ML14170A666
| ML14170A666 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/09/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jackie Jones CAROLINA POWER & LIGHT CO. |
| References | |
| NUDOCS 8006050289 | |
| Download: ML14170A666 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION o
REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA. GEORGIA 2n2f In Reply Refer To:
MAY 9 1980 RII:JPO
,Carolina Power and Light Company ATTN:
J. A. Jones Senior Executive Vice President and Chief Operating Officer 411 Fayetteville Street Raleigh, North Carolina 27602 Gentlemen:
Enclosed is IE Bulletin No. 80-12 which requires action by you with regard to your PWR power reactor facility(ies) with an operating license.
Should you have any questions regarding the Bulletin or the actions required by you, please contact this office.
Sincerely, James P. O'Reilly Director
Enclosures:
- 1.
- 2.
List of Recently Issued IE Bulletins I
8006 05028
- MAY 9 18 Carolina Power and 4
Light Company
-2 cc w/encl:
R. B. Starkey, Jr., Plant Manager Post Office Box 790 Hartsville, South Carolina 29550
UNITED STATES SSINS No.:
6820 NUCLEAR REGULATORY COMMISSION Accession No.:
OFFICE OF INSPECTION AND ENFORCEMENT 8005050053 WASHINGTON, D.C. 20555 May 9, 1980 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 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 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 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, 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 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:
- 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.)
- 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:
- a.
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: 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, 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 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
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 Enclosure May 9, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.
80-12 Decay Heat Removal System 5/9/80 Action for all Operability PWR's w/OL. Infor
-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-O1B 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 Codtinued Feedwater nearing licensing Addition
SSINS No.:
6870 UNITED STATES Accession No.:
NUCLEAR REGULATORY COMMISSION 8002280671 OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C. 20555 May 8, 1980 IE Information Notice 80-20 LOSS OF DECAY HEAT REMOVAL CAPABILITY AT DAVIS-BESSE-UNITIT 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 I 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 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 Heat 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 inadequate 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 1 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 1 and 3 was lost, all five levels of SFAS were actuated. As stated
IE Information Notice No. 80-20 May 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 LWR 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 Notice. No specific action or response is requested at this time.
Licenshes having questions regarding this matter should contact the director of the appropriate NRC Regional Office.
Enclosure:
Davis-Besse Event of April 19, 1980
Enclosure A DAVIS-BESSE EVENT OF APRIL 19, 1980 STATUS OF DAVIS-BESSE 1 PRIOR TO LOSS OF POWER TO BUSSES E-2 AND F-2:
- 1.
Refueling mode with RCS temperature at 90o and level slightly below vessel head flange. Bead detensioned with bolts in place. Manway cover on top of OTSG removed. Tygon tubing attached to lower vents of RCS hot leg for RCS level indication.
Decay heat loop 2 in service for RCS cooling.
- 2.
All non-nuclear instrument (NNI) power and Static Voltage Regulator YAR supplied from 13.8 KV Bus B via HBBF2. 13.8 K Bus A energized but not connected.
RPS and SFAS Channels 1 and 3 being supplied from YAR.
- 3.
Equipment Out of Service
- a. Source Range Channel 2 - Surveillance
- b. Emergency Diesel Generator 1 - Maintenance.
- c. Decay Beat Loop 1 - Maintenance.
- 4.
Breakers for containment spray and EPI pumps racked out.
SEQUENCE OF EVENTS TIME EVENT CAUSE/COMMENTS 2:00 p.m.
Loss of power to Ground short on 13.8 KV breaker HBBF2 Busses E-2 and F-2 which caused breaker to open. This (non-essential 480 interrupted power to busses E-2 and 7-2 VAC) which were supplying all non-nuclear*
instrument (NNI) power, channels 1 and 3 of the Reactor Protection System (RPS) and
.the Safety Features Actuation Signal (SFAS),.
the computer, and much of the control room indicators.
2:00 p.s.
SPAS Level 5 (recircu-Two out of four logic tripped upon loss lation mode) actua-of Busses E-2 and 7-2. Actuation caused tion.
ECCS pump suction valves from containment sump to open and ECCS pump suction valves from Borated Water Storage Tank to close.
During valve travel times, gravity flow path existed from BWST to containment sump.
2:02 p m.
Decay Beat (10 Operator turned off only operating DE pressure safety in-pump to avoid spillage of RCS water to jectiqk) flo secured containment via the tygon tubing for RCS
-by opeator level indication and open SG manway.
