W3F1-2014-0024, 10CFR 50.46 Significant Change Report for the Waterford 3 ECCS Performance Analysis Due to Implementation of the Replacement Steam Generators
| ML14085A465 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/25/2014 |
| From: | Jarrell J Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| W3F1-2014-0024 | |
| Download: ML14085A465 (16) | |
Text
Entergy Operations, Inc.
1 7265 River Road Killona, LA 70057-3093
,._l--w Tel 504 739 6685 Fax 504 739 6698 jjarrelentergy.com John P Jarrell Regulatory Assurance Manager Waterlord 3 W3FI-2014-0024 10 CFR 50.46 March 25, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
SUBJECT:
I 0 CFR 50.46 Significant Change Report for the Waterford 3 Emergency Core Cooling System Performance Analysis Due to Implementation of the Replacement Steam Generators Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38 REFERENCES I CENPD-1 32, Supplement 4-P-A, Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model, March 2001.
- 2. CENPD-1 37, Supplement 2-P-A, Calculative Methods for the ABB CE Small Break LOCA Evaluation Model, April 1998.
Dear Sir or Madam:
The analyses for the LBLOCA and SBLQCA ECCS performance were re-analyzed for the implementation of Waterford 3 Replacement Steam Generators (RSG). The RSGs were installed in refueling outage (RFI8), and the analyses became effective after entry into Mode I on January 17, 2013 at 14:02. Contrary to the requirements of 10 CFR 50.46(a)(3)(ii) for reporting, a 30-day report that had been prepared was not forwarded to the NRC. This has been entered into the Waterford Steam Electric Station, Unit 3 (Waterford 3) Corrective Action Program in condition report CR-WF3-201 4-00955.
This letter is submitted pursuant to 10 CFR 50.46(a)(3)(ii) to provide notification of a significant change to the peak cladding temperature of the Large Break Loss-of-Coolant Accident (LBLOCA) and Small Break Loss-of-Coolant Accident (SBLOCA) Emergency Core Cooling System (ECCS) performance analyses for Waterford 3.
As noted above, the analyses for the LBLOCA and SBLOCA ECCS performance had been re-analyzed for the implementation of Waterford 3 Replacement Steam Generators (RSG).
In addition, the new analyses incorporated an allowance for up to 10% steam generator tube plugging. The re-analyses were performed using the latest NRC accepted versions of the Westinghouse evaluation models for Combustion Engineering designed pressurized water reactors (References I and 2). The basis for this report is not associated with any model or calculational errors.
W3FI -2014-0024 Page 2 ance criteria of I 0 The results of the new LBLOCA analysis conforms to the ECCS accept CFR 50.46(b) as discussed in the attachment to this letter. Becaus e the sums of the absolute magnitudes of the changes in peak cladding temper ature (POT) associated with the tion), the changes changes implemented in the new analyses are greater than 50°F (reduc for the SBLOOA qualify as significant as defined in 10 CFR 50.46 (a)(3)(i). The POT change for reporting as a analysis does not specifically meetthe requirements of 10 CFR 50.46(b) this criterion, the significant change. However, since it results in a POT that approaches reporte d. As noted above, these change in POT for the SBLOOA analysis is also being RSG refueli ng outage (RFI 8) analyses became effective after entry into Mode I from the es were incorporated into which was completed in January 201 3. The details of these analys in compliance the latest revision of the Waterford 3 Updated Final Safety Analysis Report with 10 OFR 50.71(e).
additional This letter contains no new commitments. If you have any questions or require information, please contact me at 504-739-6685.
