ML18037A408

Provides Response to NRC Request for Addl Info Re LPCI Operability W/Rhrs Aligned for Shutdown Cooling

ML18037A408
Person / Time
Site:	Browns Ferry
Issue date:	08/12/1993
From:	Baron R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M85255, TAC-M85256, TAC-M85257, NUDOCS 9308170411
Download: ML18037A408 (11)

Text

ACCEI ERAT DOCUMENT DIST UTION SYSTEM

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REGULAT INFORMATION DISTRIBUTION STEM (RIDS) s)

ACCESSION NBR:9308170411 DOC.DATE: 93/08/12 NOTARIZED:

NO DOCKET FACIL:50;259 Browns Ferry Nuclear Power Station, Unit 1, Tennessee 05000259 50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH.NAME AUTHOR AFFILIATION BARON,R.R.

Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Provides response to NRC request for addi info re LPCI operability w/RHRS aligned for shutdown cooling.

DISTRIBUTION CODE:

D030D COPIES RECEIVED:LTR i ENCL U SIZE:

TITLE: TVA Facilities Routine Correspondence NOTES:

RECIPIENT ID CODE/NAME COPIES LTTR ENCL RECIPIENT ID CODE/NAME COPIES LTTR ENCL INTERNAL: ACRS OC LFDCB EXTERNAL: NRC PDR 01 6

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1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED:

LTTR 13 ENCL 11

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Tennessee Valley Authority, Post Olfice Box 2000, Decatur, Alabama 35609 AUG 12 1993 U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C.

20555 Gentlemen:

In the Matter Of Tennessee Valley Authority Docket Nos.

50-259 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)

NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING LOW PRESSURE COOLANT INFECTION (LPCZ)

OPERABILITY WITH THE RESIDUAL HEAT REMOVAL SYSTEM (RHRS)

AL1GNED FOR SHUTDOWN COOLING (SDC)(TAC NOS ~ M85255i M85256f and M85257)

Reference:

1)

Letter from NRC to TVA dated Zune 9,

1993, Request for Additional Information Proposed Technical Specification Amendment Regarding Low Pressure Coolant Injection Operability With Residual Heat Removal Aligned for Shutdown Cooling (TAC Nos.

M85255, M85256, and M85257) 2)

Letter from TVA to NRC dated December 23,

1992, Technical Specification (TS)

No.

328 L'ow Pressure Coolant Injection (LPCI) Operability When Residual Heat Removal System (RHRS) is Aligned to Shutdown Cooling (SDC)

Mode Units 1, 2, and 3

The enclosure provides TVA's response to NRC's request for additional information made by Reference 1 concerning TS-328 (Reference 2).

The proposed technical specification change in TS-328 would allow LPCI to be considered operable when the RHRS is aligned in the SDC mode with the reactor shut down.

Reference 1 requested TVA to describe the limiting accident for each operating mode affected by the proposed technical specifications and to describe the analysis of the postulated sequence of events.

9308i704ii;9308i2 PDR ADOCK'5000259 P

PDR

2 U.S. Nuclear Regulatory Commission AUG i2 893 If you have any questions, please contact me at (205) 729-4828.,

Since el R

R. Baron anager Nuclear Assurance and Licensing Enclosure cc (Enclosure):

Mr. R. V. Crlenjak, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident. Inspector Browns Ferry Nuclear Plant Route 12, Box 637

Athens, Alabama 35611 Mr. Thierry M. Ross, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

ENCLOSURE Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN)

LOW PRESSURE COOLANT INJECTION OPERABILITY RE UIREMENTS FOR RESIDUAL HEAT REMOVAL SYSTEM OPERATING IN SHUTDOWN COOLING MODE The proposed technical specification (TS) change allows the low pressure coolant injection (LPCI) mode of the Residual Heat Removal System (RHRS) to be considered operable when capable of manual realignment from the shutdown cooling (SDC) mode. It only applies when the plant is in hot shutdown condition at less than 105 psig or when the plant is in the cold shutdown condition.

With pressure

< 105 psig in hot shutdown, both Core Spray System (CSS) loops would be required to be operable.

In Cold Shutdown (atmospheric pressure) one CSS loop with one 'pump would be required to be operable.

The limiting accident for each operating mode affected by the proposed technical specifications would be bounded by BFN loss of coolant accident (LOCA) analyses.

The limiting LOCA analysis is a double-ended break of a recirculation discharge line with a single failure of the LPCI injection valve in the unbroken loop.

This analysis assumes a loss of normal AC power concurrent with the line break, the reactor is operating at 102% of rated power, and the reactor water level is at the scram level when the break occurs.

The postulated single failure of the LPCI injection valve in the unbroken loop results in effective core cooling being provided by the CSS only for the limiting event.

The Emergency Core Cooling Systems (ECCS) performance analysis for this event results in a Peak Cladding Temperature (PCT) of 1886'F, which is the maximum PCT calculated for any combination of recirculation line break and postulated single failure.

(Note the design PCT criterion is 2200 F.)

