NRC Generic Letter 1980-42

NRC Generic Letter 1980-042: Transmittal of IE Bulletin 1980-012, Decay Heat Removal System Operability.

ML031350394
Person / Time
Issue date:	05/09/1980
From:	Grier B NRC Region 1
To:
References
BL-80-012 GL-80-042, NUDOCS 8005200124
Download: ML031350394 (11)

1--1 UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION I 8005200 124

631 PARK AVENUE

KING OF PRUSSIA, PENNSYLVANIA 1940 Milk, May 9, 1980

Docket Nos. 50-03

50-247 G&&gW - -

Consolidated Edison Company of New York, Inc.

ATTN: Mr. W. J. Cahill, Jr.

Vice President

4 Irving Place New York, New York 10003 Gentlemen:

The enclosed IE Bulletin No. 80-12, "Decay-Heat Removal System Operability,"

is forwarded to you for action. A written response is required. If you desire additional information regarding this matter, please contact this office.

Sincerely, Boyce H. Grier goA Director Enclosures:

1. IE Bulletin No. 80-12 with Attachment

2. List of Recently Issued IE Bulletins

CONTACT

W. H. Baunack

(215-337-5253)

cc w/encls:

L. 0. Brooks, Project Manager, IP Nuclear W. Monti, Manager - Nuclear Power Generation Department M. Shatkouski, Plant Manager J. M. Makepeace, Director, Technical Engineering W. D. Hamlin, Assistant to Resident Manager J. D. Block, Esquire, Executive Vice President - Administration Joyce P. Davis, Esquire

ENCLOSURE 1 UNITED STATES SSINS No.: 6820

NUCLEAR REGULATORY COMMISSION Accession No.:

OFFICE OF INSPECTION AND ENFORCEMENT 8005050053 WASHINGTON, D.C. 20555 IE Bulletin No. 80-12 Date: May 9, 1980 DECAY HEAT REMOVAL SYSTEM OPERABILITY

Introduction:

The intent of this Bulletin is to improve nuclear power plant safety by re- ducing 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 con- current 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 Triin 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 I

incident of April 19, 1980, DHR capability was restored prior to exceeding the specified RCS temperature 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 Attachment 1). Two factors identified as major contributors to the Davis-Besse event in the Information Notice are: (1) extensive mainten- ance 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 Attachment 1.

Enclosure 1 IE Bulletin No. 80-12 Date: May 9, 1980 2. Review your facility(ies) for all DHR degradation events experienced, especially for events similar to the Davis-Besse incident.

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.

Enclosure 1 IE Bulletin No. 80-12 Date: May 9, 1980 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 affirmation and submitted within the time specified above. Reports shall be submitted 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.

Attachment:

IE Information Notice No. 80-20

Attachment to IE b6.letin No. 80-12 ENCLOSURE 1 SSINS No.: 6870

UNITED STATES Accession No.:

NUCLEAR REGULATORY COMMISSION 8002280671 OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D. C. 20555 IE Information Notice 80-20

Date: May 8, 1980 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 Attachment 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 1P and IN; Emergency Diesel-Generator No. 1; 4.16 KY Essential Switchgear Bus Cl; and 13.8 KY 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 Protection System (RPS) and Safety Features Actuation System (SFAS) were being energized 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 l 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

Enclosure 1 IE Information Notice No. 80-20

Date: May 8, 1980 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 approxi- mately 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 mainten- ance 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

-

Enclosure 1 IE Information Notice No. 80-20

Date: May 8, 1980 1 and 3 was lost, all five levels of SFAS were actuated. As stated 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.

Licensees having questions regarding this matter should contact the director of the appropriate NRC Regional Office.

Attachment A

• Attachment A to Information Notice No. 80-20

• . Enclst A

DAVIS-BESS; EVENT OF APRIL 19, 1980

STATUS OF DAVIS-BESSE 1 PRIOR TO LOSS OF POWER TtO BUSSES E-2 AMD 7-2:

1.. Refuelng mode with RCS temperature at 9007 and level slihtly below vessel head flange. Send detensioned with bls in place. Hanw.ay cover on top of OTSG removed. Tygon tubing attached to lwer vents of ReS hot leg for RCS level indication. Decay heat loop 2 In ;ervice for RCS

coling.

2. A12 zou-nuclear I,=truent (NNI) power and Static Volta,e Regulator UAR

aupplied f=om 13.8 K7 Bus B via EBE3T. 13.8 F. Bus A energized but not cozaected. RPS a-d S7AS Chanzels 1 and 3 being supplied from A.L

3. Equp;ment Out of Service

. Scuvzia ?.aze Chza--I. 2 - Surveillance b. zargenzy Diesel Ga paator I - Maintena-ce.

c. Decay Erat Lccp 3.- Mnte=nnce.

4. lraaers for contai~ent spray and EPI puis racked cut.

