are held by the Outage Coordinator to provide all disciplines onsite with a shift by shift update on the plant status and ongoing maintenance work. The outage status meetings were given twice a day until mid-April, when the once-a-day operations status meetings were recommenced in conjunction with the plant's coming out of cold shutdown following.the refueling
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In addition to the above, independent inspection effort was performed on
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the following items:
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(1) The Performance Appraisal Team from I&E Headquarters conducted inspections at Rancho Seco during the weeks of April 14 and i
April 21,1980. The inspectors attended numerious interviews and the preliminary exit meetings that were held onsite on April 21, 1980, and at the Sacramento Municipal Utility District Corporate Offices on April 24, 1980.
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(2) 179 Engineering Change Notices (ECN's) were examined by the inspectors.
Of the 179 ECN's, examined, 22 Class-1 ECN's, had
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been declared as not requiring 10 CFR 59.59 reviews by the Plant l
Review Committee (PRC) and 22 Class-1 ECN's had been declared as requiring those reviews.
All of the 179 ECN's had been screened and evaluated by the Engineering and Quality Control Supervisor (or Technical Support Supervisor) or his designated representative.
No items of noncompliance or safety issues were identified.
5.
Exit Interview The Senior Resident Inspector met with the licensee representatives denoted in Paragraph 1 on April 30, 1980.
During this meeting, the inspector sunmarized his findings of the April, 1980 inspection effort.
No items of noncompliance or deviations were identified.
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138 VAC
--118 VAC lilSTRUMENT 118 VAC FIGURE 4 INVERTER J POWER SUPPLY INVERTER D 1980 POWER SUPPLY MODIFICATION
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POWER 0'0WEo POWER POWER POWER POWER SUPPLY POWER POWER SUPtLY SUPPLY SUPPLY SUPPLY SUPPLY MONITOR SUPPLY SUPPLY
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ENCLOSURE 1 LAMPS AND AUXILIARY RELAYS
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