ML20073A413

Forwards Response to RAI Re Planned Implementation of Power Rerate Program at Util & Subject to Operating License Change Request 93-94-0,dtd 931209

ML20073A413
Person / Time
Site:	Limerick
Issue date:	09/09/1994
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9409200253
Download: ML20073A413 (7)

Text

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.PECO ENERGY Nuclear Group Headauarters 965 Chesterbrook Boulevard wayne, PA 19087-5691 l

l September 9,1994

. Docket Nos. 50-352

. 50-353

l. License Nos. NPF-39 NPF-85 i

. U. S. Nuclear Regulatory Commission l Attn: Document Control Desk i Washington, DC 20555

SUBJECT:

Limerick Generating Station, Units 1 and 2 l' Response to Request for Additional Information Regarding Power Rerate Program dated .

j. August 11,1994 (RAl-2)-

Gentlemen

i l . .

! Attached is our response to your Request for Additional Informrjon (RAI), discussed in our l telephone conversation on August 11,1994, regarding the planned implementation of the Power Rerate Program at Limerick Generating Station (LGS), Units 1 and 2. The Power Berate

Program is the subject of Operating License Change Request No. 93-24-0 which was forwarded to you by letter dated December 9,1993.

If you have any questions, please do not hesitate to contact us. ,

Ve truly yours, 4.

. A. Hunger, Jr., [

Director - Licensing j t

Attachment cc: T. T. Martin, Administrator, Region I, USNRC w/ attachment

. N. S. Perry, USNRC Senior Resident inspector, LGS - w/ attachment j l

R. R. Janati, Director, PA Bureau of Radiological Protection - w/ attachment l

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j COUNTY OF CHESTER  : l l i i: j 1 ,

D. M. Smith, being first duly sworn, deposes and saysi 7

j That he is Senior Vice President and Chief Nuclear Officer of PECO Energy Company, the Applicant herein; that he has road the enclosed response to the NRC Request for 1

Additional information involving Power Rerate discussed on August 11,1994, j concerning Operating License Change Request (No.' 93-24-0) for Limerick Generating ,

1-j Station Facility Operating Ucense Nos. NPF-39 and NPF-85, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to i c '

j the best of his knowledge, information and belief.

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i Senior Vic& Presideht and *

} Chief Nuclear Officer  !

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Subscribed and sw rn to before me this day o '

M 1994.

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Erica A.Santon, Notary Pubic TrecyMn Tve,ChesterCourtv Whnisson ExamsJuly10.1995

Dockst Nos. 50-352 50-353

. License Nos. NPF-39 NPF-85 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAl-2)

LIMERICK GENERATING STATION, UNITS 1 AND 2 (Per Telecon dated August 11,1994)

Ouestion 1

Define ICA Region stability exclusion region in Figure 2-1, " Power / Flow Operating Map for Power Rerate," of NEDC-32225P. State that the ICA region and the stability exclusion region are the same thing.

Resoonse 1:

The potential for BWR Core Thermal-Hydraulic Instability is documented in GE Service Information Letter (SIL) 380, Revision 1. The SIL's recommendations have been included in the LGS 1 and 2 Operating procedures. In addition, GE and the BWROG have developed interim corrective actions (ICA's) to further address core stability concerns. NRC Bulletin No. 88-07, Supplement 1, " Power Oscillations in Boiling Water Reactors (BWR's)", has endorsed these ICA's which have been implemented at LGS 1 4

and 2. The ICA's included " Operating Exclusion Regions" on the LGS Power / Flow Map, tht s the ICA region and the stability exclusion region are the same thing.

Inadvertent entry into these regions requires immediate action for exiting the region.

