ML20012B899

LER 89-023-01:on 891220,determined That Pipe Rupture in nonsafety-related Svc Water Subsystem Could Result in Rapid Draining of Subsystems That Serve Auxiliary Bldg.Task Force Formed to Determine Corrective actions.W/900312 Ltr

ML20012B899
Person / Time
Site:	Calvert Cliffs
Issue date:	03/12/1990
From:	Denton R, Love G BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-023, NUDOCS 9003190137
Download: ML20012B899 (11)

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., BALTIMORE Ru' ' '

OAS AND  ;

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• P.O. BOX 1475

• BALTIMORE. MARYLAND 21203 1475' f ii J H

CALVERT Cliff 8 NBCLEAR POWER PLANT DEPARTRIENT '- '

, CALVERT CUFF 8 NUCLEAR POWER PLANT

s . LUSOY, MARYAND 20657 .r t

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-I March 12, 1990 ,

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', i U. S Nuclear Regulatory Commission:

. Docket Nos. 50 317 & 50 318'-

License Nos. DPR 53 & 69 J Document Control De.sk

' 1 White Flint North MailiStop P1-317- .

11555 Rockville Pike. *

- Rockville, MD 20850 -

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Dear' Sirs:

The' attached supplemental LER 89-23.. Revision 1,'is being sent to you'as-

required under 10 CFR 50.73. guidelines.

Should you have any questions regarding this report, we would be pleased to

. discuss them with'you.' 3

'Very truly yours,

, n- -

i RL.E.-Denton Manager

~i GAL:1rr-cc: . William T.. Russell .

Director, Office of Management.Information -l

.and Program Control j G. C. Creel'

. Messrs:

C. H. Cruse ex J..'R. Lemons L. B. Russell .

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Asernact to . e.oe t.. - . an c. . an no This supplement to LER 89-023 is submitted to clearly identify our position as previously stated in the Calvert Cliffs Nuclear Power Plant's Safety Evaluation Report issued August 28, 1972, and to reclassify the reportability to "other" and the LER as 1 a voluntary report.

At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on December 20, 1989 it was postulated that a reportable condition may I have existed as a result of a plant configuration that could potentially result in <

l- unavailability of both safety-related Service Water (SRW) Subsystems. At the time of determination, Unit 1 was in cold shutdown with the Reactor Coolant System (RCS) partially filled. at atmospheric pressure, and 114 degrees F. The Unit 2 reactor was defueled, with the reactor vessel partially drained, the vessel head detensioned, and the RCS at atmospheric pressure and ambient temperature.

It was postulated that a pipe rupture in the non safety-related SRW Subsystem thatl serves the Turbine Building could result in rapid draining of both of the independent, safety-related SRW Subsystems that serve the Auxiliary Building. The loss of both Auxiliary Building SRW trains could subsequently result in unavailability of the Emergency Diesel Cenerators. The reported condition does not describe an actual event; therefore, it was not contributed to by any actual compcnent or system failures. Based on a review of the NRC's Safety Evaluation Report for Calvert Cliffs Nuclear Power Plant (CCNPP), it is clear that we were only ar.alyzed for a LOCA concurrent with a loss of off-site power. Therefore, it is clear our SRW System was not designed to cope with a seismic event and a simultaneous loss of off site power.

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rewr n m.. w , mea ,manwim I. DESCRIPTION OF EVENT -i At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on December 20, 1989 it was postulated that a reportable condition may have existed at Calvert Cliffs. The condition was the >

result of a plant configuration that could have potentially resulted in tha unavailability of both independent, safety related Service Water (SRW)-

Subsystems. This determination wrs made during a routine reportability +

review for a Non Conformance Report (NCR). . The NCR described a condition whereby a postulated pipe rupture in >the non safety-relat_ed SRW Subsystem .

that serves the Turbine Building could result in rapid draining of. both of the independent, safety related SRW Subsystems ~ that serve the Auxiliary Building. The- loss of both - Auxiliary Building . SRW Subsystems could subsequently. result in unavailability of the Emergency Diesel Generators (EDGs).  :

L At the time.of determinatica, Unit I was in cold shutdown with the Reactor Coolant System (RCS) partially filled, at atmospheric pressure, and-114 degrees F. The Unit 2 resotor was defueled, with the reacter vessel partially drained,.the vessel head detensioned, and the RCS at atmospheric pressure and ambient temperature.