2:33 p.m.
Partial restoration..
of power Power to Bus E-2 and SFAS chnnels 1 and3 restored along with one channel of NNI.
This restored all essential power for ECCS.
-2g T tY EVENT CAUSE/COKMINTS 2:44 p.m.
Attempt to Teestab-Started DM pump 1-2 then stopped it when lish DH flow it was deter-mined that air was in suction line.
Pump secured to prevent damage.
3:34 p.m.
Source Range Channel 2 energized.
4:00 p.m.
Restoration of Busses Busses restored sequentually as efforts to "(480 VAC) 7-2, F-21, progressed to isolate ground fault.
4:06 p.m.
7-22t and F-23 4:25 p.m.
DH flow restored DR pump 1-2 starteg after venting. RCS temperature at 170 F.DH flow by-passing noI5
, Icore TC's being taken and maximum 4:46 p.m.
Containment sump Precautionary measure to assure containment pump breakers sump water from BWST remained in sontainment.
opened Incore Me' range from 161 to 164 F.
5:40 p.m.
Computer returned to Incore TC's range from 158 to 1600F.
service.
6:24 p.m.
DH flow directed RCS cooldown established at less than 25 o through cooler per hour.
RCS te erature at 1500F. Incore TC's range from 151 to1a580F.
(9:50 P.M.
Power completely RCS temperature at approximately 1150F.
re fto red STATUS 0F DAVIS-BESSE I AFTER RtECOVERY FROM LOSS OF POWJER TO BUSSES E-2 AND 7-2:
- 1. Refueling mode with ACS'temperature at 1150r and level slightly below vessel head flange. Head detensioned with bolts in place. Manway cover on top of OTSG removed. Tyoon tubing attached to lower vents of RCS hot leg for RCS level indication.
Decay beat loop 2 in service for RCS cooling.
2.- Bus p-2 being upplied from 13.8 KV Bus A via breaker TrAAE2 and Bus o-2 beint supplied from 13.I8 KVTBus B via breaker HtBF2.
villedall tags clear. Maintenancew restoration of system will be less than two hours.
ECCS pupuction valves '(Dr-9A and DH-93) from containment sump closed and breakers racked out.
This will prevent the suction of air into the decay L
-3 heat loop during a level 5 actuation (recirculation mode) when there is no "ater in the sump.
- 5. Equipment Out of Service:
Emergency Diesel Geaerator 1 - maintenance
- 6. Breakers for containment spray and HPI pumps racked out.
4
IE Information Notice No. 80-20 Enclosure May 8, 1980 RECENTLY ISSUED IE INFORMATION NOTICES Information Subject Date Issued To Notice No.
Issued 80-20 Loss of Decay Heat Removal 5/8/80 All holders of power Capability at Davis-Besse reactor operation Unit 1 While in a Refueling Lic'enses or construc Mode tion permit 80-19 NIOSH Recall of Recircu-5/6/80 All holders of a power lating-Mode (Closed-Circuit) reactor OL, Research Self-Contained Breathing Reactor License, Fuel Apparatus (Rebreathers)
Cycle Facility License and Priority I Material License 80-18 Possible Weapons Smuggling 5/5/80 All power reactor Pouch facilities with an OL, fuel fabrication and processing facilities and Materials Priority I licensees (processors and distributors) 80-17 Potential Hazards Associated 5/5/80 All radiography With Interchangable Parts Licenses On Radiographic Equipment 80-16 Shaft Seal Packing 4/29/80 All power reactor Causes Binding In Main facilities in your Steam Swing Check And Region with an OL or CP Isolation Valves 80-15 Axial (Longitudinal) 4/21/80 All Light Water Reactor Oriented Cracking In Facilities holding power Piping reactor OLs or CPs 80-14 Safety Suggestions From 4/2/80 All power reactor Employees facilities with an OL or CP 80-13 General Electric Type SBM 4/2/80 All light water reactor Control Switches - Defective facilities holding power Cam Followers reactor OLs or.CPs 80-12 Instrument Failure Causes 3/31/80 All holders of power Opening of PORV and Block reactor OLs and CPs Valve 80-11 General Problems with ASCO 3/14/80 All holders of Reactor Valves in Nuclear Application OL, CP, fuel fabrica Including Fire Protection tion and processing Systems facilities