ML20195B246

Forwards Itemized Response to NRC 990506 RAI for TS Change Request 279 Re Core Protection Safety Limit

ML20195B246
Person / Time
Site:	Three Mile Island
Issue date:	05/21/1999
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1920-99-20269, NUDOCS 9905280199
Download: ML20195B246 (10)

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(- Route 441 South NUCLEAR Post Office Box 480 Middletown, PA 17057-0480 Tel717 944-7621 May 21,1999 1920-99-20269 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Ladies and Gentlemen:

Subje.t: Thw Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 Additional Information - Technical Specification Change Request No. 279 -

Core Protection Safety Limit This letter provides itemized responses (Attaciunent 1) to the NRC Request for Additional Information dated May 6,1999. Find design verification of the identified values provided in response to Question No.

1 is estimated to be completed by June 11,1999. GPU Nuclear will notify NRC when the design verification is complete.

If m.y additional information is needed, please contact Mr. David J. Distel, Nuclear Licensing and Regulatory Affairs at (973) 316-7955.

Sincerely, 7

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James W. Lange ach Vice President and Director, TMl

/DJD Attachments: (1) RAI Responses (2) Single Failure Capability Overview of the TMI-l Emergency Feedwater System \

(3) Emergency Feedwater System Schematic - .

A cc: Administrator, Region I TMI-l Senior Pwject Manager TMI-l Senior Resident inspector

.; l File No. 98195 ,,..

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY d/b/a i GPU ENERGY GPU NUCLEAR, Inc.

1 i Three Mile Island Nuclear Station, Unit 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 279 Response to Request for Additional Information (RAI)

COMMONWEALTH OF PENNSYLVANIA )

) SS:

COUNTY OF DAUPHIN )

This GPUN Inc. response to the NRC Staff s RAI on Technical Specification Change Request 279 is submitted in support ofLicensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my knowledge.

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Ece President and Direhor, ThU Sworn and subscribed before me this g/ yo M ,1999.

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Jnac b  ;

/.~ Noh Public g Notarial Seal i Suzanne C. Miklosik, Notary Public

1. @22 Mile's"fA R"%

Member, Pennsylvana Associategn of Ncp r<{

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1920-99-20269 Attachment 1 Page1of5 ATTACHMENT 1

1. NRC Ouestion Define specifically with respect to TSCR 279 what the total EFW (emergency feedwater) flow requirement is, assuming 20 percent average tubes plugged in the once-through-steam generators, including the flow breakdown, i.e., how much flow is necessary for decay heat removal, pump recirculation, bearing cooling, instrument inaccuracy, etc. and describe how the flow balance will be assured and maintained over time.

Response

The bounding design basis events for emergency feedwater (EFW) flow requirements that were reanalyzed for Technical Specification Change Request (TSCR) No. 279 are the Loss of Feedwater (LOFW) and Loss of All AC Power.

The specific event, acceptance criteria, and minimum required EFW flow are tabulated below.

Event Acceptance Criteria Minimum Required Notes EFW Flow

• Loss ofFeedwater e Thermal Power <112% 550 gpm total delivered . Single failure leaves (RCPs ON) flow @ 1065 psia 2 of the 3 EFW pumps e RCS < 2750 psig OTSG pressure available to provide flow.

• Pressurizer does not (275 gpm/OTSG) e No anticipatory reactor go solid trips credited.

• Analysisis sensitive to total net flow but not flow distribution between the OTSGs Loss of All AC e Thermal Power <112% 350 gpm total delivered . Only the TDP is available

?ower fl w @ 1065 psia on loss of all AC power e RCS < 2750 psig OTSG pressure e Resultant Dose

<10CFR100

• The accident acceptance criteria were met using these EFW flow rates in the analysis. Lower flows, although not currently analyzed, may also meet the acceptance criteria.

