ML19305C253
| ML19305C253 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/13/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Clayton F ALABAMA POWER CO. |
| References | |
| NUDOCS 8003260409 | |
| Download: ML19305C253 (2) | |
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UNITED $TATES 3
8 NUCLEAR REGULATORY COMMISSION l
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101 MARIETTA ST N.W., SUITE 3100 o
ATLANTA. GEORGIA 30303 MAR 131980 In Reply Refer To:
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Alabama Power Company Attn:
F. L. Clayton Senior Vice President Post Office Box 2641 Birmingham, AL 35291 Gentlemen:
The enclosed Bulletin 80-06 is forwarded to you for action. A written response is required. If you desire additional information regarding this matter, please contact this office.
Sincerely, N.
(~~3 %7 James P. O'Reilly Director
Enclosures:
1.
List of IE Bulletins Recently. Issued
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MAR 131980 0
Alabama Power Company cc w/ encl:
A. R. Barton Executive Vice President Post Office Box 2641 Birmingham, Alabama 35291 R. P. Mcdonald Vice President-Nuclear Generation Post Office Box 2641 Birmingham, Alabama 35291 J. R. Campbell Manager-Operations Quality Assurance Post Office Box 2641 Birmingham, Alabama 35291 H. O. Thrash Manager-Nuclear Generation Post Office Box 2641 Birmingham, Alabama 35291 W. G. Hairston, III Plant Manager Drawer 470 Ashford, Alabama 36312 i :.q' l
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SSINS: 6820 Accession No.:
UNITED STATES 8002280639 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 13, 1980 4
IE Bulletin No. 80-06 ENGINEERED SAFETY FEATURE (ESF) RESET CONTROLS Description of Circumstances:
On November 7, 1979, Virginia Electric and Power Company (VEPCO) reported that following initiation of Safety Injection (SI) at North Anna Power Station.
Unit 1, the use of the SI Reset pushbuttons aloce resulted in certain ventila-tion dampers changing position from their safety or emergency mode to their normal mode. Further investigation by VEPCO and the architect-engineer resulted in discovery of circuitry which similarly affected components actuated by a Containment Depressurization Actuation (CDA, activated on Hi-Hi Containment Pressure). The circuits in question are listed below:
Component / System Problem Outside/Inside Recirculation Spray Pump motors will not start after Pump Motors actuation if CDA Reset is depressed prior to starting timer running out (approx. 3 minutes)
Pressurized Control Room Dampers will open on SI Reset Ventilation Isolation Dampers Safeguards Area Filter Dampers Dampers reposition to bypass filters when CDA Reset is depressed Containment Recirculation Cooler Fans will restart when CDA Reset Fans is depressed Service Water Supply and Discharge If service water is being used as Valves to Containment the cooling medium prior to CDA actuation, valves will reopen upon depressing CDA reset Service Water Radiation Monitoring Pumps will not start after Sample Pumps actuation if CDA reset is depressed prior to motor starting timers running out Main Condenser Air Ejector Exhaust After receiving a high radiation Isolation Valves to the Containment monitor alarm on the air ejector exhaust, SI actuation would shut these valves and depressing SI Reset would reopen them e
e IE Bulletin No. 80 06 March 13, 1980 Page 2 of 3 Review of circuitry for ventilation dampers, motors, and valves reported by VEPCO resulted in discovery of similar designs in ESF-actuated components at Surry Unit I and Beaver Valley; where it has been found that certain equipment would return to its normal mode following the reset of an ESF signal;
- thus, protective actions of the affected systems could be compromised once the associated actuation signal is reset.
These two plants had Stone and Webster Engineering Corporation for the architect-engineer as did the North Anna Units.
The Stone and Webster Engineering Corporation and VEPCO are preparing design changes to preclude safety-related equipment from moving out of its emergency mode upon reset of an Engineered Safety Features Actuation Signal (ESFAS).
This corrective action has been found acceptable by the NRC, in that, upon reset of ESFAS, all affected equipment remains in its emergency mode.
The NRC has performed reviews of selected areas of ESFAS reset action on PWR facilities and, in some cases, this review was limited to examination of logic diagrams and procedures.
It has been determined that logic diagrams may not adequately reflect as-built conditions; therefore, the requested review of drawings must be done at the schematic / elementary diagram level.
There have been several communications to licensees from the NRC on ESF reset actions. For example, some of these communications have been in the form of Generic Letters issued in November, 1978 and October, 1979 on containment venting and purging during normal operation.
Inspection and Enforcement Bulletins Nos. 79-05, 05A, 05B, 06A, 06B and 08 that addressed the events at THI-2 and NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations. However, each of these communications has addressed only a limited area of the ESF's.
We are requesting that the reviews undertaken for this Bulletin address all of the ESF's.
Actions To Be Taken By Licensees:
For all PWR and BWR facilities with operating licenses:
1.
Review the drawings for all systems serving safety-related functions at the schematic level to determine whether or not upon the reset of an ESF actuation signal, all associated safety-related equipment remains in its emergency mode.
2.
Verify the actual installed instrumentation and controls at the facility are consistent with the schematics reviewed in Item 1 above by conducting a test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of the various isolating or actuation signals. Provide a schedule for the performance of the testing in your response,to this Bulletin.
3.
If any safety-related equipment does not remain in its emergency mode upon reset of an ESF signal at your facility, describe proposed system modification,, design change, or other corrective action planned to resolve the problem.
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e IE Bulletin No. 80-06 March 13, 1980 Page 3 of 3 4.
Report in writing within 90 days, the results of your review and include a list of all devices which respond as discuss'ed in item 3 above, actions taken or planned to assure adequate equipment control, and a schedule for implementation of corrective action. This information is requested under the provisions of 10 CFR 50.54(f). Accordingly, you are requested to provide within the time period specified above, written statements of the above information, signed under oath or affirmation. Reports shall be submitted to the Director of the appropriate NRC Regional Office and a copy shall be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C.
20555.
For all power reactor facilities with a construction permit, this Bulletin is for information only and no written response is required.
Approved by GAO, B180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic problems.
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IE Bulletic No. 80-06 Enclosure March 13, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.
80-05 Vacuum Condition Resulting 3/10/80 All PWR power reactor In Damage To Chemical Volume facilities holding Control System (CVCS) Holdup OLs and to those with Tanks a CP 79-01B Environmental Qualification 2/29/80 All power reactor of Class IE Equipment facilities with an OL 80-04 Analysis of a PWR Main 2/8/80 All PWR reactor facilities Steam Line Break With holding OLs and to those Continued Feedwater nearing licensing Addition 80-03 Loss of Charcoal From 2/6/80 All holders of Power Standard Type II, 2 Inch, Reactor OLs and cps Tray Adsorber Cells 80-02 Inadequate Quality 1/21/30 All BWR licenses with Assurance for Nuclear a CP or OL 80-01 Operability of ADS Valve 1/11/80 All BWR power reactor Pneumatic Supply facilities with and OL 79-01B Environmental Qualification 1/14/80 All power reactor of Class IE Equipment facilities with an OL 79-28 Possible Malfunction of 12/7/79 All power reactor Namco Model EA 180 Limit facilities with an Switches at Elevated OL or a CP Temperatures 79-27 Loss Of Non-Class-1-E 11/30/79 All power reactor Instrumentation and facilities holding Control Power System Bus OLs and to those During Operation nearing licensing 79-26 Boron Loss From BWR 11/20/79 All BWR power reactor Control Blades facilities with an OL 79-25 Failures of Westinghouse 11/2/79 All power reactor BFD Relays In. Safety-Related facilities with an Systems OL or CP 1
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