Information Notice 1993-28, Failure to Consider Loss of DC Bus in the Emergency Core Cooling System Evaluation May Lead to Nonconservative Analysis

Failure to Consider Loss of DC Bus in the Emergency Core Cooling System Evaluation May Lead to Nonconservative Analysis

ML031080005
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	04/09/1993
From:	Griems B Office of Nuclear Reactor Regulation
To:
References
IN-93-028, NUDOCS 9304050032
Download: ML031080005 (9)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 April 9, 1993 NRC INFORMATION NOTICE 93-28: FAILURE TO CONSIDER LOSS OF DC BUS IN THE

EMERGENCY CORE COOLING SYSTEM EVALUATION MAY

LEAD TO NONCONSERVATIVE ANALYSIS

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees that a single failure of a 125-Vdc bus has been

identified as potentially the worst case single failure in the emergency core

cooling system (ECCS) evaluation for certain boiling water reactors (BWR-3 and

BWR-4). It is expected that recipients will review the information for

applicability to their facilities and consider actions, as appropriate, to

avoid similar problems. However, suggestions contained in this information

notice are not NRC requirements; therefore, no specific action or written

response is required.

Description of Circumstances

On July 30, 1992, the Nebraska Public Power District, licensee for the Cooper

Nuclear Station, notified the NRC-that the worst case single failure in the

ECCS had not been correctly identified in the analysis of the loss-of-coolant

accident (LOCA) in conjunction with a loss of offsite power. To satisfy

regulatory requirements, the most limiting single failure, which results in

the most severe calculated consequences, must be considered in performing the

LOCA analysis. The previous licensee analysis assumed that the worst case

single failure was a failure of the low pressure coolant injection (LPCI)

system injection valve in the ECCS train that is connected to one

recirculation loop, concurrent with a pipe break in the other recirculation

loop. In July 1992, the licensee recognized that failure of a 125-Vdc bus

that provides the control power for the LPCI injection valve serving the

unbroken recirculation loop is the worst single failure for this accident.

The licensee discovered the problem while performing a plant design basis

reconstitution.

Discussion

The low-pressure trains of the ECCS are shown conceptually in Figures 1 and 2.

Trains A and B each include a core spray pump and two residual heat removal

(RHR) pumps that function as LPCI pumps during a LOCA. The two RHR pumps in

q PD R I7cA-tc S3-o°8> 3° qi j-

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IN 93-28 April 9, 1993 each train share common piping and a common injection valve to the associated

recirculation loop. When offsite power is lost, two emergency diesel

generators (EDGs) supply power to the pump motors and two 250-V batteries

supply motive power to the valve motors. Two 125-V batteries provide control

power to the valve motors and the output circuit breakers for the EDGs.

In the previous ECCS analysis, as shown in Figure 1, the licensee assumed a

guillotine break in one recirculation loop and failure of the LPCI injection

valve for the other recirculation loop. If this event were to occur, coolant

from RHR pumps BI and B2 would not reach the reactor vessel because the

coolant would flow out through the broken recirculation loop, and RHR pumps Al

and A2 would not pump coolant to the reactor vessel because of the failure of

the LPCI injection valve to open. However, it was concluded that both core

spray pumps would pump coolant to the reactor vessel and the temperature of

fuel cladding would increase but remain less than the regulatory limit.

The licensee has now determined that the worst single failure would be the

failure of one of the two 125-Vdc buses. As shown in Figure 2, failure of the

125-Vdc bus for train A would prevent closure of the output breaker for EDG A.

Motors for RHR pumps Al and B2 and core spray pump A would not receive

emergency power and would not supply coolant to the reactor vessel. Likewise, the LPCI injection valve for recirculation loop A would fail to open and RHR

pump A2, which receives power from EDG B, would be unable to pump coolant to

the reactor vessel. If the break were in recirculation loop B, then RHR pump

B1 would pump coolant to the broken loop, leaving only core spray pump B to

pump coolant to the reactor vessel, which is insufficient to perform the

intended ECCS function. To avoid the possibility that the temperature of the

fuel cladding might exceed the regulatory limit, the licensee reduced the

reactor power level pending completion of modifications to correct the

problem. The licensee took this action based on analyses by the General

Electric Company.

Although the combined probability of occurrence of a guillotine rupture, loss

of offsite power, and failure of the 125-Vdc bus for the ECCS train serving

the Intact recirculation loop is very low, the licensee has modified the

control power for the LPCI injection valves and recirculation pump discharge

valves so that they are powered from the 250-Vdc buses, while leaving the

125-Vdc control power for the EDG output breakers unchanged.

Related Generic Communications

The General Electric Company submitted a report to the NRC as an attachment to

a letter dated November 1, 1978, addressing the effect of a failure of a

direct current power supply on BWR-3 and BWR-4 reactors. General Electric

concluded that peak cladding temperature would be greater than previously

anticipated for a small break LOCA, but less than the regulatory limit, and

that the peak clad temperature would be unaffected for large breaks. The

report indicated that at least two ECCS pumps would be available.

