ML20198N693
| ML20198N693 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/09/1998 |
| From: | Subalusky W COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9801210173 | |
| Download: ML20198N693 (10) | |
Text
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- m o w a s 2iu in. a Mawilks 1161 S a l T4' f rl Hi % 4%"4.*bt January 9,1998 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555
Subject:
Plant Specific ECCS Evaluation Changes -
10 CFR 50.46 Report LaSalle County Station, Units 1 and 2 Facility Operating License NPF 11 and NPF-18 NRC Docket Nos. 50-373 and 50 374
References:
1.
"LaSalle County Station Units 1 and 2 SAFER /GESTR LOCA Loss-of Coolant Accident Analysis", NEDC 32258P, October 1993.
2.
Comed Letter W. T. Subalusky (LaSalle) to U.S. NRC, "LaSalle County Nuclear Power Station Units 1 and 2 Plant Specific ECCS Changes - 10 CFR 50.46 Report Facility Operating Licenses NPF 11 and NPF-18 NRC Dockets Nos. 50 373 end 50-374",
March 21,1997.
This leiter 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 cf new, approved models) has resulted in an estin ated Peak Cladding Temperature (PCT) difference of more than 50*F for Unit 2. This transmittal also includes the PCT assessments for Unit 1 to provide a complete report for both LaSalle Units 1 and 2. This letter also fulfills the annual requirement for 10 CFR 50.46 reporting for both Units 1 and 2.
E The following attachments provide updated information regarding the PCTs g
for the Loss of Coolant Accident (LOCA) analyses of record.
_, : LaSalla Unit 1 10 CFR 50.46 Repoit (GE Fuel)
~ # 'b i ll : LaSalle Unit 210 CFR 50.46 Report (GE Fuel) : LaSalle Unit 210 CFR 50.46 Report (SPC Fuel)
, : LaSalle Units 1 and 2 PCT Assessment Notes o
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Attachments 13 provide PCT information for the limiting Loss of Coolant Accident evaluations for LaSalle County Nuclear Power Station, including all assessments as of January 9,1998. The assessment notes (Attachment 4) provide a detailed description foi each change or error reported.
Unit 1 The current General Electric LOCA analysis (Reference 1) was approved in 1993 and utilizes appreved methodology. It appliec to all fuel operating in Unit 1 (currently all GE fuel), and the MAPLHGR limits calculated by GE apply to all fuelin the core. T he 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.
Unit 2 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 have been no changes to the Unit 2 PCT assessments for GE fuel since the Reference 210 CFR f0.46 transmittal.
SPC Fuel:
Siemens Power Corporation (SPC) notified Corned 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 an estimated Peak Cladding Temperature (PCT) difference of more than 50 F for Unit 2. Altuugh LaSalle has not operated Unit 2 with the ATRIUM 9B fuel, the LOCA analysis errors are still oeing tracked and reported per 10 CFR 50.46. A detailed description of the errors and PCT changes is given in Attachment 4.
The previous 10 CFR 50.46 report for the ATRIUM 9B fuel was docketed in Reference 2. The previous 10 CFR 50.46 report documented assessments for the HUXY time step error, the HUXY system pressure error and the changes for the suppl 9 mental reflood requirement. Since the last report, these changes have been incorporated into a revised MAPLHGR analysis. The MAPLHGR analysis is a heatup calculation for the limiting break size and single failure and produces the Analysis of Record (AOR)
PCT. Therefore, the SPC Reference PCT includes the PCT impact from the previous model assessments. SPC also concluded that the HUXY time step error, the HUXY system pressure error and the changes for the supplemental reflood requirement did not change the Break Spectrum report conclusion for the limiting break size and single failure.
1 t
Comed plans for SPC to reanalyze the LOCA for the limiting case, correcting the latest errors. This reanalysis is scheduled for completion and approval per 10 CFR 50.59 prior to Unit 2 startup.
If there are any questions or comments conceming this letter, please refer them to Perry Barnes, Regulatory Assurance Supervisor, at (815) 357 6761, extension 2383.
