ML20137C608

Fulfills 30 Day Reporting Requirement of 10CFR50.45(a)(3) for LaSalle County Nuclear Power Station,Units 1 & 2

ML20137C608
Person / Time
Site:	LaSalle
Issue date:	03/21/1997
From:	Subalusky W COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9703250058
Download: ML20137C608 (11)

Text

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Coninu>nwealth litiwn Company j *

, 12alle Generating $tation i

JNil Nortlilist iload Marwilles 11. 61311 -9757

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l i March 21,1997 i

j United States Nuclear Regulatory Commission I i Attention: Document Control Desk l Washington, D.C. 20555

Subject:

LaSalle County Nuclear Power Station Units 1 and 2 i Plant Specific ECCS Evaluation Changes -

i 10 CFR 50.46 Report Facility Operating Licenses

! i NPF-11 and NPF-18 NRC Dockets Nos. 50-373 and I 50-374 i l

References:

1. "LaSalle County Station Units 1 and 2 l

SAFER /GESTR-LOCA Loss-of-Coolant

. Accident Analysis", NEDC-32258P, l October,1993.

! 2. Comed Letter W.T. Subalusky (LaSalle) to USNRC, "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", January 16,1997.

3. Siemens Power Corporation letter (H. D.

Curet) to USNRC, " Notification of '

Supplemental Reflood Requirement",

HDC:97:008, January 21,1997.

I I 9703250058 970321 PDR ADOCK 05000373.

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4. 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", j Siemens Power Corporation, May 1996", i 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 i transmittal also includes the PCT assessments for Unit 1 to provide a i

complete report for both LaSalle Units 1 and 2.

1 2 The following attachments provide updated information regarding the PCTs for the Loss of Coolant Accident (LOCA) analyses of record.

i 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) l 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 March 21,1997. The assessment notes (Attachment 4) provide a detailed description for each change or error reported.

Unit 1 The current General Electric 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 MAPLHGR limits calculated by GE will still apply to the GE fuel. The accumulation of the absolute magnitude of all previous changes described in Attachments 1 and 4 is less than 50 F for Unit 1. There have been no changes to the Unit 1 PCT assessments since the Reference 210 CFR 50.46 transmittal.

4 4

Unit 2 Siemens Power Corporation (SPC) notified Comed of errors in the current <

LOCA analysis for the ATRIUM -9B fuel for LaSalle Unit 2 Cycle 8. The accumulation of the absolute magnitude of changes or errors 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. Although LaSalle has not operated Unit 2 with the ATRIUM -9B fuel, the LOCA analysis errors are still being tracked and reported per 10 CFR 50.46. The initial 10 CFR 50.46 report for the introduction of ATRIUM -9B fuel was docketed in Reference 2. A detailed i description of the errors and PCT changes is given in Attachment 4.

SPC has recalculated the LOCA/ECCS ATRIUM -9B fuel analysis PCT for  !

the limiting break and single failure using approved methodology. This  !

calculation establishes the revised PCT value for ATRIUM -9B fuel. The PCT change also includes the revised supplemental reflood criteria that was presented to the NRC and documented in Reference 3. 1 In addition to the reporting of the PCT and assessments for SPC fuel for Unit 2, 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.  !

There have been no changes to the Unit 2 PCT assessments for GE fuel since the Reference 210 CFR 50.46 transmittal.

in May 1996, SPC submitted a revised LOCA Evaluation Model methodology _

(Reference 4) 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 NRC approved LOCA Evaluation Model with the overly conservative BWR Jet pump modeling. This I has resulted in reduced thermallimit and PCT margins. Upon NRC approval I of the SPC revised LOCA Evaluation Model, Comed plans to revise the ,

reference LOCA analysis for ATRIUM -9B 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. Comed will

then submit a revised 50.46 letter to document the PCT for the revised I

analysis of record.

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' If there are any questions or comments concerning this letter, please refer them to Perry Bames, Regulatory Assurance Manager, at (815) 357-6761, 1 extension 2383.

Respectfully, l

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l W. T. Subalusky Site Vice President LaSalle County Station j Enclosure i cc: A. B. Beach, NRC Region ill Administrator 1 M. P. Huber, NRC Senior Resident inspector - LaSalle i D. M. Skay, Project Manager - NRR - LaSalle F. Niziolek, Office of Nuclear Facility Safety - IDNS .

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Attachment 1 LaSalle Unit 110 CFR 50.46 Report (GE Fuel)

PLANT NAME: LaSalle Unit 1  !