The- purpose .of the SRW System is to remove heat from the main

. turbine-generator plant components, containment cooling unito, spent fuel pool heat exchangers, and various EDG heat exchangers, and to transfer that heat to the Saltwater System. Although the SRW piping configuration differs slightly between Unit i and Unit 2, each unit is basically '

comprised of two independent, safety-related SRW Subsystems. in the Auxiliary . Building which operate in parallel with a single, non-safety-

, related SRW Subsystem-in the Turbine Building.

Both - Auxiliary Building - SRW Subsystems and the Turbine Building SRW Subsystem are needed during normal plant operations. For Unit 2, the two Auxiliary Building SRW Subsystems are connected to the-- Turbine Building SRW Subsystem by a common, non safety-related connection from the SRW discharge header where the SRW- System enters the Turbine Building. For ,

Unit 1, .the two Auxiliary Building SRW Subsystems are onnected to the Turbir.e Building SRW Subsystem by a common, non-safety-related pipe located where the SRW System exits the Turbine Building and connects to the SRW suction header. As a result of these common piping connections, the non-safety-related Terbine Building SRW Subsystem essentially cross-connects with the two safety-related Auxiliary Buildieg SRU Subsystems.

1 The ability to isolate the Turbine Building SRW Subsystem from the Auxiliary Building SRW Subsystems is provided by dual, air operated isolation valves on the discharge header piping of each Auxiliary Building SRW Subsystem, and by check valves in the suction header piping of each Auxiliary Building SRW Subsystem. The isolation valves are located in the l

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- safety-related Auxiliary Subsystem piping prior to connection with the Turbine Building Subsystem piping, and the check valves are located in the m

safety related portions of the Auxiliary Building SRW Subsystem suction

. header piping - The Turbine Building isolation valves can be operated from the main Control Room, and close automatically following receipt of a Safety Injection Actuation Signal (SIAS) or loss of instrument air. Other than SIAS or loss of instrument air, there are no other automatic closure l functions associar.ed with the' isolation valves.

Calculations have been - performed assuming a worst-case, double ended guillotine pipe break in 'the non safety-related Turbine Building SRW piping. It was also assumed that the break would occur'in the Unit 2 SRW piping . configuration. The piping configuration for~ Unit 2 is much less

= conservative .than Unit 1 configuration because its' cross-connection.

occurs just downstream of the non critical service water valves. The Unit 1 cross-connection is . downstream of all Turbine Building loads., The calculation results indicate that under the previously mentioned -*

conditions, breakflow could empty the SRW System in less. time than is '

required , for the isolation valves to close following receipt.of.a SIAS.

The calculations also indicate that the SRW System could be drained before an operator could act to isolate the break under non SIAS conditions. It should be noted that assumption of a double-ended guillotine type break is more conservative than is required under our licensing and design basis for a moderate energy break in a line that is designed as Seismic Category II, and was constructed to ANSI B31.1. However, informal calculations ,

indicate that even a moderately sized pipe break would result in a rapid loss of SRW' inventory.

Calvert Cliffs Nuclear Power' Plant (CCNPP) is only analyzed for a loss of. )

coolant accident (IDCA) concurrent with a loss of off-site power, and not

! for a seismic event concurrent with a loss of off-site power. Therefore, the condition described in this report is being reported under "other" as L a vcluntary LER.

, The following identifies CCNPP current design assumptions, active and L passive failures as described in our Updated Final Safety Analysis Report L (UFSAR), Section 9.5.

Current Design Assumotions I

L Single Failure Analvj h 4 Needed 4 Needed Comoonent For Normal OPS For LOCI SRW Heat Exchangers Note (a) 1 SRU Pumps 2 1 EDCs 0 1 estC pow 3eea *3 .

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s a) Two are needed. However, the two subsystems may be cross connected and one heat exchanger may be utilized to remove the full heat load.