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[' Attachment 1

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L The LOFW event assumes that the EFW system delivers 550 gpm total flow to the once-through-steam generators (OTSG) at event pressure. . Assuming a single failure in the 1

- EFW system (See Attachment 2 for single failure discussion), the system must be able to deliver required flow to the OTSGs with 2 of the 3 pumps and only one (1) control valve

.(EF-V-30) open to each generator. All three'of the possible two-pump combinations must .i be capable of delivering the required flow (i.e.,2 motor driven pumps (MDP), the turbine  ;

driven pump (TDP) and MDP "A", 'or, the TDP and MDP "B"). I iThe Loss of All AC Power event assumes that the EFW system delivers 350 gpm total flow

- to the OTSGs at the event pressure. With both onisite and off-site AC power unavailable, only the TDP is ~available. Therefore, the TDP must be capable of delivering the required l '

flow (350 gpm total delivered flow). This event must cope with decay heat only, as tiie

- reactor coolant pumps (RCPs) are tripped with the loss of motive power.

Significant differences in EFW flow requirements are due to:

1 Loss of Feedwater Loss of All AC Power e Reactor Coolant Pump Heat e No Reactor Coolant Pump Heat e Reactor trip on resulting high RCS pressure e Immediate R.cactor trip on loss of AC power i The capability of the EFW system to deliver the above design basis flow requirements will  !

l continue to be verified as required by the TMI-1 Technical Specifications (once per l refueling interval or whenever the reactor is in cold shutdown for more than 30 days). The  !

surveillance test acceptance criteria for the EFW pumps and the method for generating those criteria are as follows:

• EFW Pumo Test Acceptance Criteria The surveillance' test acceptance criteria for each individual EFW pump are developed 4

- from the most limiting flow requirement. The LOFW event (RCPs ON) requires l 550 gpm total EFW flow delivered to the OTSGs at 1055 psia from any two of the three  !

l EFW pumps.' The approximate system head requirement with 550 gpm total flow

- delivered to the OTSGs at event pressure is 2662 feet . System head includes the effects of elevation differences, friction and velocity head due to flow and generator pressure. l The system head was established using a hydraulic model benchmarked to actual system l L data to validate' accuracy.

Since two pumps are available to deliver flow, individual pump performance is acceptable ifit develops this approximate value of head while delivering approximately half of the total flow requirement (e.gc, ~ 275 gpm), to the generator. The uncenainties due to instrumentation used b measure head and flow will be included in the values to establish the acceptance criteria used in the surveillance test. )

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' ~ Calculation is complete and being design verified.

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1920-99-20269 Attachment 1 -

Page 3 of 5

. EFWTDP Test Acceptance Ciiteria for Loss of All AC Power Additional acceptance criteria were developed for the turbine driven EFW pump based on the flow requirement for the Loss of All AC Power event. The event requires 350 gpm total EFW flow delivered to the OTSGs at 1065 psia. Only the TDP is available and it alone must deliver the flow requirement. The approximate system head requirement at 350 gpm total delivered flow to the OTSGs at event pressure is 2540 feet 4 System head includes the effects ofelevation differences, friction and velocity head due to flow and generator pressure.

The TDP performance was evaluated with respect to the TDP acceptance criteria for both LOFW and Loss of All AC Power events. The evaluation

• determined that the LOFW event is more limiting and, therefore, establishes the operability acceptance criteria for the turbine driven EFW pump.

TMI-l is currently revising the EFW refueling interval surveillance test procedure for the upcoming 13R outage (Fall 1999) to incorporate the new pump acceptance criteria. The surveillance test will operate each pump individually, injecting water from the condensate storage tanks into a single OTSG. The plant will be at cold shutdown with the OTSG depressurized during the test. Flow to the steam generators will be throttled until total flow delivered to the generator is at least the required rate (~ 275 gpm) including instrument uncertainty.

- Pump discharge pressure measured at pressure instrumentation locations downstream of the pump recirculation and bearing / seal cooling water branch lines, combined with pump suction pressure, will produce a conservative total dynamic head (TDH). EFW pump TDH acceptance criteria will assure the pump adequately meets the design basis head requirements including instrument uncertainty.-

- Analysis of past EFW system test results show that all pumps wil! meet the above i acceptance criteria. Test data also shows that the performance of all three pumps is essentially unchanged since initial installation in 1974.