IN 93-28 April 9, 1993 The NRC sent letters to licensees on April 25, 1980, requesting that

addressees confirm the validity of the conclusions made by the General

Electric report. The letters asked that responses include lists of ECCS

equipment that would be available for breaks in the suction and discharge

piping connected to pumps in the recirculation loops. Based on the recent

analysis performed at Cooper, the conclusion in the General Electric report

that at least two ECCS pumps would remain available may not be true for all

scenarios involving direct current power supply failure.

The NRC recently issued Information Notice 93-11, 'Single Failure

Vulnerability of Engineered Safety Features Actuation Systems," to alert

licensees to a design deficiency identified at Millstone Nuclear Power

Station, Unit 2, that causes a spurious engineered safety feature actuation

when one train of DC electrical power is deenergized. In addition, in Generic

Letter 89-18, "Systems Interactions in Nuclear Power Plants,' the NRC

highlighted concerns regarding actuation system designs, including electrical

power system designs, that could cause adverse system interactions.

This information notice requires no specific action or written response. If

you have any question about the information in this notice, please contact one

of the technical contacts listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

6$ 0

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: David L. Skeen, NRR

(301) 504-1174 Elmo E. Collins, RIV

(817) 860-8291 Attachments: el ee(-&

2

1. Figure 1: tow Pressq4& ECCS Trains...

Figure 2: Break in iecirculation Loop...

2. List of Recently Issued Information Notices

Attachment 1 IN 93-28 April 9, 1993 r--0U011110114 .11 I m__

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Figure 1: Low pressure ECCS trains showing break In

recirculation loop B with loss d offsite power and

failure of the 'A' LPCI Injection value.

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Figure 2: Break in recirculation loop B with loss of offsite power

and one 125-Vdc bus.

Attachment 2 IN 93-28 April 9, 1993 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

93-27 Level Instrumentation 04/08/93 All holders of OLs or CPs

Inaccuracies Observed for nuclear power reactors.

during Normal Plant

Depressurization

93-26 Grease Solidification 04/07/93 All holders of OLs or CPs

Causes Molded Case for nuclear power reactors.

Circuit Breaker

Failure to Close

93-25 Electrical Penetration 04/01/93 All holders of OLs or CPs

Assembly Degradation for nuclear power reactors.

93-24 Distribution of 03/31/93 All holders of operator and

Revision 7 of NUREG-1021, senior operator licenses at

"Operator Licensing nuclear power reactors.

Examiner Standards'

93-23 Weschler Instruments 03/31/93 All holders of OLs or CPs

Model 252 Switchboard for nuclear power reactors.

Meters

93-22 Tripping of Klockner- 03/26/93 All holders of OLs or CPs

Moeller Molded-Case for nuclear power reactors.

Circuit Breakers due to

Support Level Failure

93-21 Summary of NRC Staff 03/25/93 All holders of OLs or CPs

Observations Compiled for light water nuclear

during Engineering Audits power reactors.

or Inspections of Licen- see Erosion/Corrosion

Programs

93-20 Thermal Fatigue Cracking 03/24/93 All holders of OLs or CPs

of Feedwater Piping to for PWRs supplied by

Steam Generators Westinghouse or Combustion

Engineering.

93-19 Slab Hopper Bulging 03/17/92 All nuclear fuel cycle

licensees.

OL - Operating License

CP - Construction Permit

IN 93-28 April 9, 1993 The NRC sent letters to licensees on April 25, 1980, requesting that

addressees confirm the validity of the conclusions made by the General

Electric report. The letters asked that responses include lists of ECCS

equipment that would be available for breaks in the suction and discharge

piping connected to pumps in the recirculation loops. Based on the recent

analysis performed at Cooper, the conclusion in the General Electric report

that at least two ECCS pumps would remain available may not be true for all

scenarios involving direct current power supply failure.

The NRC recently issued Information Notice 93-11, "Single Failure

Vulnerability of Engineered Safety Features Actuation Systems," to alert

licensees to a design deficiency identified at Millstone Nuclear Power

Station, Unit 2, that causes a spurious engineered safety feature actuation

when one train of DC electrical power is deenergized. In addition, in Generic

Letter 89-18, "Systems Interactions in Nuclear Power Plants," the NRC

highlighted concerns regarding actuation system designs, including electrical

power system designs, that could cause adverse system interactions.

This information notice requires no specific action or written response. If

you have any question about the information in this notice, please contact one

of the technical contacts listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Original igned by

Brian K.Grimei

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: David L. Skeen, NRR

(301) 504-1174 Elmo E. Collins, RIV

(817) 860-8291 Attachments:

1. Figure 1: Low Pressure ECCS Trains...

Figure 2: Break in Recirculation Loop...

