ML20149M803

Forwards Plant Specific ECCS Evaluation Changes 10CFR50.46 Rept

ML20149M803
Person / Time
Site:	LaSalle
Issue date:	01/16/1997
From:	Subalusky W COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9701280023
Download: ML20149M803 (9)

Text

. Osmmonw calth Ediv>n Compan)-

. 12%tle Generating Mation

,. 2(d)I North list Road Marsci.lles 11 6134131'5' Tel H15-3574 61 January 16,1997 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

LaSalle County Nuclear Power Station Units 1 and 2 Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report Facility Operating Licenses NPF-11 and NPF-18 NRC Dockets Nos. 50-373 and 50-374

References:

1. "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident i Analysis", NEDC-32258P, October,1993.

2. NRC SER, D. M. Skay to I. M. Johnson,

" Issuance of Amendments (TAC NOS. M95156 and M95157)", October 29,1996.

3. SPC Letter (R. A. Copeland) to USNRC (Document Control Desk), "ANF-91-048(P),

Supplement 1 and ANF-91-048(NP), Supplement 1, "BWR Jet Pump Model Revision for RELAX",

. Siemens Power Corporation, May 1996.",

RAC:96:042, May 6,1996.

This letter fulfills the thirty day reporting requirement of 10 CFR 50.46(a)(3) for LaSalle County Nuclear Power Station Unit 2. The accumulation of the absolute magnitude of changes in the ECCS evaluation models (or the application of new, approved models) has resulted in a calculated Peak Cladding Temperature (PCT) difference of more than 50 F for Unit 2. This letter also fulfills the annual reporting requirement of 10 CFR 50.46(a)(3) for Unit 1 and Unit 2 of LaSalle County Nuclear Power Station.

9701280023 970116  !

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PDR ADOCK 05000373 P PDR A l'nioim o,mpan>

The following attachments provide updated information regarding the PCTs for the Loss of Coolant Accident (LOCA) analyses of record.

Attachment 1: LaSalle Unit 1 10 CFR 50.46 Report (GE Fuel)

Attachment 2: LaSalle Unit 210 CFR 50.46 Report (GE Fuel)

Attachment 3: LaSalle Unit 210 CFR 50.46 Report (SPC Fuel)

Attachment 4: LaSalle Units 1 and 2 PCT Assessment Notes Attachments 1-3 provide PCT information for the limiting Loss of Coolant Accident evaluations for LaSalle County Nuclear Power Station, including all assessments as of December 13,1996. The assessment notes (Attachment 4) provide a detailed description for each change or error reported.

Unit 1 The current General Electric (GE) LOCA analysis was approved in 1993 (Reference 1) and utilizes approved methodology. It applies to all fuel operating in Unit 1 (currently all GE fuel), and the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits calculated by GE will still ,

apply to the GE fuel. The accumulation of the absolute magnitude of all changes described in Attachments 1 and 4 is less than 50 F for Unit 1.

Unit 2 With the introduction of the Siemens Power Corporation (SPC)

ATRIUM -9B fuel for LaSalle Unit 2 Cycle 8, vendor specific LOCA analyses were required for operation with co-resident GE and SPC fuel. The <

GE LOCA analysis (Reference 1) is applicable to the GE fuel and the new SPC LOCA analysis is applicable to the SPC ATRIUM -9B fuel. Technical Specification Amendments (Reference 2) have been approved, allowing use of the SPC methodology at LaSalle. Comed has completed 10 CFR 50.59 safety evaluations for the introduction of ATRIUM -9B fuel and for the SPC LOCA analysis, which was performed with the methodologies approved for use at LaSalle in Reference 2.

SPC has calculated the LOCA/ECCS analysis PCT for the ATRIUM -9B fuel with their approved methodology. This calculation establishes the base PCT value for ATRIUM -9B fuel. The SPC calculation resulted in a PCT change for Unit 2 of more then 50 F. This 10 CFR 50.46 report also includes the PCT and all of the assessments for the co-resident GE fuel.

The GE fuel PCT is calculated by General Electric, and it is the same analysis as described for Unit 1 above.

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Siemens Power Corporation has demonstrated the hydraulic compatibility of the ATRIUM *-9B and GE fuel and concluded that the mixed core effects ,

have a negligible impact on , he PCT calculation. Therefore, since the GE calculated MAPLHGR limits i ill continue to be applied to the GE fuel, the l GE PCT calculation remains applicable for the GE fuel and the SPC PCT calculation is applicable to the ATRIUM -9B fuel in the mixed core.

In May 1996, SPC submitted a revised LOCA Evaluation Model methodology (Reference 3) that corrects the overly conservative modeling of the BWR jet pump used in the current approved methods. The SPC analyses performed for the LaSalle Unit 2 SPC fuel transition utilized the current NRC approved LOCA Evaluation Model with the overly conservative BWR jet pump modeling. This has resulted in reduced thermallimit and PCT margins.

Upon NRC approval of the SPC revised LOCA Evaluation Model, Comed plans to revise the reference LOCA analysis for ATRIUM -98 fuel with the revised methodology. The revised methodology would increase the thermal limit and PCT margins and would prevent LaSalle Station from experiencing thermal limit and operating power restrictions later in the operating cycle.