The test acceptance criteria described above are based on total EFW flow delivered to the OTSGs. Changes in the pump flow breakdown (recirculation and bearing / seal cooling) will not affect operability provided the pump produces adequate head while deliverina required flow to the OTSGs. The surveillance testing is representative of normal system conditions, in that bearing / seal cooling and pump recirculation will operate normally during the test (i.e., are not isolated or throttled to improve test results). Therefore, using pump head and delivered flow alone as test acceptance criteria is appropriate to verify operability even if the pump flow distribution changes over the life of the pir.nt. I

• Calculation is complete and being design verified.

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1920-99-20269 Attachment 1 Page 4 of 5 As good engineering practice, additional information (e.g., total pump flow) will be recorded during refueling interval testing to assess the overall condition of the pumps. This l data allows trending of changes in recirculation or bearing / seal cooling flows. However, flow balancing of the EFW system is not performed, and branch line flows do not afTect the

' EFW pump acceptance criteria. The OTSG delivered flow is the key parameter (vice total pump flow) to which pump operability is appropriately verified to deliver required flow and total dynamic head.

Quarterly surveillance testing, as required by TMI-l Technical Specifications, operates each pump on recirculation and is used to monitor the EFW pumps for changes in recirculation and bearing / seal cooling flows between refueling outages.

2. NRC OuestioD

. Assuming maximum allowed pump degradation (for all three EFW pumps) as permitted by the ASME Code, compare the established EFW flow requirement above with the EFW flow that will  !

be delivered assuming the worst-case pump combination (assuming single failure). j i

Response

As recommended by NUREG 1482, TMI-l will take action before the pumps degrade beyond the most restrictive ofeither ,

1 (1) the operability test acceptance criteria described in response to Question 1 above, or  ;

(2) 10% degradation allowed by ASME Section XI and OM-6.

EFW pump degradation has been evaluated and determined to be limited by the test acceptance '

criteria based on the design basis event required flow and TDH. Allowing 10% degradation per i ASME Section XI and OM-6 would not produce acceptable pump performance for operability. -l Therefore, TMI-l will maintain the appropriate operability acceptance criteria that are more conservative than code limits for the EFW pumps.

3. NRC Ouestian l In responding to Question 1 of the NRC's letter dated March 10,1999, regarding TMI-l TSCR 279, you discussed the loss of feedwater accident and stated in a letter dated March 26,1999, that the EFW system was conservatively assumed to deliver flow (550 gpm) to the steam generators starting 43 seconds after the initiation signal with any (emphasis added) combination i ofEFW pumps (1 TDP or 2 MDPs or 1 TDP and 1 MDP). However, the information you

. provided during the April 23,1999, meeting with the staffindicated that the maximum predicted flow capability of the TDP is 536 gpm. Please either confirm the capability of the TDP to deliver the required 550 gpm flow, or amend your response to Question 1 to clarify the correct combination of pumps necessary to achieve the required 550 gpm flow.

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'1920 9t)-20269 Attachment 1 Page 5 of 5 This information is necessary for the staff to accurately understand and evaluate the TMI-l EFW pump capability as it applies to the flow requirement of 550 gpm total flow to the once-through-steam generators as stated in your March 26,1999, letter.

Response

l The GPU Nuclear letter to the NRC dated March 26,1999 (1920-99-20088), discussed EFW flow rates, initiation and delivery time and pump combinations in conjunction with the LOFW event. The LOFW analysis uses 550 gpm EFW flow to the steam generators beginning 43 seconds after the low OTSG level initiation signal. In this analysis, all EFW pumps are l bounded by the 43 second delay assumption. As such, the analysis does not rely upon any specific pump individually, or in combination, to deliver 550 gpm steam generator flow. . EFW flows actually beginning prior to the assumed 43 second delay or in excess of 550 gpm would benefit the event and are therefore not credited.

The GPU Nuclear March 26,1999, response cited the TDP alone as a possible EFW pump combination, assuming it could deliver 550 gpm to the OTSGs during the LOFW event.