Active Failures System Comoonent Tvoe of Failure Service Water Turbine Building. Fails to close on SIAS isolation valves. ,

Consecuences

.Valvee are actuated by a redundant channel and would shut, isolating service water as required. ,

System Component Tyne of Failure Service Water 12.EDG Supply / Fails to seek-header return CVs with pressure Consecuences Diesel generator 12 does not receive any cooling water. This could result in cross-connecting one subsystem of each unit _ and possibly draining one subsystem by over-flowing in the other unit's head tank. However, each unit would still have a subsystem in operation and this is . sufficient to remove all necessary heat. Diesel generators 11 and 21 are cooled - and provide sufficient electrical power, System- Comoonent Tvoc of Failure Service Water Turbine building Fails to close'under return check valve. reverse flow.

Consecuences Since in all cases two check valves are provided in-series,'the second valve would close_providing isolation.

Note: As shown above sufficient numbers of all other active components are supplied to provide sufficient redundancy for all modes of operation.

Passive Failure Durine Containment Sumo Recirculation System Location of Ruoture Service Water 12 EDG supply / return manual cross connect valves.

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Conseauengg.g One . subsystem from each unit ~ would be drained and rendered inoperable.' i However, one subsystem - in each unit would continue to operate. This is ,

' adequate to provide the necessary cooling'for each-unit. No single rupture in any location could cause the loss of both subsystems of a unit as two, normally closed valves are provided where two subsystems are tied together.

Flooding Due to a Passive Failure  ;

Indication in ,

Structure Flooded Control Room System Ruotured i Service Water Room High level alarm Saltwater in the room with normal service water head tank level.

Consecuences Saltwater to the Service Water Room would be stopped by closing remote manual valves from the Control Room. The containment coolers would be shut down and_ heat removed fro.n the containment would be via the spray system.

Service water would continue to operate until the service water temperature reached' 120 F, which would occur approximately 21 minutes after loss of .

saltwater, based on an- initial service water. temperature of 95 F. This is  !

considered. to be sufficient time,' to determine which subsystem has ruptured  ;

and to re-establish saltwater flow in the other subsystem.

1 When both units are in operation, cool ' service water would be provided to diesel generator No. 12 by the other unit's Service Water System.. Saltwater for this other unit is functioning normally. This- is accomplished as follows:

a. Diesel generator No. 12 automatically provides power to the accident unit-channel ZB for Unit 1 (ZA for Unit 2),

b. Valves on the-discharge-of service water pump No. 13 (23) are normally open to Service Water Subsystem ZB (ZA).

c. The circuit breaker is remotely closed to provide power to service water pump No. 13(23) from channel ZA (ZB).

d. The service water pumps on the unit which has the ruptured Saltwater System are shutdown,

e. The pressure seeking valves automatically supply diesel generator No.12 cooling water from the other unit.

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Structure' Flooded Control Room System Ruotured q Service Water Room- High level. alarm Saltwater i in the room with i normal service i water head tank i level, i Conseaugngga ,

.One subsystem from each un'it would be drained. However, the other subsystem would continue to operate and'is sufficient to provide all necessary service ~!

' water. The entire-contents of one Service Water-System would not-flood out  ;

the service water pumps and motors.

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- Based on the above and a review of NRC's Safety Evaluation Report for CCNPP, it is clear that we were only analyzed for a 1DCA concurrent with a loss of off-site '

l' power. Therefore, it-is clear our Service Water System was not designed to cope- S

-with a seismic event with a simultaneous loss of off site power.

Report' ability as an event or condition prohibited by Technical Specifications (T.S.) is-related to T.S._3.7.4.1, which requires that "at least two independent service water loops shall be OPERABLE in modes 1, 2, 3, and 4" is.not an' issue as=  ;

described in LER 89 023, Revision O. It is determined 'at - this time that the current . design and configuration- of the . SRW System meets. the - intent of this ~

Technical Specification and the original licensing' and design. bases.. The original licensing and design bases are the same as the current plant: license and d

design bases.

The condition described in this. report is'not being considered for reportability as a condition that-is outside the plant design basis for the SRW System and'the

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EDGs. The SRW design basis is described in Section 9.5.2.2 of the. Updated Final Safety Analysis Report (UFSAR), and states that the SRW System "has been divided into two subsystems in the Auxiliary Building.to meet single failure criteria."