2. List of Recently Issued Information Notices

• SEE PREVIOUS CONCURRENCES

OEAB SC:DRP:RIV RPB:ADM SC:OEAB PM:PD41 D:DRP:RIV

RWWoodruff* EECollins* JDMain* RLDennig* RBBevan* ABBeach*

02/04/93 02/04/93 11/17/92 02/04/93 01/26/93 02/04/93 C:EELB C:SRXB C:OEAB OGCB C:OGCB

CHBerlinger* RCJones* AEChaffee* JBirmingham* GHMarcus* BKGr

02/10/93 02/19/93 02/22/93 03/01/93 03/01/93 04/7 /93 Document Name: 93-28.IN

al

IN 93-XX

April xx, 1993 The NRC sent letters to licensees on April 25, 1980, requesting that

addressees confirm the validity of the conclusions made by the General

Electric report. The letters asked that responses include lists of ECCS

equipment that would be available for breaks in the suction and discharge

piping connected to pumps in the recirculation loops. Based on the recent

analysis performed at Cooper, the conclusion in the General Electric report

that at least two ECCS pumps would remain available may not be true for all

scenarios involving direct current power supply failure.

The NRC recently issued Information Notice 93-11, 'Single Failure

Vulnerability of Engineered Safety Features Actuation Systems," to alert

licensees to a design deficiency identified at Millstone Nuclear Power

Station, Unit 2, that causes a spurious engineered safety feature actuation

when one train of DC electrical power is deenergized. In addition, in Generic

Letter 89-18, "Systems Interactions in Nuclear Power Plantsithe NRC

highlighted concerns regarding actuation system designs, incTuding electrical

power system designs, that could cause adverse system interactions.

This information notice requires no specific action or written response. If

you have any question about the information in this notice, please contact one

of the technical contacts listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: David L. Skeen, NRR

(301) 504-1174 Elmo E. Collins, RIV

(817) 860-8291 Attachment: Figure 1: Low Pressure ECCS Trains...

Figure 2: Break in Recirculation Loop...

List of Recently Issued Information Notices

• SEE PREVIOUS CONCURRENCES

OEAB SC:DRP:RIV RPB:ADM SC:OEAB PM:PD41 D:DRP:RIV

RWWoodruff* EECollins* JDMain* RLDennig* RBBevan* ABBeach*

02/04/93 02/04/93 11/17/92 02/04/93 01/26/93 02/04/93 C:EELB C:SRXB C:OEAB OGCB C:OGCB D:DORS

CHBerlinger* RCJones* AEChaffee* JBirmingham* GHMarcus* BKGrimes

02/10/93 02/19/93 02/22/93 03/01/93 03/01/93 / /93

N 3-XX

April xx, 1993 The NRC sent letters to licensees on April 25, 1980, requesting that

addressees confirm the validity of the conclusions made by the General

Electric report. The letters asked that responses include lists of ECCS

equipment that would be available for breaks in the suction-and discharge

piping connected to pumps in the recirculation loops. Based on the recent

analysis performed at Cooper, the conclusion in the General Electric report

that at least two ECCS pumps would remain available may not be true for all

scenarios involving direct current power supply failure.

This information notice requires no specific action or written response. If

you have any question about the information in this notice, please contact one

of the technical contacts listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Roger W. Wo.druffL NRR

(301) 504 t152j) D -- . -, 4--

Elmo E. Collins, RIV

(817) 860-8291 Ottachmetg: ur 1: Low Pressure ECCS Trains...

,J gure 2: Break in Recirculation Loop...

of Recently Issued Information Notices

• SEE PREVIOUS CONCURRENCES

OEAB SC:DRP:RIV RPB:ADM SC:OEAB PM:PD41 D:DRP:RIV

RWWoodruff* EECollins* JDMain* RLDennig* RBBevan* ABBeach*

02/04/93 02/04/93 11/17/92 02/04/93 01/26/93 02/04/93 C:EELB C:SRXB C:OEAB OGCB C:OGCB D:DORS

CHBerlinger* RCJones* AEChaffee* JBirmingham* GHMarcus* BKGrimesOLL

02/10/93 02/19/93 02/22/93 03/01/93 03/01/93 / /93V

Document Name: G:\DLS\COOPER.IN

\h. N.

IN 93-XX

January xx, 1993 This information notice requires no specific action or written response. If

you have any question about the information in this notice, please contact one

of the technical contacts listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: Roger W. Woodruff, NRR

(301) 504-1152 Elmo E. Collins, RIV

(817) 860-8291 Attachment: List of Recently Issued Information Notices

Ar; r

PAS

OEABI i . SC:DRP:R IV RPB:ADM PM: PD41 D:DRP:RIV

RWWoopruff EECollin!S JDMain *- RBBevan*- ABBeach

-/c49 3~ a g2/IA 193 / /a* 2/ A./93 t* :SRXB C:OEABgA,, C:OGCB D:DORS

RCJones AEChaffee GHMarcus BKGrimes

/93 k/011/93 a-ld),193 / /93 / /93 Document Name: G:\DLS\COOPER.IN

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