However, the EFW system design basis does not require the TDP alone to provide the required system flow for the .LOFW event, as discussed in the response to Question 1 above.

Information presented during the April 23,1999, NRC meeting correctly describes the EFW system design basis as withstanding a single active failure. As such, system function relies upon the TDP to operate as a single pump only in the Loss of All AC Power event. No reliance has been placcd on the TDP alone to satisfy required EFW system flow for any other design basis event. This discussion clarifies the information previously provided in the GPU Nuclear letter dated March 26,1999.

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Attachment 2 Page1of2

, ATTACHMENT 2 Single Failure Capability Overview of the TMI-l Emergency Feedwater System Instrumentation and Control The Emergency Fcedwater System (EFW) is automatically initiated and controlled by the single failure proof Heat Sink Protection System (HSPS). Each of two actuation trains (A and B) are separately powered from diverse supplies initiated with 2 out of 4 channel logic. Four (4) instrument channels are powered from diverse IE supplies.

The worst case active failure (entire HSPS Train) affects only one pump and one flow control valve in each flow path. The system meets the following criteria:

1. No single failure will prevent the system design function. Sufficient redundancy to neither cause unwanted actuation nor prevent necessary system operation.

2. Mechan :nally and electrically separated so that a fault in one channel will not affect another.

HSPS Train A HSPS Train B Initiates MDP EF-P-2A Initiates MDP EF-P-2B Opens MS-V-13A & B starting the TDP EF-P-1 Opens MS-V-13A & B starting the TDP EF-P-1 Transfers flow control valves EF-V-30A & C from Transfers flow control valves EF-V-30B & D from 0" Startup Range (SU) to 25" SU/50% Operating 0" Startup Range (SU) to 25" SU/50% Operating Range OTSG level setpoint. Range OTSG level setpoint.

Mechanical Desian The EFW System was modified to make it single failure proof and safety grade including:

1. Parallel EFW flow control valves (EF-V-30C & D) to the system

2. Removed low OTSG pressure isolation signal to the flow contro? valves

3. Cavitating venturi in the injection line to each OTSG

4. Locked open pump recirculation lines Mechanical components in the system required to change state are:

1. Three pumps and associated drivers (EF-P-1, EF-P-2A/B),

2. Parallel flow control valves (EF-V-30A/B/C/D), and,

3. Steam admission and pressure regulating valves (MS-V-13 A/B, MS-V-6).

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Attachment 2 Page 2 of2 The failure of any one mechanical component would only result in either:

1. . Loss of one EFW pump .

2. Loss of one flow control valve associated with one OTSG

3. Failure of MS-V-6, causing a loss of the turbine driven pump

4. Failure of one MS-V-13 valve, causing a delay in the start of the TDP

Electrical Desian -

Power supplies for the significant components of the EFW system are shown on Attachment 3.

The motor driven pumps are powered from separate' emergency diesel backed IE switchgear.

The IC ES power supply provides power to both MS-V-2A and MS-V-2B. However, the valves are normally open and are not required to change state to perform their safety function.

Therefore, no single electrical failure will prevent performance of the EFW system safety function.

' Environmental Considerations The EFW system function considers the consequences of design basis events. These consequences include pipe whip, jet impingement and harsh environment (pressure, temperature, radiation and humidity).

The EFW system is not subjected to pipe whip orjet impingement effects from high energy line breaks or LOCAs. The system function continues in the post-LOCA radiation environment.

. Instruments located inside the Reactor Building have been environmentally qualified for LOCA and HELB temperature and pressure conditions. HSPS equipment located in the' control building is not subject to the harsh temperature or pressures resulting from LOCA or HELB events.-

EFW pumps and valves are located in the Intermediate Building. This equipment is not subjected to harsh temperature or pressures caused by LOCAs or HELBs in the Reactor Building. EFW equipment in the Intermediate Building has been environmentally qualified for the worst-case HELB environment (steam line break). The Intermediate Building has available volume to provide sufficient time for the operator to detect and isolate a feed line break in the building b.Gre the flood level can affect EFW system operation.

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