ECCS EVALUATION MODEL: SAFER /GESTR LOCA REPORT REVISION DATE: 3/21/97 CURRENT OPERATING CYCLE: 8 ANALYSIS OF RECORD Evaluation Model Methodology: "GESTR-LOCA and SAFER Models for the 4 Evaluation of the Loss-of-Coolant Accident",

Volumes I,11 and III, 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 i

"LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA less-of-Coolant Accident Analysis", NEDC-31510P, December,1987. i Fuel: P8x8R, GE8x8EB and GE8x8NB (Note 1)

Limiting Single Failure: HPCS Diesel Generator l Limiting Break Size and Location: Double Ended Guillotine of Recirculation Suction Piping Reference PCT: PCT = 1260 F 1

MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS I

Bottom Head Drain Issue (Note 2) APCT =+10 F SAFER /GESTR Automation Error (Note 3) APCT =+30 F B. CURRENT LOCA MODEL ASSESSMENTS None NET PCT: PCT = 1300 F Page 1 of 7

i 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: 3/21/97 CURRENT OPERATING CYCLE: 8 ANALYSIS OF RECORD l Eva'uation Model Methodology: "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident",

Volumes I, Il and Ill, NEDE-23785-1-P-A, ,

February,1985. I 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 I and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", NEDC-31510P, December,1987.

Fuel: P8x8R, GE8x8EB and GE8x8NB (Note 1) l Limiting Single Failure: HPCS Diesel Generator ,

Limiting Break Size and Location: Double Ended Guillotine of Recirculation Suction  !

Piping Reference PCT: PCT = 1260 F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS Bottom Head Drain issue (Note 2) APCT =+10 F S.AFER/GESTR Automation Error (Note 3) APCT =+30 F B. CURRENT LOCA MODEL ASSESSMENTS None NET PCT: PCT = 1300 F Page 2 of 7

l Attachment 3 j LaSalle Unit 210 CFR 50.46 Report (SPC Fuel) l PLANT NAME: LaSalle Unit 2 1 ECCS EVALUATION MODEL: EXEM BWR Evaluation Model REPORT REVISION DATE: 3/21/97 CURRENT OPERATING CYCLE: 8 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, EMF-96-153(P), Revision 1, i August,1496. (Notes 2,4 and 5) and LOCA Break Spectrum Analysis for LaSalle Units 1 l and 2, EMF-96-152(P), August,1996. (Notes 2 and 4) i Fuel: ATRIUM -9B Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: Discharge side 0.5 ft' Recirculation Line Break Reference PCT: PCT = 2161 F I MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None B. CURRENT LOCA MODEL ASSESSMENTS l

HUXY Time Step Error (Note 6) APCT = +25 F HUXY System Pressure Error (Note 7) APCT = +58 F l Supplemental Reflood Requirement (Note 8) APCT = -115 F l

NET PCT: PCT = 2129 F Page 3 of 7

Attachment 4 LaSalle Units 1 and 2 PCT Assessment Notes

1. GE Fuel Types The GE S AFER/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 nath (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 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 believed that the impact is less than 10 F. Comed has 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 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.

I SPC has conservatively incorporated the effects of the BHD into the LaSalle LOCA analysis for ATRIUMm-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. l

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. Annlication of the EXEM BWR Evaluation Model Tojustify use of the ATRIUM *-9B fuel for 1,2C8, 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 ATRIUMm-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 Setnoint 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 12 inches above the top of active fuel. The GE LOCA analysis utilized a 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 Units 1 and 2 was also performed with a level setpoint at the top of active fuel. Although the SPC Break 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 inches of available margin between the Nominal Setpoint and the GE LOCA analysis input (top of active fuel).

Therefore,12 inches of this available margin was used in generating the LaSalle LOCA-ECCS Analysis MAPLHGR limits for SPC ATRIUM-9B fuel. The 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 analysis. This analysis setpoint change still bounds all of the instrument error, and this change is documented here for PCT margin tracking. Pending UFSAR 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 NRC currently approved LOCA Evaluation Model with the overly conservative BWR jet pump modeling used for this ATRIUM -9B fuel analysis resulted in reduced thermal limit and PCT margins. Upon NRC approval of the revised SPC LOCA Evaluation Model, Comed plans to reanalyze the reference LOCA analysis for ATRIUM -98 fuel with the revised methodology. The revised methodology will allow an 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 will maintain consistency with the previous GE analysis, even though adequate setpoint margin exists at the Level I setpoint assumed in the current SPC LOCA analysis.