Respectfully, i
~2==
W. T. Subalusky Site Vice President LaSalle County Station Attachment cc:
A. B. Beach, NRC Region 111 Administrator M. P. Huber, NRC Senior Resident inspector - LaSalle D. M. Skay, Project Manager NRR LaSalle F. Niziotek, Office of Nuclear Facility Safety - lDNS a
LaSalle Unit 110 CFR 50.46 Report (GE Fuel)
PLANT NAME:
LaSalle Unit 1 ECCS EVALUATION MODEL:
SAFER /GESTR LOCA REPORT REVISION DATE:
1/9/98 CURRENT OPERATING CYCLE:
8 ANALYSIS OF RECORD Evaluation Model Methodology:
Volumes I,11 and Ill, NEDE 237851-P A, February,1985.
Calculation:
"LaSalle County Station Units 1 and 2 SAFER /GESTR LCDA 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 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
' SAFER /GESTR Automation Error (Note 3)
APCT =+30*F B.
CURS?NT LOCA MODEL ASSESSMENTS None NET PCT:
PCT = 1300'F Page 1 of 7
LaSalle Unit 210 CFR 50.46 Report (GE Fuel)
PLANT NAME:
LaSalle Unit 2 ECCS EVALUATION MODEL:
SAFER /GESTR LOCA REPORT REVISION DATE:
1/9/98 CURRENT OPERATING CYCLE:
8 (upon startup) i ANALYSIS OF RECORD Evaluation Model Methodology:
"CESTR-LOCA and SAFEA Models for the Evaluation of tha Loss-of-Coolant Accident",
Volumes I,11 and lil, NEDE-237851 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" Nr" 31510P, December, 1987.
Fuel:
P8x8R, G'
..J8x8NB (Note 1) 1 Limiting Single Failure:
HPCS Diesv. denerator Limiting Brsak 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 SAFER /GESTR Automation Error (Note 3)
APCT =+30*F B
CURRENT L OCA MODEL ASSESSMENTS None NET PCT:
PCT = 1300*F Page 2 of 7
Attachmsnt 3 LaSalle Unit 210 CFR 50.46 Report (SPC Fuel)
PLANT NAME:
LaSalle Unit 2 ECCS EVALUATION MODEL:
EXEM BWR Evaluation Model REPORT REVISION DATE:
1/9/98 CURRENT OPERATING CYCLE:
8 (upon staitup)
ANALYSIS OF RECORD Evaluation Model Methodology:
Advanced Nuctsar Fuels Corporation Methodology for Bolling Water Reactors EXF.M BWR Evaluation Model, ANF-91048(P)(A),
January,1993.
Calculation:
LaSalle LOCA ECCS Analysis MAPLHGR Limits for ATRIUM -9L Fuel, EMF-96-153(P),
o Revision 2, March 5,1997 (Notes 2, 4, 5 and 6) and iOCA Break Spectrum Analysis for LaSalle Units 1 and 2, EMF-96-152(P), August,1996.
(Notes 2,4 and 6)
Fuel:
ATRIUM -9B Liiniting Single Failure:
HPCS Diesel Generator D
Limiting Break Size and Location:
Discharg side 0.5 ft' Recirculation Line Break Reference PCT:
PCT = 2129'F MARGIN ALLOCATION A.
PRIOR LOCA MOOEL ASSESSMENT 3 None (Note 6)
B.
CURRENT LOCA MODEL ASSESSMENTS Fuel Pellet Grain Size (Note 7)
APCT = 0*F HUXY Claddin;; Strain Error (Note 8)
APCT = -100*F NET PCT: -
PCT = 2029'F Page 3 of 7-l
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 FCT 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 ccolant 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 3HD 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 preciselv the BHd, evaluate the PCT impact of the recirculation line break LOCA including it is believed that the impcct is less than 10 F. Comed determined that inis 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 j
the increased minimum makeup flow is well within the margins available in the ECCS systems. The minirnum makeup flow corresponds to that necessar makeuo for decay heat and 'or system leakages such as the BHD flow path. y to SPC has cc,nservatively incorporated the effects of the BHD into the LaSalle LOCA analysis for ATRIUM -0B 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 (PUTincrease)
In June of 1996, GE re 3orted an error to the USNRC for some applications of the GE LOCA Evaluation Vlo el SAFER /GESTR. It was determined that in some d
analyses an algorithm und to compute the number of fuel rods in a BWR lattice was incorrectly specified. As a result, LOCA input prepared with the automation procecs may have included incorrect c'ata. This error had impact on fuel designs containing a large water rod and analyses where the input generation was autoniated Calculations performed to assess the signihcance of this error indicate that the im aact on the calcu!ated peak cladding temperature is less than 30 F. GE informsc Comed on Septomber 26,1996 that this error applies to the GE analysis for LaSalle Units 1 and 2.
