ML19312E558
| ML19312E558 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/09/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Mcdonald R ALABAMA POWER CO. |
| References | |
| NUDOCS 8006050309 | |
| Download: ML19312E558 (2) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION
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REGION 11
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101 MARIETTA ST., N.W., SUITE 3100 O
ATLANTA, GEORGIA 30303 NAY 9 1980 In Reply Refer To:
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Alabama Power Company Attn:
R. P. Mcdonald Vice President-Nuclear Generation Post Office Box 2641 Birmingham, Alabama 35291 Gentlemen:
The enclosed Bulletin 80-12 is forwarded to you for information. No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, James P. O' Reil Director
Enclosures:
1.
IE Bulle*in No. 80-12 2.
List of Recently Issued IE Bulletins t
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8006050 309 D
MY 9 1989 Alabama Power Company,
cc w/ encl:
A. R. Barton Executive Vice President Post Office Box 2641 Birmingham, Alabama 35291 H. O. Thrash Manager, Nuclear Generation Post Office Box 2641 Birmingham, Alabama 35291
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F. L. C1ayton, Jr.
Senior Vice President Post Office Box 2641 Birmingham, Alabama 35291 W. C. Petty Manager-Quality-Design Assurance (Design and Construction)
Post Office Box 2641 Birmingham, Alabama 35291 W. G. Hairston, III Plant Manager Drawer 470 Ashford, Alabama 36312 R. E. Hollands, Jr.
QA Supervisor Post Office Box U Ashford, Alabama 36312
- 0. D. Kingsley, Jr.
Manager, Nuclear Engineering and Technical Services Post Office Box 2641 Birmingham, Alabama 35291
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UNITED STATES SSINS No.: 6820 NUCLEAR REGULATORY COMMISSION Accession No.:
OFFICE OF INSPECTION AND E.VFORCEMENT 8005050053 WASHINGTON, D.C. 20555 May 9, 1980 IE Bulletin No. 80-12 DECAY HEAT REMOVAL SYSTEM OPERABILITY
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Introduction:==
The intent of this Bulletin is to improve nuclear power plant safety by reducing the likelihood of losing decay heat removal (DHR) capability in operating pressurized water reactors (PWRs).
PWRs are most susceptible to losing DHR capability when their steam generators or other diverse means of removing decay heat are not readily available. Such conditions often occur when the plants are in a refueling or cold shutdown mode, and during which time concurrent maintenance activities are being performed.
There is a need to assure that all reasonable means have been taken to provide redundant or diverse means of DHR during all modes of operation.
(Note: A redundant means could be provided by having DHR Train A AND Train B operable; a diverse means could be provided by having either DHR Train A OR Train B operable AND a steam generator available for DER purposes.) There is also need to assure that all reasonable means have been taken to preclude the loss of DHR capability due to common mode failures during all modes of operation.
Background:
On several occasions, operating PWRs have experienced losses of DHR capability.
In each instance, except that of the Davis-Besse Unit 1 incident of April 19, 1980, DHR capability was restored prior to exceeding the specified RCS temper-ature limit for the specific mode of operation. Nonetheless, the risk and frequency associated with such events dictate that positive actions be taken to preclude their occurrence or at least ameliorate their effects.
The most noteworthy example of total loss of DHR capability occurred at Davis-Besse Unit 1 on April 19, 1980.
(See IE Inforniation Notice No. 80-20, attached hereto as Enclosure 1).
Two factors identified as major contributors to the Davis-Besse event in the Information Notice are:
(1) extensive maintenance activities which led to a loss of redundancy in the DHR capability, and (2) inadequate procedures and/or administrative controls which, if corrected, could have precluded the event or at least ameliorated its effects.
ACTIONS TO BE TAKEN BY LICENSEES OF PWR FACILITIES:
1.
Review the circumstances and sequence of events at Davis-Besse as des-cribed in Enclosure 1.
2.
Review your facility (zes) for all DHR degradation events experienced, especially for events similar to the Davis-Be.22 incident.
IE Bulletin No. 80-12 May 9, 1980 Page 2 of 3 3.
Review the hardware capability of your facility (ies) to prevent DHR loss events, including equipment redundancy, diversity, poor source reliability, instrumentation and control reliability, and overall reliability during the refueling and cold shutdown modes of operation.
4 Analyze your procedures for adequacy of safeguarding against loss of redundancy and diversity of DHR capability.
5.
Analyze your procedures for adequacy of responding to DHR loss events.
'Special emphasis should be placed upon responses when maintenance or refueling activities degrade the DHR capability.
6.
Until further notice or until Technical Specifications are revised to resolve the issues of this Bulletin, you should:
Implement as soon as practicable administrative controls to assure a.
that redundant or diverse DHR methods are available during all modes of plant operation; (Note: When in a refueling mode with water in the refueling cavity and the head removed, an acceptable means could include one DHR train and a readily accessible source of borated water to replenish any loss of inventory that might occur subsequent to the loss of the available DHR train.)
b.