Sin
Attachment:
10 CFR 50.46 Significant Ohange Report for Ohanges to the Waterford 3 ECOS Performance Analysis
W3FI -2014-0024 Page 3 cc: Mr. Mark L. Dapas, Regional Administrator U. S. NRC, Region IV RidsRgn4MailCenter@nrC.g0V NRC Project Manager for Waterlord 3 Alan .Wang©nrc.gOV Michael.Orenak©nrc.gOV NRC Resident Inspectors for Waterford 3 Marlone. Davisnrc.goV Chris Speernrc.goV
Attachment to W3FI -2014-0024 10 CFR 50.46 Significant Change Report for Changes to the Waterford 3 ECCS Performance Analysis
Attachment to W3FI -2014-0024 Pagelofl2 I NTRODUCTION the This significant change report is provided for Waterford 3 in accordance with requirements of I 0 CFR 50.46(a)(3)(ii) (Reference 1 ) for reporti ng:
a model that (1) changes in an acceptable evaluation model or the application of such affects the temperature calculation and Cooling System (2) the estimated effect of the changes on the limiting Emergency Core (ECCS) analysis.
s described herein Because the effects on the Peak Clad Temperature (PCT) of the change CFR 50.46(a)(3)(i) are greater than 50°F, the changes qualify as significant as defined in 10 are being and, consequently, are provided in this significant change report. No errors reported.
of Coolant Emergency Core Cooling System performance for the Large Break Loss nt (SBLO CA) has been re Accident (LBLOCA) and the Small Break Loss of Coolant Accide ccepted analyzed for Waterford 3. The analyses were performed with the latest NRC-a stion Engineering versions of the Westinghouse Appendix K evaluation models for Combu the implementation designed Pressurized Water Reactors (PWRs). The analyses modeled steam generator of Replacement Steam Generators (RSG) with allowance for up to 10%
tube plugging (SGTP).
do not provide an The new LBLOCA and SBLOCA analyses are not assessments (i.e., they
, they are estimate of the effect of the changes on the limiting ECCS analysis). Rather ble to complete re-analyses that use acceptable evaluation models that are applica the new analys es and their compl iance with 10 Waterford 3. A summary description of CFR 50.46 is provided below.
LBLOCA ECCS PERFORMANCE ANALYSIS LBLOCA Evaluation Model 1999 Evaluation The new LBLOCA ECCS performance analysis was performed with the Model (EM) version of the Westinghouse LBLOCA evalua tion model for Combustion ence 2).
Engineering designed PWRs (Reference 3), an Appendix K EM (Refer entation of Additionally, the analysis used methodology supplements for the implem CE I 6x1 6 Next TM and Optimized ZIRLOTM cladding types, the implementation of ZIRLO al Steam Cooling Generation Fuel (NGF) assemblies, and the use of the I 999 EM Option by the NRC for Model. Each of these I 999 EM submittals has been generically accepted elements of the licensing applications for Combustion Engineering designed PWRs. These LBLOCA EM are described in the subsections below.
6.9.1 .1 1 the in accordance with the requirements of Waterlord 3 Technical Specification Limits Report (COLR) 1999 EM topical reports are listed in Section III of the Core Operating operating limits in as approved analytical methodologies that can be used to determine core the COLR.
Attachment to W3FI -2014-0024 Page 2 of 12 nstraints imposed by the Safety The new LBLOCA analysis complies with the limitations/co as well as the applicable Evaluation Reports (SERs) for the I 999 EM topical reports s of the LBLOCA evaluation limitations/constraints imposed by the SERs for earlier version model.
1999 EM (CENPD-132, Suprlement 4-P-A)
Performance in CE plants is the The Westinghouse Appendix K Evaluation Model for ECCS 3). The SERs documenting 1999 Evaluation Model (1999 EM) for LBLOCA (Reference ed in Refere nces 5, 6, 7, and 23. The NRC acceptance of the evaluation model are provid for analys is of ZIRLOTM cladding I 999 EM for LBLOCA is augmented by CENPD-404-P-A and by Addendum I to CENPD (Reference I 3 and approved by NRC in Reference 24),
ZIRLOTM cladding (Reference 1 4 and approved by NRC 404-P-A for analysis of Optimized WCAP-1 6072-P-A (Reference I 5 in Reference 1 9). Also, the I 999 EM is supplemented by n of ZrB2 IFBA fuel assembly and approved by NRC in Reference 25) for implementatio coolin g heat transfer component model designs. The 1999 EM includes an optional steam heat transfer effects as for less than 1 in/sec core reflood that includes spacer grid nce 4. The implementation of documented in Reference 12 and approved by NRC in Refere ented in Reference 1 1 and CE 1 6x1 6 NGF into the I 999 EM methodology is docum approved by NRC in Reference 16.
