ML20206K708

Forwards 10CFR50.46(a)(3) Rept Re Significant Change in Calculated Pct.Loca Analyses for Both GE Fuel & Siemens Power Corp Fuel Demonstrates Results within All of Acceptance Criteria Set Forth in 10CFR50.46

ML20206K708
Person / Time
Site:	LaSalle
Issue date:	05/07/1999
From:	Jamie Benjamin COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9905130253
Download: ML20206K708 (7)

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May 7,1999 United States Nuclear Regulatory Commission i Attention: Document Control Desk j Washington, D.C. 20555

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LaSalle County Station, Units 1 and 2 l Facility Operating License Nos. NPF-11 and NPF-18 j NRC Docket Nos. 50-373 and 50-374  ;

Subject:

Report of Significant Change in Calculated Peak j Cladding Temperature (PCT)- 10 CFR 50.46 Report i

Reference:

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

(2) Comed Letter from J. A. Benjamin (Comed) to USNRC, 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, January 8,1999.

This letter fulfills the thirty day reporting requirement of 10 CFR 50.46(a)(3) for LaSalle County Nuclear Power Station Unit 2. The previous 50.46 report (Reference 2) was an annual report. Since the Reference 2 report, there has been a significant change to the calculated PCT for Siemens Power Corporation (SPC) fuel based on reanalysis using the latest approved evaluation rnodel. This re-baseline analysis resulted in an improvement of '

227 from the PCT value provided in the Reference 2 report. /

p i

in addition to reporting tt.:s significant change, we are including in our submittei all other changes to Peak Cladding Temperature since the submittal of Reference 2. This submittal will also fulfill the annual reporting (\3 00 requirements of 10CFR50.46(a)(3).

The Loss of Coolant Accident analyses of record for both GE fuel and SPC fuel demonstrates the results within all of the acceptance criteria set forth in 10 CFR 50.46, therefore no further action is required.

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May 7,1999 U.S. Nuclear Regulatory Commission Page 2 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 May 5,1999. The assessment notes (Attachment 4) ,

provide a detailed description for each change or error reported.  !

Unit 1 The current General Electric LOCA analysis (Reference 1) was approved in 1993 and utilizes approved methodology. It applies to all fuel operating in l Unit 1 (currently all GE fuel), and the MAPLHGR limits calculated by GE ,

apply to all fuel in the core. There has been one change to the Unit 1 PCT '

assessments for GE fuel since the Reference 210 CFR 50.46 transmittal.

This change is a result of the leakage determined as a part of the jet pump riser flaw evaluation. The PCT impact of this change was conservatively determined to be 1 F.

Unit 2 The core design for Unit 2 Cycle 8 will be mixed core containing both GE and Siemens Fuel. The following changes are specified by fuel type.

GE Fuel:

General Electric calculates the GE fuel PCT for the GE fuel in the Unit 2 mixed core, and it is the same analysis as described for Unit 1 above. There has been one change to the Unit 2 PCT assessments for GE fuel since the Reference 210 CFR 50.46 transmittal. This change is a result of the leakage determined as a part of the jet pump riser flaw evaluation. The PCT impact of this change was conservatively determined to be 1 F.

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I May 7,1999 q

l U.S. Nuclear Regulatory Commission

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l SPC Fuel.  ;

i The LaSalle Unit 2 restart will contain a reload batch of SPC ATRIUM -98 l 1 fuel. The previous 10 CFR 50.46 report for the ATRIUM -9B fuel was  !

! docketed in Reference 2. Since the Reference 2 report, SPC has performed )

a new break spectrum and MAPLHGR limit analysis for LaSalle correcting ]

the known errors and assessments documented in the Reference 2 report. {

, These analyses were performed using the latest approved evaluation model l l as referenced in the LaSalle Unit 2 Technical Specifications. This report  !

! fulfills the 30 day reporting requirement of 10 CFR 50.46.  !

if there are any questions or comments concerning this letter, please refer i them to Frank A. Spangenberg,111, Regulatory Assurance Manager, at  !

(815) 357-6761, extension 2383. j l ,

I Respe ully, l l

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, i r FL J re - ~Benjamig j L alle C unt S ation I

Attachment l

cc: Regional Administrator - NRC Region lll  !

l NRC Senior Resident inspector - LaSalle County Station 1

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

PLANT NAME: LaSalle Unit 1 l ECCS EVALUATION MODEL: SAFER /GESTR LOCA l REPORT REVISION DATE: 5/5/99 l CURMNT OPERATING CYCLE: 8 ANALYSIS OF RECORD

, Evaluation Model Methodology: "GESTR-LOCA and SAFER Models for the l Evaluation of the Loss-of-Coolant Accident",

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

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

i and "LaSalle County Station Units 1 and 2 i

SAFER /GESTR-LOCA Loss-of-Coolant i

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 Ended Guillotine of Recirculation Suction Piping l

Reference PCT: PCT = 1260*F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS Bottom Head Drain issue (Note 2) APCT =+10 F SAFER /GESTR Automation Error (Note 3) APCT =+30*F B. CURRENT LOCA MODEL ASSESSMENTS (Since 1/8/99 submittal)

Jet Pump Riser Flaw Evaluation (Note 5) APCT =+1*F NET PCT: PCT = 1301*F Page1of5

'. Attachment 2 LaSalle Unit 210 CFR 50.46 Report (GE Fuel)  ;

PLANT NAME: LaSalle Unit 2 ECCS EVALUATION MODEL: SAFER /GESTR LOCA REPORT REVISION DATE: 5/5/99 i CURRENT OPERATING CYCLE: 8 (upon startup) l l

ANALYSIS OF RECORD Evaluation Model Methodology: "GESTR-LOCA and SAFER Models for the

))

Evaluation of the Loss-of-Coolant Accident", l Volumes I, il and Ill, NEDE-23785-1-P-A, i February,1985.