It also states that "during normal operation both [SRW) subsystems are . . . .;

independent to the degree necessary to assure the safe operation and shutdown of the plant assuming a single failure."

The system description for the EDGs is found in UFSAR.Section 8.4.1.2, and states that "the emergency diesel generators and their auxiliaries are designed to w ithstand Seismic Category 1 accelerations and are installed in Category 1 structures." The SRW System directly supplies cooling water to the EDGs and is considered to be auxiliary equipment to the EDCs. UFSAR Section 9.5.2.2 and 8.4.1.2 meet the intent of the original design basis, which is the same as our current design basis.

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ol0 0l 7 0F 1l0 rent u n . m .me w sca mnswim The assumptions assumed in LER 89 023, Revision 0 are more conservative than the current licensed plant design bases. Investigations perforn.ed in late 1989 previously assumed a double end guillotine pipe rupture in the non safety-related )

portion of the SRW System as the event initiator. We recognize a seismically ,

inducou pipe rupture and concurrent loss of off site power as a possible event J scenario, however, we were not required to analyze this scenario. However, an analysis of the postulated scenario and determination if a significant safety l concern exists is provided in Section III, Analysis of Event. 1 4  :

Calvert Cliffs Nuclear Power Plant Units 1 & 2 Safety Evaluation, issued August j 28, 1972', Section 3.2.5 stat'es,"in part "The Auxiliary Systems include the Chemical and Volume Control System, Shutdown Cooling System, Component Cooling Water System,, Service Water System, Saltwater System....

The; Service Water and Saltwater Systems' provide cooling required' for vital plant safety features. These systems were revised during our review to provide greater separation and redundancy so that they could sustain single failure of active or passive components without loss of the required cooling capability. l The design bases, functions, and descriptions of the Calvert Cliffs Auxiliary Systems are substantially the same as for other plants. that have been recently reviewed and approved for operating licenses. On the basis of our - comparison of these systems with those of other approved plants and our evaluation of the adequacy of each system we concluded that the Calvert Cliffs Auxiliary Systems are acceptable."~ 1 1

1 II. CAUSE OF EVENT N/A l 1

-III. ANAINSIS OF EVENT -

The postulated event was discovered during a routine reportability review for Non-Conformance Report (NCR). NCR 8391 stated a concern that "a rupture, without l-a SIAS (turbine building isolation valves do not shut), occurring in the Turbine Building will cause a loss of both subsystems. In the event of a loss of l off-site power this could render both EDGs inoperable." l An analysis of the problem and determination if a significant safety concern exists is provided below. I 1

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Information obtained from our Probabilistic Risk Assessment (PRA) Unit for East .I Coast earthquake risks put the probability for low energy earthquakes at 1.1E 7 1 per hour per event and 1E 9 per hour per event for potentially damaging earthquakes.

Even if we postulate a damaging earthquake, a catastrophic failure of the  ;

non safety-related (NSR) portion of the Service Water System is unlikely because l the Turbine Building is a Seismic Class II structure. It has a working stress design for 0.08 g horizontally and 0.053 g vertically (OBE accelerations). While l a conservative analysis was not performed , the building is relatively stiff and would not collapse'under Safe Shutdown Earthquake (SSE) conditions.

Inspections of industrial mill buildings and power plants have been performad-after earthquakes much stronger than our SSE. Except for those on soft soil or. ,

4 of very unusual configurations, those buildings performed well. - In addition, there were very few instances of piping damage. These points have been nade many

' times - with' the NRC and ACRS by Seismic Qualification Utility Group '(SQUG) consultants. .

The few piping fhilures noted by SQUG consultants were caused by: ,

- Unanchored equipment

- Severe building displacement / relative motion with little piping flexibility (such as buried pipes entering buildings, closely spaced rigid supports at' expansion joints).

Steel piping is inherently rugged. '"his is borne out by the testing which lead to ASME Code Case _N-411. and by the recent reports suggesting 'far less conservative design for small bore pipe.

The NRC has placed a relatively low priority-on seismic qualification (USI A 46).