CE 16x16 NGF (WCAP-16500-P) are documented in The methodologies for licensing CE 16x16 NGF assemblies 16 x 16 Next Generation Fuel Westinghouse Topical Report WCAP-16500-P, titled CE was approved by the NRC in Core Reference Report (Reference 1 1 ). WCAP-1 6500-P ECCS performance methods Reference I 6. Section 5.2 of Reference I I documents the The final SER for WCAP-1 6500-P suitable for use to analyze the implementation of NGF.
with these Limitations and Conditions contains 10 Limitations and Conditions. Compliance in Reference 17, and the ECCS for implementation of NGF in Waterford 3 was documented Performance analysis was documented in Reference I 8.
Optimized ZIRLOTM (CENPD-404-P-A Addendum I)
ZIRLOTM, an advanced cladding alloy. The The CE 16x16 NGF design utilizes Optimized implementation of Optimized ZIRLO TM in CE plants is documented in Reference 14 and SER Limitations and Conditions approved by the NRC in Reference I 9. As required by the codes have been updated to in Reference 19, the ECCS performance analysis computer s detailed in the topical report.
include the Optimized ZIRLOTM cladding property change lies requires a cladding The use of Optimized ZIRLOTM cladding in the NGF assemb 10 CFR Part 50, Appendix K, which exemption from the requirements of 10 CFR 50.46 and by NRC in Reference 21.
was submitted to the NRC in Reference 20 and accepted ment 4-P-A Addendum I -P)
I 999 EM Optional Steam Cooling Model (CENPD-1 32 Supple LOCA Evaluation Model, Calculative Methods for the CE Nuclear Power Large Break g Model for Less Than 1 in/sec Improvement to I 999 Large Break LOCA EM Steam Coolin 1-P was approved by the NRC Core Reflood, CENPD-132, Supplement 4-P-A, Addendum
Attachment to W3FI -2014-0024 Page 3 of 12 LBLOCA analysis in Reference 4. For the implementation of RSGs at Waterford 3, the r effects.
credited the use of the modified model including spacer grid heat transfe 4, and 5 are applicable The conclusion of the SER stated that Limitations and Conditions 3, Limita tions and Conditions to CE I 6 x I 6 NGF design fuel assemblies. The first two dum 1 -P are for fuel included in the SER for CENPD-1 32, Supplement 4-P-A, Adden Conditions and the designs other than CE 16 x 16 NGF. The applicable Limitations and supplementary ECCS means of satisfying them were documented in Reference I 7, and the Performance analysis was documented in Reference 22.
Fuel Design Changes no fuel design changes For ECCS Performance modeling using the 1999 EM, there were implemented for the new analysis with RSGs.
Plant Parameter Changes The RSGs were The new LBLOCA analysis includes the design impact changes for RSGs.
is inputs.
implemented assuming an allowance for up to 10% SGTP in the analys on (i.e., greater than 50 Replacement of the Waterlord 3 RSGs leads to a significant reducti
°F) in predicted PCT.