The exclusion regions are bounded by the natural circulation line, the minimum recirculation pump speed line and the 45% of rated core flow line. These flow boundaries are essentially unaffected by power rerate since only the power axis of the

power / flow map is rescaled by power rerate and rated core flow is not changed. The exclusion regions are also bounded by the 80% rod line, the 100% rod line and the maximum flow control rod line. The maximum flow control rod line is unaffected by power rerate; however, the 80% and 100% rod lines are affected. To ensure that the ICA's provide the same level of protection at power rerate conditions as they do currently, the 80% and 100% rod lines have been redefined such that the absolute power (MWT) values of these boundaries are unchanged (i.e.,80% becomes 76.2%

(80/1.05) and 100% becomes 95.2% (100/1.05). Thus, the power rerate ICA exclusion region boundaries provide the same level of protection against potential stability events as the current boundaries.

Question 2:

Section 2.5.1, " Control Rod Drive Hydraulic System", of NEDC-32225P.

CRD pump modifications - What are they?

Please state what the mods are.

i l .

l Docket Nos. 50-352 50-353

. Ucense Nos. NPF-39 NPF-85 l

l Resoonse 2:

Mod. P00195 replaces the existing CRD pumps, motors, and gear boxes with new, higher-capacity, direct-drive (i.e., pump and motor, no gear box) pumps. This is the only modification being performed on the CRD system involving Power Rerate.

This modification is being performed because, the evaluation for power rerate determined that the existing pumps do not have sufficient capacity under rerate l

conditions during normal CRD positioning operations. This is due.to high line losses from the discharge of the pump to the CRD flow control station. Maintaining the required differential pressure of 250 psid between the CRD system and reactor-pressure would result in cooling water flows of approximately 35 gpm, which may result in the increase of the number of drives running hot. After discussion with the existing pump manufacturer and review of the pump operating history, a determination  !

l was made that the pumps be replaced. This replacement includes pumps and motors l and the removal of the existing pumps, motors, and gear boxes.

Question 3:

l Section 3.4, " Reactor Recirculaton System" (RRS), of NEDC-32225P. Vibration l Evaluation - Does it exist? Provide information on vibration analysis. . What is the LGS

( plan to test for it? Commit to a Vibration test.

Resoonse 3:

A detailed vibration analysis was perforn.ad for the reactor recirculation system piping for rerate conditions. The impact of rerate resulted in a negligible effect.

A qualitative assessment of the rerate vibration conditions for the recirculation pump shaft was performed. The increased speed of the recirculation pump due to rerate (less than 1%) will not lead to significant increases in shaft mechanical stresses.

A qualitative analysis of the impact of rerate conditions on the recirculation pump and pump motor vibration was performed by General Electric. It was concluded that rerate conditions would not impact pump or pump motor vibration levels.

The recirculation pump, pump motor, and piping were determined to remain within ASME code allowable values for stresses.

i i

Docket Nos. 50-352 50-353

. License Nos. NPF-39 NPF-85 Recently, a phenomenon described as " Containment Noise" has occurred at another

' U.S. BWR. This phenomenon has been related to operation at increased core flow (ICF) above 100% rated flow. The increased recirculation pump speed associated with ICF is theorized to be a contributing factor to this phenomenon. ' This phenomenon is not believed to be related to power rerate since only a very small increase in recirculation pump speed is required to maintain a'given core flow at rerate conditions.

An investigation at this other U.S. BWR to determine the source of this phenomenon is l underway. PECO Energy is following this investigation closely.

LGS Units 1 and 2 have operated with ICF up to 105% rated flow for many cycles. No incidents of the " Containment Noise" phenomenon have been reported at LGS.

Analyses have been performed for LGS Justfying ICF operation up to 110% rated flow l including rerate conditions. PECO Energy plans to implement 110% ICF near the end I of the LGS. Unit 2 Cycle 3 Operating Cycle (September 1994), prior to the implementation of power rerate the following cycle.

PECO Energy is developing a monitoring program for the implementation of 110% ICF.

This program is likely to include baseline measurements and trending of noise and vibration levels in key areas of the reactor building and monitoring the vibration-instrumentation currently installed on the recirculation pump motors and shafts.