Corned will then submit a revised 50.46 letter to document the PCT change for this analysis.

1 If there are any questions or comments conceming this letter, please refer them to me at (815) 357-6761, extension 3600. l l

Respectfully, W. T. Subalusky Site Vice President LaSalle County Station Enclosures cc: A. B. Beach, NRC Region ill Administrator .

M. P. Huber, NRC Senior Resident inspector - LaSalle j D. M. Skay, Project Manager - NRR - LaSalle F. Niziolek, Office of Nuclear Facility Safety - IDNS DCD - Licensing (Hardcopy: Electronic: )

Central File

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Att: chm:nt 1

'. LaSalle Unit 110 CFR 50.46 Report (GE Fuel)

PLANT NAME: LaSalle Unit 1 ECCS EVALUATION MODEL: SAFER /GESTR LOCA REPORT REVISION DATE: 12/13/96 CURRENT OPERATING CYCLE: 8 ANALYSIS OF RECORD Evaluation Model Methodology: "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident",

Volumes I,11 and Ill, NEDE-23785-1-P-A, February,1985.

Calculation: "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", NEDC-32258P, October, 1993.

and "LaSalle County Station Units 1 and 2 i

SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", NEDC-31510P, December,1987.

Fuel: P.8x8R, GE8x8EB and GE8x8NB (Note 1)

. Limiting Single Failure: HPCS Diesel Generator  ;

Limiting Break Size and Location: Double Ended Guillotine of Recirculation 4

Suction Piping i Reference PCT: PCT = 1260 F MARGIN ALLOCATION l A. PRIOR LOCA MODEL ASSESSMENTS 1

None B. CURRENT LOCA MODEL ASSESSMENTS Bottom Head Drain issue (Note 2) APCT =+10 F SAFER /GESTR Automation Error (Note 3) APCT =+30 F

< NET PCT: PCT = 1300*F Page 1 of 6

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. l AttcChm:nt 2

'. LaSalle Unit 210 CFR 50.46 Report (GE Fuel) l PLANT NAME: LaSalle Unit 2 ,

ECCS EVALUATION MODEL: SAFER /GESTR LOCA  :

REPORT REVISION DATE: 12/13/96 l CURRENT OPERATING CYCLE: 8 ANALYSIS OF RECORD Evaluation Model Methodology: "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident",

Volumes I,11 and Ill, NEDE-23785-1-P-A, February,1985.

Calculation: "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", NEDC-32258P, October, 1993.

and "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", NEDC-31510P, December,1987.

Fuel: P8x8R, GE8x8EB and GE8x8NB (Note 1)

Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: Double Erided Guillotine of Recirculation Suction Piping Reference PCT: PCT = 1260*F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None B. CURRENT LOCA MODEL ASSESSMENTS Bottom Head Drain Issue (Note 2) APCT =+10 F SAFER /GESTR Automation Error (Note 3) APCT =+30 F NET PCT: PCT = 1300 F l

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s Attcchm:nt 3

' aSalle Unit 210 CFR 50.46 Report (SPC Fuel)

PLANT NAME: LaSalle Unit 2  ;

ECCS EVALUATION MODEL: EXEM BWR Evaluation Model l REPORT REVISION DATE: 12/13/96 l CURRENT OPERATING CYCLE: 8 i

. ANALYSIS OF RECORD Evaluation Model Methodology: Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A), January,1993.

Calculation: LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM -9B Fuel, a EMF-96-153(P), August,1996.

(Notes 2,4 and 5) l and j LOCA Break Spectrum Analysis for LaSalle Units 1 and 2, EMF-96-152(P), August,1996.

(Notes 2 and 4)

Fuel: ATRIUM -9B

) Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: Discharge side 0.5 fta Recirculation Line

, Break Reference PCT: PCT = 2161*F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None B. CURRENT LOCA MODEL ASSESSMENTS None NET PCT: PCT = 2161 F Page 3 of 6 .

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. 1 Attachment 4 l l

LaSalle Units 1 and 2 PCT Assessment Notes '

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1. GE Fuel Tvoes i The GE SAFER /GESTR LOCA analysis calculated the PCT for the P8x8R, GE8x8EB and GE8x8NB fuel types. The PCT reported is the highest PCT of the three fuel types (P8x8R). Although only the GE8x8NB fuel will be used for the current operating cycle (the P8x8R and GE8x8EB fuel types have been discharged to the fuel pool), the bounding PCT is used as the reference PCT for all GE fuel types available.

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2. Bottom Head Drain (BHD) flow path (PCT increase) in March of 1995, Comed asked GE to evaluate the impact of additional reactor coolant loss during a LOCA due to the cross tie of the bottom head drain (BHD) to the recirculation piping. General Electric reported this issue via a 50.46 report to the USNRC in a December 151995 submittal. Reactor Water Cleanup (RWCU) suction p, system iping, which operation takes are connected at a suction commonfrom point.the BHDbasis A design and LOCA from the recirculati where the break is on the recirculation suction piping would allow water in the lower plenum of the reactor vessel to be lost through the RWCU piping where it connects to the recirculation suction piping.

