ML17095A355
| ML17095A355 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/09/1980 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Crane P PACIFIC GAS & ELECTRIC CO. |
| References | |
| NUDOCS 8005200291 | |
| Download: ML17095A355 (16) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V 1990 N. CALIFORNIABOULEVARD SUITE 202, WALNUTCREEK PLA2A WALNUTCREEK, CALIFORNIA94596 Hay 9, 1980 Docket Nos. 50-275, 50-323 Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94106 Attention:
Hr. Philip A. Crane, Jr.
Assistant General Counsel 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, R.
H. Engelken Director
Enclosures:
l.
List of Recently Issued IE Bulletins E.
B. Langley, Jr.,
PG8E W.
- Raymond, PG8E R.
- Ramsey, PGSE, Diablo Canyon
I
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 May 9, 1980 SSINS No.:
6820 Accession No.:
8005050053 IE Bulletin No. 80-12 DECAY HEAT REMOVAL SYSTEM OPERABILITY
==
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 DHR purposes.)
There is also need to assure tttat all reasonable means have Seen -taken to preclude the loss of DHR capability due to common mode failures during all modes of operation.
Ba'~~r ound 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 Information 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(ies) for all DHR degradation events experienced, especially for events similar to the Davis-Besse incident.
IE Bulletin Ho. 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, power 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 OHR 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:
a ~
Implement as soon as practicable administrative controls to assure 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.)'.
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 HRC within 30 days of the date of this Bulletin the results of the above reviews and analyses, describing:
a 0 Changes to procedures (e.g.,
emergency, operational, administrative, maintenance, refueling) made or initiated as a result of your reviews and analyses, including the scheduled or actual dates of accomplish-ment; (Note:
HRC suggests that you consider the following: (I) limiting maintenance activities to assure redundancy or diversity and integrity of OHR capability, and (2) bypassing or disabling, where applicable, automatic actuation of ECCS recirculation in addition to disabling High Pressure Injection and Containment Spray preparatory to the cold shutdown or refueling mode.)
b.
The safegua} ds 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
IE Bulletin No. 80-12 May 9, 1980 Page 3 of 3 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.
Enclosure:
IE Information Notice No. 80-20
IE Bulletin No. 80-12 May 9, 1980 RECENTLY ISSUED IE BULLETINS Enclosure Bulletin No.
Subject Date Issued Issued To 80-11 80-10 80-09 Masonay Wall Design Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release to Environment Hydramotor Actuator Deficiencies 5/8/80 5/6/80 4/17/80 All power reactor facilities with an OL, except Trojan All power reactor facilities with an OL or CP All power reactor operating facilities and holders of power reactor construction permits 80-08 Examination of Containment 4/7/80 Liner Penetration Welds All power reactors with a
CP and/or OL no later than April 7, 1980 80-07 BWR Jet Pump Assembly Failure 4/4/80 All GE BWR-3 and BWR-4.facilities with an OL 80-06 80-05 79-01B Engineered Safety Feature 3/13/80 (ESF)
Reset Controls Vacuum Condition Resulting 3/10/80 In Damage To Chemical Volume Control System (CVCS) Holdup Tanks Environmental qualification 2/29/80 of Class IE Equipment All power reactor facilities with an OL All PWR power reactor facilities holding OLs and to those with a
CP All power reactor facilities with an OL 80-04 80-03 Analysis of a PWR Main Steam Line Break With Continued Feedwater Addition Loss of Charcoal From Standard Type II, 2 Inch, Tray Adsorber Cells 2/8/80 2/6/80 All PWR reactor facilities holding OLs and to those nearing licensing All holders of Power Reactor OLs and CPs
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT PJASHINGTON, D.
C.
20555 May 8,,
1980 SSINS No.:
6870 Accession No.:
8002280671 IE Information Notice 80-20 LOSS OF DECAY HEAT REMOVAL CAPABILITY AT DAVIS-BESSE UNIT 1 WHILE IN A REFUELING MODE Description of Circumstances:
On April 19,
- 1980, decay heat removal capability was lost at Davis-Besse Unit 1 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 deactiva'ted included:
Containment Spray System; High Pressure Injection System; Source Range Channel 2;
Decay Heat Loop No. 1; Station Battery 1P and 1N; Emergency Diesel-Generator No. 1;
- 4. 16 KV Essential Switchgear Bus Cl; and 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 loss of power to Channels 1 and 3 bistables also resulted in actuation of SFAS Channels 2 and 4.