Page 4 of 7 l
1
Y 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 methodolo y.
As a result of using this methodology, SPC calculated a different limitin break size and location than the previous GE analysis. The change in the limiti break and location is a result of a
the SPC methodology and it is not due to the use of the SPC
-9B fuel. SPC has demonstrated the hydraulic compatibility of the ATRIUM -9B and GE fuel and concluded that the mixed core effects have a negl'gible impact on the PCT calculation. Therefore, the GE PCT calculation for i
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 Setpoir.t 4
For the LaSalle LOCA-ECOS 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 additionel PCT margin to the 10 CFR 50.46 acceptance criteria.
A setpoint error analysis performed by Comed showed 14" 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 ebtain 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 received NRC approval for their revised LOCA Evaluation Mcdel. The revised LOCA Evaluation Model corrects the overly conservative modeling of the i
BWR Jet aump. 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 1 setpoint assumed in the current SPC LOCA analysis. Comed plans to revise the LOCA analysis using the new methodology as soon as analysis can be scheduled with SPC.
Page 5 of 7
1 LaSalle Units 1 and 2 PCT Assessment Notes i
- 6. Prior Assessments for SPC A previous 10 CFR 50.46 report for the ATRIUM -9B fuel was docketed la the i
March 21,1997 letter, "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".
This previous 10 CFR 50.46 report documented assessments for the HUXY time step error, the HUXY system pressure error and the changes for the supplemental reflood requirement. Since that report, these changes have been incorporated into a revised MAPLHGR analys.s.
The MAPLHGR analysis is a heatup calculation for the limiting break size and single failure and produces the Analysis of Record (AOR) PCT. Therefore, the SPC Reference PCT includee the PCT impact from these previous model assessments. SPC also concluded that the HUXY time step error, the HUXY system pressure error and the changes for the supplemental reflood requirement did not change the Break Spectrum report conclusion for the limiting break size and single faifure.
- 7. Fuel Grain Size (no PCT impact)
In October of 1997, SPC informed Comed of an inconsistency between fuel grain size production and the grain size asst aed in the mechanical analyses.
The analysis of record used a pellet Crain size of 16 microns, however, the most recent fuel productions have averaged values down to about 10 microns. SPC performed an evaluation for the impact on the LOCA analysis and concluded that the difference in fuel grain size will have no impact on the analysis of record.
The fuel grain size can have a small affect on the intemal rod pressure and stored energy at higher exposures. A smaller grain size would reduce the stored energy. Howeve
' ce the limiting PCT cccurs at very low exposures, there is i
no impact on the anaysis of record.
- 8. HUXY Claddina Strain Error (PCT decr9ase)
In December of 1997, SPC reaorted an error on the application of the HUXY code. The HUXY code is usec to perform heatup calculations during the entire LOCA and yields peak cladding.er..perature and local oxidation at the axial plane of interest. Calculations are made in HUXY to predict the rupture or failure temperature of the fuel cladding. One of the inputs to the cladding rupture temperature calculation is the intemal rod pressure.
Prior to rupture, the cladding experiences strain and the gas volume changes. A change in the intemal dimensions of the cladding affects the internal gas pressure and consequently the predicted rupture temperature.
It has recently-been determined that the instantaneous strain is used in the calculation instead of the maximum strain. The instantaneous strain differs from the maximum strain in that it can fluctuate up and down based on the instantaneous stress. This fluctuation causes the rod intemal pressure and the rod failure temperature to oscillate. This had the affect of prematurely predicting rod failures. SPC has determined that it is more appropriate to use the maximum strain for the roa failure temperature calculation. Using the maximum strain assumes the strain Page 6 of 7
LaSalle Units 1 and 2 PCT Assessment Notes plastically deforms the cladding. Using the maximum strain is consistent with the strain caIculations used in the calculation for the rod dimensions to determine the radiation heat transfer view factors SPC has estimated the impact of the HUXY strain error to be -100*F.
4 Page 7 of 7
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