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. Attachment 4 LaSalle Units 1 and 2 PCT Assessment Notes

6. HUXY Time Sten Error (PCT increase)

In February of 1997 SPC reported an error on the application of the HUXY code. The HUXY code is used to perform heatup calculations during the entire LOCA and yields peak clad temperature and local oxidation at the axial plane ofinterest. In the HUXY calculation, when the rod fails, the inside surface of the cladding is exposed to the coolant and significantly increases the Metal-Water-Reaction (MWR) rate. The MWR rate is inversely proportional to the oxide thickness layer en the surface and exponentially proportional to the metal temperature. At the time of rod failure, this relationship creates a situation where a large discrete change in the cladding heat source term is generated. This rapid increase in thermal source term, along with the nonlinear nature of the MWR, can create convergence problems in the numerical solution. SPC reported that previous analyses did not use a sufficiently small time step to ensure converged results following the failure. SPC reperformed the limiting analyses and determined the PCT impact from the HUXY time step selection was +25 F.

7. HUXY System Pressure Error (PCT increase)

In February of 1997, SPC reported an error on the application of the HUXY code. The HUXY code is used to perform heatup calculations for the entire LOCA transient and yields peak clad temperature and local oxidation at the axial plane ofinterest. HUXY uses the time dependent pressure difference across the fuel cladding to determine the amount of strain experienced by the cladding and the resulting potential for rod ballooning and rod failure. l The internal rod pressure is calculated by HUXY while the time dependent system pressure i is obtained from the blowdown and refill /reflood calculations. Recent LOCA analyses have used a constant 14.7 psia system pressure to provide the largest possible cladding pressure 1 difference during the event. The larger cladding pressure difference could cause rods to fail 1 that otherwise might not. The appicach of using a constant system pressure of 14.7 psia was thought to be conservative since the occurrence of fuel failures during an accident increases PCT and MWR. However, it was discovered that using a larger cladding pressure difference is not always conservative. SPC reperformed the limiting analyses and applied the time dependent pressure calculated during blowdown and refill /reflood to the HUXY heat up ,

calculation. SPC determined the PCT impact from the HUXY system pressure error for the  !

limiting case to be +58 F.

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, Attachment 4 I

' LaSalle Units 1 and 2 PCT Assessment Notes i

8. Supplemental Reflood Reauirement (PCT decrease)

The SPC FLEX computer code is used to determine the core and system response during the l reflood and refill phases of a LOCA. FLEX calculates the time when two-phase flow reaches i the hot node in the core. This is referred to as the " time of hot node reflood" or more simply I i

the time of core reflood. At core reflood, the heat transfer coefficients (HTCs) used in the

fuel pin heatup analysis increase by about an order of magnitude from the Appendix K spray ,

i HTC to the Appendix K reflood HTC. The higher HTC is sufficient to remove decay heat 1 l from the fuel rods and results in the termination of the fuel rod temperature excursion. It is the entrained liquid at the hot node that results in increased heat transfer effectiveness. The

! FLEX computer code uses the time that two-phase entrained liquid reaches the plane of interest as the definition of the time of hot node (core) reflood. The time of core reflood is

more explicitly defined in FLEX using the RELENT parameter. RELENT (relative l l

entrainment) is defined as the entrained liquid flow rate at the hot node divided by the core ,

I inlet liquid flow rate. A sustained non-zero value of RELENT is the criterion used to l 1

determine the time of core reflood for LOCA analyses using the NRC approved SPC EXEM BWR LOCA methodology.

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In recent analyses, including the analysis of record for LaSalle, SPC has applied a very f conservative supplemental reflood criteria of an absolute entrained liquid flow rate at the i plane of interest along with the sustained non-zero value of RELENT to determine the time of core reflood. Based on recent experimental data, SPC determined that a revised absolute entrained liquid flow rate is appropriate for the ATRIUM -9B fuel. The revised supplemental criteria would decrease the reflood time by several seconds and reduce the

' PCT, but the LOCA analyses would still be conservative relative to using the approved methodology with the RELENT criteria only. SPC presented the revised supplemental i

criteria to the NRC on January 9,1997 and provided an information letter on January 21,

1997 to document the supplemental criteria.

l SPC reperformed the limiting analyses with the improved supplemental reflood criteria. SPC

determined the PCT impact for the limiting case to be -115 F.

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