L Implementation of 110% ICF will consist of a gradual progression from 105% core flow up to 110% core flow over a period of approximately one week. This approach will allow PECO Energy to closely monitor the effects of the increasing recirculation pump _

speed and respond appropriately if the " Containment Noise" phenomenon is encountered.

Question 4:

Section 4.2.1, "High Pressure Coolant injection" (HPCI). GE SIL 480 (O ierspeed) should be implemented for HPCI pump. Has it been implemented? If not, does LGS plan to implement it? i Resoonse 4:

GE SIL 480 has been incorporated into the LGS HPCI system.

Question 5:  ;

Section 9.3.1, " Anticipated Transients Without Scram" (ATWS), of NEDC-32225P. How will ATWS acceptance criteria be met? What was " previous analysis"? GE Generic Analysis shows 20 psi, LGS is 40 psi. Need clarification.

Docket Nos.- 50-352 50-353'

. License Nos. NPF-39 NPF-85  !

i Response 5:

All the plant parameter changes for LGS Power Rerate are within the generic criteria, i except for the ATWS high pressure'setpoint, as stated in Section 9.3.1. The setpoint ,

was increased 40 psi rather than the 20 psi covered by the generic power uprate . . i ATWS evaluaton. A series of test cases were run for a similar BWR/4 to evaluate the ' ,

effects of certain inputs on the peak pressure for the limiting ATWS event, namely .

MSIV closure. ' A reduction of the ATWS high pressure setpoint by 30 psi produced .

only an 8 psi drop.in the peak pressure in a sensitivity test case. Although the 20 psi  !

additional increase was expected to increase the peak vessel pressure by less than 10 t psi, plant-speedic ATWS analysis was performed for LGS Power Rerate. j The LGS-plant specdic ATWS analysis was performed at 3458 MWt power level. The .

analysis assumed 28% boron enrichment and 'a 40 psi increase in the high pressure j setpoint. ATWS acceptance' criteria are met in the results of this plant-specdic  !

analysis. Peak vessel pressure is 1371 psig,'which is within the 1375 psig LGS l criterion. Peak cladding temperature is 1501* F,which is less than the 2200* Fcriterion'. f Peak suppression pool temperature is 188.8* F, which is within the 190* F generic q

criterion.  ;

i I L Question 6: -

4 Page 5-5 of SAFER /GESTR LOCA Report, NEDC-32170P, Rev.- 1 ' Jet pump  ;

j uncovery" - is it the same as 2/3 core coverage? State what it is. .;

l Resoonse 6.  ;

Both terms refer to the same elevation, but at different times in the event. " Jet pump  !

uncovery" is a term used to describe when the water level in the downcomer drops below 2/3 core height during the short-term blowdown phase for a large recirculation- -

line break (analyzed using LAMB code) in a LOCA event. The term "2/3 core l coverage" is generally used to describe the long-term response in which Jet pumps i enable flooding of the core up to the level of the. jet-pump nozzles after a large break. '

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! i Docket Nos. 50-352 50-353

. License Nos. NPF-39 NPF-85 i Question 7: i Section 5.3.2 of SAFER /GESTR LOCA report, NEDC-32170P, Rev.'1 (Section on-MELLLA calculation) What are PCT MELLLA values? l.E.,20 Fdelta between 2 PCT's-(What are these PCT's?)

Resoonse 7: .

The PCT impact from MELLLA is conservatively stated in section 5.3.2 as less than L 20 F. SAFER calculations were performed at 75%_ initial core flow with BP/P8x8R and l GE11 fuel using Appendix K assumptions. The results are tabulated below: ,

I PCT ( F)at PCT ( F)at }

Fuel Tvoe Rated Core Flow 75% Ogre Flow i BP/P8x8R 1619 1604- -

GE11 1577 1582 l.

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