The actuation of SFAS Channels 2 and 4, in turn, affected Decay Heat Loop No. 2, the operating loop.
Since the initiating event was a loss of power event, all five levels of SFAS were actuated (i.e., Level 1 - High Radiation; Level 2 - High Pressure Injec-tion; Level 3 - Low Pressure Injection; Level 4 - Containment Spray; and
IE Information Notice No. 80-20 May 8, 1980 Page 2 of 3 Level 5 -
ECCS Recirculation Mode).
Actuation of SFAS Level 2 and/or 3
resulted in containment isolation and loss of normal decay heat pump suction from RCS hot leg No. 2.
Actuation of SFAS Level 3 aligned the Decay fleat Pump No.
2 suction to the Borated Water Storage Tank (BWST) in the low pressure injection mode.
Actuation of SFAS Level 5 represents a low level in the BWST; therefore, upon its actuation, ECCS operation was automatically transferred from the Injection Mode to the Recirculation Mode.
As a result, Decay Heat Pump No. 2, the operating
- pump, was automatically aligned to take suction from the containment sump rather than from the BWST or the reactor coolant system.
Since the emergency containment sump was dry, suction to the operating decay heat pump was lost.
As a result, the decay heat removal capability was lost for approximately two and one-half hours, the time required to vent the system.
Furthermore, since Decay Heat Loop No.
1 was down for maintenance, it was not available to reduce the time required to restore decay heat cooling.
MAJOR CONTRIBUTORS TO THE EVENT:
The rather extended loss of decay heat removal capability at Davis-Besse Unit 1 was due to three somewhat independent factors, any one of which, if corrected, could have precluded this event.
These three factors are:
(i) Inadequate procedures and/or administrative controls; (ii) Extensive maintenance activities; and (iii) The two-out-of-four SFAS logic.
Regarding inadeouate procedures and/or administrative controls, it should be noted that the High Pressure Injection Pumps and the Containment Spray Pumps were deactivated to preclude their inadvertent actuation while in the refuel-ing mode.
In a similar vein, if the SFAS Level 5 scheme had been by-passed or deactivated while in the refueling mode, or if the emergency sump isolation valves were closed and their breakers
- opened, this event would have been, at
- most, a minor interruption of decay heat flow.
Regarding the extensive maintenance activities, it appears that this event would have been precluded, or at least ameliorated, if the maintenance activi-ties were substantially reduced while in the refueling mode.
For example, if the maintenance activities had been restricted such that two SFAS channels would not be lost by a single event (e.g., serving Channels I and 3 from separate sources),
this event would have been precluded.
Likewise, if maintenance activities had been planned or restricted such that a backup decay heat removal system would have been readily available, the consequences of the loss of the operating decay heat removal loop would have been ameliorated.
Regarding the two-out-of-four SFAS logic used at Davis-Besse, even under normal conditions, it appears that this type of logic is somewhat more suscep-tible to spurious actions than other logic schemes (e.g.,
a one-out-of-two taken-twice scheme).
This susceptibility is amplified when two SFAS channels are served from one source.
Consequently, when the source feeding SFAS Channels I and 3 was lost, all five levels of SFAS were actuated..
As stated
Q
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IE Information Notice No. 80-20 Hay 8, 1980 Page 3 of 3 previously, this particular event would have been precluded if SFAS Channels 1
and 3 were being served from separate and independent sources.
In a similar vein, this specific event would have been precluded by a one-out-of-two taken twice type of logic that requires the coincident actuation of or loss of power of an even numbered SFAS Channel and an odd numbered SFAS Channel.
Since each LllR can be expected to be in a refueling mode many times during its lifetime, licensees should evaluate the susceptibility of their plants to losing decay heat removal capability by the causes described in this Informa-tion hotice.
No specific action or response is requested at this time.
Licensees having questions regarding this matter should contact the director of the appropriate NRC Regional Office.
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