Results and Conclusion of the New LBLOCA Analysis coolant pump The new LBLOCA analysis analyzed a break spectrum of four reactor to a 0.4 double-ended discharge leg breaks ranging in size from a full double-ended break g single failure of ECCS break. The analysis included a study to determine the most limitin diesel generator, and equipment. The study analyzed no failure, failure of an emergency injection pump. The analysis also failure of a low pressure and a high pressure safety ons in initial safety injection tank included studies that investigated the impact of variati um claddin g oxidation.
conditions and refueling water tank temperature on PCT and maxim ions by fuel rod condit The analysis also included studies that investigated the limiting initial le absorb fuel er performing burnup dependent calculations for both U02 and ZrB2 burnab rods for a full core of CE 1 6x1 6 NGF assemblies.
new LBLOCA Table I compares several important inputs used in the current and the A more detailed analyses. Table 2 compares important results from the two analys es.
t the results of the description of the new analysis, including tables and figures that presen al steam cooling break spectrum analysis as well as results confirming use of the option Analys Report in is model, has been incorporated into the Waterford 3 Updated Final Safety accordance with 10 CFR 50.71(e).
new LBLOCA analysis As shown in Table 2, the net change in PCT that resulted from the the net change in maximum implementing RSGs is a minus 74°F. Also shown in Table 2, is implementing RSGs is local cladding oxidation percentage from the new LBLOCA analys a minus 3.9%.
m to the acceptance As summarized below, the results of the new LBLOCA analysis confor criteria of 10 CFR 50.46(b).
Attachment to W3FI -2014-0024 Page 4 of 12 Parameter Criterion Result Peak Cladding Temperature 2200°F 2092°F Maximum Cladding Oxidation 17 % 13.0 %
Maximum Core-Wide Oxidation i % <1 %
Coolable Geometry Yes Yes tor tubes The results are applicable to Waterford 3 with RSGs and up to I 0% steam genera Heat Genera tion Rate (PLHG R) of 1 2.9 kW/ft plugged and for operation at a Peak Linear and a core power of 3735 MWt (rated core power of 3716 MWt with a 0.5% power measurement uncertainty).
licensing The new LBLOCA analysis uses the 1999 EM, which is accepted by the NRC for applications for Combustion Engineering designed PWRs such as Waterlord 3. The The analysis complies with the limitations/constraints imposed by all applicable SERs.
are either applica ble to or bound the current analysis uses values for plant design data that have ongoin g proces ses that configuration of Waterford 3. Entergy and Westinghouse ed by ensure that the as operated plant values for PCT-sensitive parameters remain bound ce analys is to the values used in the analysis. The new analysis will be used as the referen its evaluate the impact on PCT of future changes to or errors in the I 999 EM and application to Waterlord 3.
SBLOCA ECCS PERFORMANCE ANALYSIS SBLOCA Evaluation Model S2M or The small break LOCA analysis used the Supplement 2 version (referred to as the break LOCA NRC-a ccepte d Evalua tion Supplement 2 Model) of the Westinghouse small ed plants. The SERs docum enting Model (Reference 8) for Combustion Engineering design nces 5, 9, and I 0. The NRC acceptance of the evaluation model are contained in Refere es using methodology for modeling the NGF assembly design in ECCS Performance Analys This the 52M is contained in the CE 1 6x1 6 NGF Core Topical Report (Reference I 1 ).
ions on the report was approved by the NRC and contains the SER constraints and limitat The S2M for SBLO CA is augme nted by application of the S2M for the analysis of NGF.
ZIRLOTM claddin g (Refer ence I 3) and by Adden dum I to CENPD-404-P-A for analysis of ZIRLOTM cladding (Refer ence 14). Also, the CENPD-404-P-A for analysis of Optimized IFBA fuel assemb ly S2M is supplemented by WCAP-16072-P-A for implementation of ZrB2 designs (Reference 15).
Attachment to W3FI -2014-0024 Page 5 of 12 Specification 69. I I I the In accordance with the requirements of Waterford 3 Technical ed analytical 52M topical reports are listed in Section Ill of the COLR as approv in the COLR.
methodologies that can be used to determine core operating limits ts imposed by the SERs The new SBLOCA analysis complies with the limitations/constrain ions/co nstrain ts imposed by the for the S2M topical reports as well as the applicable limitat SERs for earlier versions of the SBLOCA evaluation model.