The GE evaluation concluded that while no analysis had been performed to ,

a precisely evaluate the PCT impact of the recirculation line break LOCA including l the BHD, it is believed that the impact is less than 10 F. Comed has determined I that this error applied to LaSalle and the 10 F penalty has been included in the current LOCA model PCT assessments. The im 3act of the BHD exiting flow on maintaining level inside the shroud was also eva uated to be insignificant since l

the increased minimum makeup flow is well within the margins available in the ECCS systems. The minimum makeup flow corresponds to that necessary to makeup for decay heat and for system leakages such as the BHD flow path.

. SPC has consentatively incorporated the effects of the BHD into the LaSalle l

LOCA analysis for ATRIUM -9B fuel. The PCT impact of the BHD is reflected I in the Reference PCT for the SPC analysis, which is being applied at this time to Unit 2. l 1

3. SAFER /GESTR Automation Error (PCTincrease) l 1

In June of 1996, GE reported an error to the USNRC for some applications of the I GE LOCA Evaluation Model SAFER /GESTR. It was determined that in some I analyses an algorithm used to compute the number of fuel rods in a BWR lattice was incorrectly specified. As a result, LOCA input pre aared with the automation ,

process may have included incorrect data. This error lad impact on fuel designs l i containing a large water rod and analyses where the input generation was l automated. Calculations performed to assess the significance of this error  !

indicate that the impact on the calculated peak cladding temperature is less than Page 4 of 6 l

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Attachment 4 LaSalle Units 1 and 2 PCT Assessment Notes I

30 F. GE informed Comed on September 26,1996 that this error applies to the GE analysis for LaSalle Units 1 and 2.

4. Application of the EXEM BWR Evaluation Model To justify use of the ATRIUM -9B fuel for L2C8, the LaSalle LOCA analysis has utilized the NRC approved SPC methodology. As a result of using this methodology, SPC calculated a different limiting break size and location than the previous GE analysis. The change in the limiting break and location is a result of applying the SPC methodology and it is not due to the use of the SPC ATRIUM -9B fuel. SPC has demonstrated the hydraulic compatibility of the ATRIUM -9B and GE fuel and concluded that the mixed core effects have a negligible impact on the PCT calculation. Therefore, the GE PCT calculation for the GE fuel remains applicable and the SPC PCT calculation is appropriate for the ATRIUM -9B fuel.

5. Reactor Water Level Low-Low-Low Level 1 Setooint For the LaSalle LOCA.-ECCS Analysis MAPLHGR Limits for ATRIUM -9B Fuel analysis, the Reactor Water Level Low-Low-Low Level 1 Setpoint was increased to 1P inches above the top of active fuel. The GE LOCA analysis utilized a l Reactor Water Level Low-Low-Low Level 1 Setpoint at the top of active fuel.

Note, the SPC LOCA Break Spectrum Analysis (EMF-96-152(P)) for LaSalle i Units 1 and 2 was also aerformed with a level setpoint at the top of active fuel.

Although the SPC Brea < Spectrum Analysis met all of the 10 CFR 50.46 acceptance criteria, the PCT margins were low. Therefore, the Reactor Water Level Low-Low-Low Level 1 Setpoint was increased to obtain additional PCT margin to the 10 CFR 50.46 acceptance criteria.

A setpoint error analysis performed by Comed showed 14" of available margin i between the Nominal Setpoint and the GE LOCA analysis input (top of active fuel). Thcafore,12 inches of this available margin was used in generating the LaSalle LOCA-ECCS Analysis MAPLHGR limits for SPC ATRIUM-98 fuel. The 4

level setpoint in this analysis was changed to obtain additional PCT margin to the 10 CFR 50.46 limit and the PCT value is reflected in the Reference PCT for this 1 analysis. This analysis setpoint change still bounds all of the instrument error, I and this change is documented here for PCT margin tracking. Pending UFSAR 1 changes also reflect the changed setpoint for the SPC LOCA analysis.

SPC has submitted a revised LOCA Evaluation Model methodology for NRC review. The revised LOCA Evaluation Model corrects the overly conservative modeling of the BWR jet pump. The current NRC approved LOCA Evaluation Model with the overly conservative BWR jet aump modeling that was used for this ATRIUM -98 fuel analysis has resultec in reduced thermallimit and PCT margins. Comed plans to reanalyze the reference LOCA analysis for ATRIUM *-98 fuel with the revised methodology u aon NRC approval of the revised SPC LOCA Evaluation Model. The revisec methodology will allow an l

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Attachment 4 LaSalle Units 1 and 2 PCT Assessment Notes increase to the thermal limit and PCT margins which will allow the analysis input for the Reactor Water Level Low-Low-Low Level 1 Setpoint to be reduced back to the top of active fuel. This is planned to be done to maintain consistency with the previous GE analysis, even though adequate setpoint margin exists at the Level 1 setpoint assumed in the current SPC LOCA analysis.

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