Implement administrative controls as soon as practicable, for those cases where single failures or other actions can result in only one DHR train being available, requiring an alternate means of DHR or expediting the restoration of the lost train or method.
7.
Report to the NRC within 30 days of the date of this Bulletin the results of the above reviews and analyses, describing:
Changes to procedures (e.g., emergency, operational, administrative, a.
maintenance, refueling) made or initiated as a result of your reviews and analyses, including the scheduled or actual dates of accomplish-ment; (Note: NRC suggests that you consider the following: (1) limiting maintenance activities to assure redundancy or diversity and integrity of DHR capability, and (2) bypassing or disabling, where applicable, t
automatic actuation of ECCS recirculation in addition to disabling High l
Pressure Injection and Containment Spray preparatory to the cold l
shutdown or refueling mode.)
b.
The safeguards at your facility (ies) against DHR degradation, including your assessment of their adequacy.
The above information is requested pursuant to 10 CFR 50.54(f). Accordingly, written statements addressing the above items shall be signed under oath or affir-mation and submitted within the time specified above. Reports shall be submitted
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IE Bulletin No. 80-12 May 9, 1980 Page 3 of 3 J
to the director of the appropriate NRC regional office, and a copy forwarded to the Director, Division of Reactor Operations Inspection, NRC Office of Inspection and Enforcement, Washington, D. C. 20555.
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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Enclosure:
i IE Information Notice No. 80-20 I
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IE Bulletin No. 80-12 Enclosure May 9, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subj ect Date Issued Issued To No.
80-12 Decay Heat Removal System 5/9/80 Action for all Operability
- PWR's w/0L.
In fo r-mation for all PWR's
.with Construction Permit. and all BWR's.
80-11 Masonay Wall Design 5/8/80 All power reactor facilities with an OL, except Trojan 80-10 Contamination of 5/6/80 All power reactor Nonradioactive System and facilities with an Resulting Potential for OL or CP Unmonitored, Uncontrolled Release to Environment 80-09 Hydramotor Actuator 4/17/80 All power reactor Deficiencies operating facilities and holders of power reactor construction permits 80-08 Examination of Containment 4/7/80 All power reactors with Liner Penetration Welds a CP and/or OL no later than April 7, 1980 80-07 BWR Jet Pump Assembly 4/4/80 All GE BWR-3 and Failure BWR-4 facilities with an OL 80-06 Engineered Safety Feature 3/13/80 All power reactor (ESF) Reset Controls facilities with an OL 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-OlB Environmental Qualification 2/29/80 All power reaccor 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
SSINS No.:
6870 UNITED STATES Accession No.:
NUCLEAR REGULATORY COMMISSION 8002280671 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C. 20555 Hay 8, 1980 IE Information Notice 80-20 i
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' ~ ' ~ LOSS 0F DECAY HEAT REMOVAL CAPABILITY AT DAVIS-BESSE UNIT'l WHILE IN A
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REFUELING MODE Description of Circumstances:
On April 19, 1980, decay heat removal capability was lost at Davis-Besse Unit I for approximately two and one-half hours. At the time of the event, the unit was in a refueling mode (e.g., RCS temperature was 90F; decay heat was being removed by Decay Heat Loop No. 2; the vessel head was detensioned with bolts in place; the reactor coolant level was slightly below the vessel head flanges; and the manway covers on top of the once through steam generators were removed).
(See Enclosure A, Status of Davis-Besse 1 Prior to Loss of Power to Busses E-2 and F-2 for additional details regarding this event.)
Since the plant was in a refueling mode, many systems or components were out of service for maintenance or testing purposes.
In addition, other systems and components were deactivated to preclude their inadvertent actuation while in a refueling mode. Systems and components that were not in service or deactivated included:
Containment Spray System; High Pressure Injection System; Source Range Channel 2; Decay Heat Loop No. 1; Station Battery IP and IN; Emergency Diesel-Generator No. 1; 4.16 KV Essential Switchgear Bus C1; and l
13.8 KV Switchgear Bus A (this bus was energized but not aligned).
In brief, the event was due to the tripping of a non-safeguards feeder breaker in 13.8 KV Switchgear Bus B. Because of the extensive maintenance and testing activities being conducted at the time, Channels 1 and 3 of the Reactor Protec-tion System (RPS) and Safety Features Actuation System (SFAS) were being ener-gized from only one source, the source emanating from the tripped breaker.
Since the SFAS logic used at Davis-Besse is a two-out-of-four input scheme in which the loss (or actuation) of any two input signals results in the actuation of all four output channels (i.e., Channels 1 and 3, and Channels 2 and 4), the l
loss of power to Channels 1 and 3 bistables also resulted in actuation of SFAS l
Channels 2 and 4.
The actuation of SFAS Channels 2 and 4, in turn, affected l
Decay Heat Loop No. 2, the operating Ic-
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