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

1993.  !

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

Fuel: P8x8R, GE8x8EB and GE8x8NB (Note 1) l i

Limiting Single Failure: HPCS Diesel Generator i

Limiting Break Size and Location: Double Ended Guillotine of Recirculation Suction Piping Reference PCT: PCT = 1260*F l l

MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS Bottom Head Drain Issue (Note 2) APCT =+10*F SAFER /GESTR Automation Error (Note 3) APCT =+30*F B. CURRENT LOCA MODEL ASSESSMENTS (Since 1/8/99 submittal)

Jet Pump Riser Flaw Evaluation (Note 5) APCT =+1*F L

NET PCT: PCT = 1301*F i

)

Page 2 of 5

I i Attachment 3 LaSalle Unit 210 CFR 50.46 Report (SPC Fuel)  !

l PLANT NAME: LaSalle Unit 2 ECCS EVALUATION MODEL: EXEM BWR Evaluation Model REPORT REVISION DATE: 5/05/99 CURRENT OPERATING CYCLE: 8 (upon startup)

ANALYSIS OF RECORD Evaluation Model Methodology: Advanced Nuclear Fuels Corporation l Methodology for Boiling Water Reactors EXEM l

BWR Evaluation Model, ANF-91-048(P)(A), '

January,1993.

BWR Jet Pump Model Revision for RELAX, ANF-91-048(P)(A), Supplement 1 and Supplement 2, Siemens Power Corporation, October 1997.

Calculation: LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM -9B Fuel, EMF-2175(P),

March,1999. (Notes 2,4 and 5) and LOCA Break Spectrum Analysis for LaSalle Units 1 and 2, EMF-2174(P), March 1999.

(Notes 2,4 and 5)

Fuel: ATRIUM -9B Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: Discharge side 1.1 ft2 Recirculation Line Break Reference PCT: PCT = 1807*F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None B. CURRENT LOCA MODEL ASSESSMENTS None NET PCT: PCT = 1807'F Page 3 of 5

L l Attachment 4 LaSalle Units 1 and 2 PCT Assessment Notes

1. GE Fuel Tvoes 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.

2. Bottom Head Drain (BHD) flow path (PCT increase)

In March of 1995, Comed asked GE to evaluate the impact of additional reactor coolant less 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 15,1995 submittal. Reactor Water Cleanup (RWCU) system operation takes suction from the BHD and from the recirculation suction piping, which are connected at a common point. A design basis LOCA 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 precisely evaluate the PCT impact of the recirculation line break LOCA including the BHD, it is Lelieved that the impact is less than 10 F. Comed determined that this error applied to LaSalle and the 10 F penalty has been included in the current LOCA model PCT assessments. The impact of the BHD exiting flow on maintaining level inside the shroud was also evaluated to be insignificant since the increased minimum makeup flow is well within the rnargins 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 conservatively incorporated the effects of the BHD into the LaSalle LOCA analysis for ATRIUM -9B fuel. The PCT impact of the BHD is reflected in the reference PCT for the SPC analysis, which is being applied at this time to Unit 2.

3. SAFER /GESTR Automation Error (PCT increase)

In June of 1996, GE reported an error to the USNRC for some applications of the GE LOCA Evaluation Model SAFER /GESTR. It was determined that in some analyses an algorithm used to compute the number of fuel rods in a BWR lattice was incorrectly specified. As a result, LOCA input prepared with the automation process may have included incorrect data. This error had impact on fuel designs containing a large water rod and analyses where the input generation was automated. Calculations performed to assess the significance of this error indicate that the impact on the calculated peak cladding temperature is less than 30 F. GE informed Comed on September 26,1996 that this error applies to the GE analysis for LaSalle Units 1 and 2.

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Attachment 4 LaSalle Units 1 and 2 PCT Assessment Notes

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. Leakaae Determined as a Part of the Jet Pump Riser Flaw Evaluation During L2R07, flaw indications were identified in jet pump risers 1/2 and 19/20.

GE has performed a LOCA evaluation assuming the maximum allowable flaw sizes are combined together and the worst of the units is evaluated. These maximum allowable flaw sizes correspond to a leakage that is documented in "LaSalle County Nuclear Power Station Jet Pump Riser Safety Evaluation, Evaluation of Riser Leakage Impact," GENE-A1300439-00-02P, dated ' March i 1999. This calculation shows a PCT increase of 1 F. GE has confirmed (and Comed concurs) that the LaSalle daily jet pump Technical Specification surveillance will detect a jet pump riser crack that results in a failed jet pump

("LaSalle County Nuclear Power Station Jet Pump Riser Safety Evaluation, Evaluation of Surveillance Monitoring Parameters," GE-NE-A13-00439-00-01P, Dated February 1999).

The jet pump riser flaw leakage was evaluated in the break spectrum analysis for l ATRIUM-9B fuel. The jet pump riser flaw leakage values identified during L2R07 are listed in "LaSalle Units 1 and 2 Principal LOCA Analysis Parameters," EMF-95-041, Revision 1, Siemens Power Corporation, dated March 1999 for the SPC LOCA Analysis. Siemens evaluated the impact of the jet pump riser flaw leakage and determined that there was no PCT impact on the LOCA analysis.

Hence the SPC Break Spectrum and MAPLHGR reports listed in Attachment 3 reflect the PCT including the jet pump riser flaw leakage.

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