SEQUM4CE OF EVEMS

0 pTm. EvN1E CAUSSE/ColEsnS

2:00 P.=; Less of power to Grotmd short on 13.8 KV buakarr EBF2 Busses Z-2 and F-2 Whfich caused breaker to open. This (non-essential 480- intarzuptad power to busses E-2 and 7-2 VAC) which were supplying all nou-nuclear Instrment (1tl) pcwer,. channes 1 and 3 of' the Reactor Protection System (ITS) and

.tba Safety Features ActuatLon Signal (S7AS),

the computer, and =ch of the control roon i -dicators.

2:00 p.m. SFAS Level 5 (recircu- Two out of four logic t-Ipped uspon less laticn. node) actua- of 3usses Z-2 a=d 7-2. Actuation caused tlou. zCCS P _ :cstin Val-,va3 from cwctaizreat sump to open and ECCS p=p suction valves froe Berated Water Storage Tank to close.

During valve travel times, gravity flow path existed from BWST to contalze=t sunp.

2:02 p.m. Decay Reat (low Operator turned off only operating DS

pressure safety in- pump to avoid spillage of RCS water to jection) flow secured cctainme=t via the rygon tubing for ICS

by operator level indication and open SG =anway.

2:33 p.z. Partial restoration of power Power to Bus E-2 anzd SAS chattels I and 3 restored along with oze channel of KNJI. This

-restoed zll anecr4sm _

Attachment to Ir tnformation Notice No. 80-20

TIM EVENT CAUSE/COK4ENTS

2:44 p.m. . Attempt to reestab- Started DE pump 1-2 then sthpped it when lish DE flow it was determined that air was in suction line. Pump secured to prevent da-ge.

334 p.m. Source Range Charmel 2 energized.

4:00 p.=. Restoration of Busses Busses restored sequentually as efforts to (480 VAC) F-2, F-21, progressed to isolate ground fault.

4:06 p.M. F-22, and F-23

4:25 p.m. DE flow restored DE pump 1-2 started after venting. RCS

temperature at 170 0F.DE flow bypassing

°°aF6 nmcore TC's being taken and maxi==

4:46 p.m. Containment sump Precautionary measure to assure containment.

P=up breakers s3p vatar fr= 3SWST rcsained i= qonntainze=.

- opened Incore TC's range from 16. to 164 P.

5:40 p.m. Computer returned to corse TC's range frm 358 to 160*F.

.. ervice.

6t24 p.mz. DE flow directed RCS cooldowm established at less Shan 25aF

thrcugh cooler per hour. RCS te elatu re at 150 F. Incare C'

- .: p* m. . -!. - range from 15. to or.o

.I:.; 9:S0 P. M

,,$@.

Pawer completely RCS temperature at app oMattely flSY.

-ratored

• • • )

t-_ w.

M &

sTAuS OF DMVUB-ESSE AFTER RECOVERY FROM LOSS OF POWER TO BUSSES E-2 AND F-2:

1 Refueling =de with RCS temperature at 1151F and level slightly below vessel beac flzane. Eead datensiazed with bolts in place. Hznvay cover on. top of OTSG

renove4. Tygon tubing attached to lawer vents of flCS hot leg for RCS level indicaiion. Decay beat loop 2 in sezvice for RCS cooling.

2. Bus E-2- being.-upp ied from 13.8 V Bus A via breaker EAAE2 and Bus F-2 being

~plied from 13.8 KV Bus B via breaker "BF2.

3. -lDecay heat loop filled, all tags clear. Maintenance work restricted so restoration of system will be less than two bours.

4. ECCS pup suction valves (DU-9A and DE-9B) from containment sump closed and breakers racked out. This will prevent the suction of air into the decay

Attachme~nt to IF Tnformatibn Notice No. 80'-20

..

i heat loop during a level 5 actuation (recirculation mode) whets there is no water in the.sumup.

5. Equipzent Out of Service:

6.

Emaergency Diesel Generator 1 - Maintenance Breakers for contaliment spray and BPI pumps racked out.

.K

b I

I

J.

6

.9

1 ENCLOSURE 2 IE Bulletin No. 80-12 Date: May 9, 1980 RECENTLY ISSUED IE BULLETINS

Bulletin Subject Date Issued Issued To No.

80-06 Engineered Safety 3/13/80 All Power Reactor Feature (ESF) Reset Facilities with an Controls Operating License (OL) (For Action)

All Power Reactor Facilities with a Construction Permit (CP) (For Information)79-03A Longitudinal Weld 4/4/80 All Power Reactor Defects in ASME Facilities with an SA-312, Type 304 OL or CP

Stainless Steel Pipe

80-07 BWR Jet Pump Assembly 4/4/80 BWR 3 & 4's with OL

Failure (For Action)

BWR's with CP (For Information)

80-08 Examination of Con- 4/7/80 All Power Reactor tainment Liner Facilities with Penetration Welds an OL or CP

80-09 Hydromotor Actuator 4/17/80 All Power Reactor Deficiencies Facilities with an OL or CP

80-10 Contamination of 5/6/80 All Power Reac- Nonradioactive tors with OL

System and Resulting (for action)

Potential for Unmoni- with CP (for tored, Uncontrolled information)

Release to Environment

80-11 Masonry Weld Design 5/8/80 All Power Reactor Facilities with an OL, Except Salem 2