Since the proposed TS change is only applicable when the reactor is shutdown with reactor pressure less than 105 psig, the reactor heat. load is considerably less than that at 100% power with normal operating temperature and pressure.

The BFN licensing basis does not assume a double-ended guillotine break at atmospheric pressure.

With the proposed TS change, the worst single failure would be failure of one CSS injection valve which would result in only one loop of core spray available for automatic actuation for core cooling (Note that LPCI would still be available for manual injection).

The consequences of any postulated pipe failure during these shutdown conditions would be

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considerably less severe than the Design Basis Event (DBE) described above since there is not enough latent and decay heat.

after shutdown to cause fuel damage with Core Spray available.

Piping failures are considered to have an extremely low probability of occurrence due to the low pressure and margins inherent in reactor system piping design.

NRC ITEM 1 Describe the limiting accident for each operating mode affected by the proposed technical specifications.

TVA RESPONSE There are two operating modes affected by the proposed technical specifications:

1) hot shutdown condition with reactor pressure less than 105 psig, and 2) cold shutdown condition.

For the first case, the worst case accident is considered a

recirculation suction line break with a single failure of one CSS injection valve.

For the second

case, the cold shutdown condition, the most severe accident is considered to be an inadvertent drain down of the reactor vessel due to a break in SDC piping or a leak in the reactor coolant pressure boundary due to maintenance or valve mispositioning.

NRC ITEM 2 For the limiting accidents, describe the analysis of the postulated sequence of events, including time to core uncovery if mitigating actions are not taken.

Discuss the assumptions made in this evaluation, including a summary of equipment assumed to

operate, required operator actions, and a description of the worst single failure of equipment.

The description should address operations in accordance with both the existing technical specifications and the proposed changes.

TVA RESPONSE For the postulated accident in hot shutdown (see TVA response to Item 1), the unit is assumed to be at 105 psig.

One loop of RHR is assumed to be in the SDC mode.

The other loop is assumed to be unavailable.

No credit needs to be taken for manual

realignment of LPCI although immediate operator actions would be to realign.

A pipe break in the recirculation suction piping is assumed.

One of several possible single failures occurs which renders one CSS injection valve inoperable resulting in one CSS loop injecting into the vessel.

With these conservative assumptions, core cooling would be accomplished by Core Spray.

Time to core uncovery is not critical to this analysis because the success criteria (PCT

< 2200 F) is dependent upon the very low value of the linear heat generation in the fuel assemblies in the shutdown condition compared to that at full power.

For the Cold Shutdown case, a leak is assumed in the primary system piping due to an event such as valve mispositioning or maintenance.

One loop of RHR is assumed to be in the SDC mode.

The other loop is assumed to be unavailable.

The most likely leak would be a coolant loss through the SDC suction line.

A single failure is assumed to render one CSS injection valve inoperable while the other CSS loop is inoperable per Technical Specification 3.5.A.4.

As a result, no automatic ECCS injection capability would be available.

The ensuing reactor drain down would result in a primary containment isolation system (PCIS) isolation, which includes the SDC primary containment isolation valves (PCIVs), at reactor vessel low water level 538" above vessel zero.

Closure of the SDC PCIVs will terminate the event within 40 seconds of the isolation signal (the maximum stroke time from full open to full close is 40 seconds).

The control system for each SDC isolation valve is designed to provide closure of the valve in time to prevent uncovering the fuel as a

result of a break in the pipeline.

Therefore, no mitigating actions are required to preclude uncovering the core (top of the active fuel is approximately 360" above vessel zero).

Isolation of the line in 40 seconds will also preclude reaching the CSS and LPCI automatic initiation setpoint at reactor vessel low water level 398" above vessel zero.

However, operators are assumed to begin manually realigning RHR to the LPCI mode upon receipt of the reactor vessel low water level scram/isolation signal.

NRC ITEM 3 Describe the indications and alarms available to the operator which will inform him of the need to realign the residual heat removal (RHR) system from shutdown cooling (SDC) to low pressure coolant injection (LPCI) mode.

Discuss the actions required for this realignment, and provide an estimate of the time required to complete them and refill the reactor.

TVA RESPONSE The operator would be initially aware of a loss of coolant inventory upon receipt of an abnormal water level alarm at approximately 555" above vessel zero (6" below normal water level) and an ADS blowdown permissive alarm at approximately 546" above vessel zero.

The operator would be prompted to realign the RHR system from the SDC mode to the LPCI mode upon receipt of reactor vessel low-water scram/isolation indications and alarms (at 538" above vessel zero) in the control room.

LPCI is arranged for automatic operation and for remote-manual operation from the control room.

The operator would realign RHR by closing the RHR pump shutdown cooling suction valves, opening the RHR pump suppression pool suction valves, resetting actuation logic and opening the LPCI injection valves.

These actions can be completed in approximately 3.5 minutes.

If reactor water level reached 398" above vessel zero, the Emergency Operating Instructions would direct the operator to manually initiate LPCI if it has not automatically initiated.

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