Their position is that seismic is not a major contributor to nuclear risk.

Walkdowns by expert teams on five of the oldest plants resulted in few

. corrections. Some of these plants weren't even designed for earthquakes.

l It is postulated that a seismic event could occur following a LOCA. Emergency Operating Procedure E0P-5-(Loss of Coolant) directs operators during recovery to restart Turbine Building service water and~ restart equipment such as instrument -

air compressors. A calculation was performed and showed that a seismic event following a LOCA is an extremely low probability event scenario - on the order of IE-8 events in any 30 day period.

Anothe' failure mechanism that needs to be considered is passive failure. NUREG g CR.4407 (Pipe Break Frequency Estimates for Nuclear Power Plants) puts passive

pipe failure for balance of plant systems at 4.4E-8 failures per hour per event.

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I Our UFSAR only discusses passive failures during recirculation following a LOCA.

By comparing passive failure risk to a potentially damaging seismic risk, it is ,

obvious that the passive failure is the higher risk - even though Sth are small.

Our Abnormal Operating Procedures (AOP) recognize a pipe ruptwe as a possible )

event. AOP 78 (loss of Service Water) Sections IV and V (Rupture of a subsystem) I provides direction to operators in the event of a loss inventory in the Service  ;

Water System. There is explicit direction given to isolate the Turbine Building 1 a so that a rupture in one subsystem will not drain the other subsystem".

The NRC also recognized that plants of our vintage were not designed to withstand I a loss of their safety related (SR) Service Water System during non LOCA events.

Their evaluation of our May 20, 1980 loss of service water accident concluded the loss of service water event at Calvert Cliffs did not result in damage to any j plant equipment either safety c,r non safety related, and taken by itself does nog )

represent a cause for concern. The significance of nhe event lies in the fact i that it involved two fundamental aspects consitered in the design of .

safety related systems: J

1. Interaction between safety and non safety related systems and components; .

2. Common caused failure of redundant safety systems.

The review of the event by the Office for Analysis and Evaluation of Operational Data (AOD) revealed no immediate safety concerns; however, there is a need to reevaluate the isolation provisions at the interface between the safety at d non safety related portions of the Service Water System at Calvert Cliffs as weil as generically.

The primary concern with a loss of service water during non 14CA events is loss of cooling for the EDGs. This scenario would render EDGs inoperable and may place us in a station blackout. Calvert Cliffs is currently able to maintain the plant in a safe shutdown condition for four hours with no AC power. Station Blackout Procedure (EOP-7) provides direction to operators on how to restore both i the Service Water System and AC System to operation.

Given the low probability of a damaging earthquake and the small likelihood that it will cause a catastrophic failure of the Service Water System, we can cor.clude that there is no exigent need to take immediate action to modify the system due to an earthquake risk.

There is sufficient operator guidance to cope with a postulated loss of service water with a simultaneous loss of off site power. As an additional measure, Operations has increased the frequency of leak rate monitoring of the Service Water System, ggen. m.

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olo 1l0 or 1 l0 vert u . n.w anc s mmw nn IV. 00RRECTIVR ACTIONS TAKEN A task force has been assembled to determine appropriate long term corrective actions. In the interim, prior to startup, compensatory actions will be established, although the described is not part of our current licensing and design bases. The following compensatory actions will be established prior to startup.

- Change Alarm Manual to include immediate isolation of Turbine Building header on large rupture indications.

. Inform operators of the status of this issue prior to Unit 1 startup, i

V. ADDITIONAL INfDENATION N/A l l

VI. IDENTIFICATION.0F.00KPONENTS REFERRED IV IN THE LER ,

IEEE 803 IEEE 805  ;

Componenr/Systeu Component ID Code System ID Code Auxiliary Building NF Auxiliary Feedwater System BA Containment Cholers BK Control Room NA l Emergency Diesel Generator EK p

Isolation Valve ISV l

Reactor Coolant System AB

! Reactor Vessel RCT l Safety Injection System JE/BQ/BP L Saltwater System BS l Service Water System BI Spent Fuel Pool Cooling System DA Turbine Building NM Turbine Generator T- ,

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