Fuel Design Changes no fuel design changes For ECCS Performance modeling using the S2M, there were implemented for the new analysis with RSGs.
Plant Parameter Changes s for RSGs. The RSGs were The new SBLOCA analysis includes the design impact change SGTP in the analysis inputs.
implemented assuming an allowance for up to 10%
Results and Conclusion of the New SBLOCA Analysis reactor coolant pump The new SBLOCA analysis analyzed a break spectrum of three
, 0.05 ft, and 0.06 ft 2 . The 0.05 ft 2 /PD break was 2
discharge leg breaks, namely, 0.04 ft 2 is was performed using the determined to be the limiting small break LOCA. The analys failure of a direct current (DC) bus as the most limitin g single failure of the ECCS. A DC tor that would cause the bus failure would prevent startup of an emergency diesel genera pressure safety injection loss of a High Pressure Safety Injection (HPSI) pump and a low being available to cool the (LPSI) pump, and results in a minimum of safety injection water heric dump valves core. The analysis credits operation of the steam generator atmosp ent. They are model ed in automatic mode with (ADVs). The ADVs are safety grade equipm of a DC bus, which an opening pressure of I 040 psia. The most limiting single failure DC power to an ADV controller.
prevents start up of a diesel generator, results in loss of for contro l of secondary side Thus only one of the two ADVs (one ADV per SG) is available rod condit ions that result in the pressure. The SBLOCA analysis was performed for the fuel ed the analysis of both maximum initial stored energy in the fuel. The calculations includ assemblies.
U02 and ZrB2 burnable absorber fuel rods for a full core of NGF t and the new SBLOCA Table 3 compares several important inputs used in the curren es. A more detailed analyses. Table 4 compares important results from the two analys present the results of the description of the new analysis, including tables and figures that ord 3 Updated Final Safety break spectrum analysis, have been incorporated into the Waterf Analysis Report in accordance with 10 CFR 50.71(e).
the new SBLOCA analysis As shown in Table 4, the net change in PCT that resulted from 4, the net change in maximum implementing RSGs is minus 48 °F. Also shown in Table is implementing RSGs is local cladding oxidation percentage from the new SBLOCA analys minus 3.1%.
is conform to the acceptance As summarized below, the results of the new SBLOCA analys criteria of 10 CFR 50.46(b).
Attachment to W3FI -2014-0024 Page 6 of 12 Parameter Criterion Result Peak Cladding Temperature 2200°F 1925°F 11.2 %
Maximum Cladding Oxidation 17 %
Maximum Core-Wide Oxidation i % <0.65 %
Yes Yes Coolable Geometry I 0% steam generator tubes The results are applicable to Wateriord 3 with RSGs and up to tion Rate (PLHGR) of I 32 kW/ft plugged and for operation at a Peak Linear Heat Genera with a 0.5% power and a core power of 3735 MWt (rated core power of 371 6 MWt measurement uncertainty).
the NRC for licensing The new SBLOCA analysis uses the S2M, which is accepted by Waterf ord 3. The analys is complies with the applications for CE designed PWRs such as The analys is uses values for plant limitations/constraints imposed by all applicable SERs.
t config uration of Waterford 3.
design data that are either applicable to or bound the curren that the as operated plant Entergy and Westinghouse have ongoing processes that ensure used in the analysis.
values for PCT-sensitive parameters remain bounded by the values te the impact on PCT of The new analysis will be used as the reference analysis to evalua ord 3.
future changes to or errors in the S2M and its application to Waterf Summary of Compliance with 10 CFR 50.46 50.46 as follows:
The new LBLOCA and SBLOCA analyses comply with I 0 CFR s and included
. The analyses were performed with acceptable evaluation model CA were analyzed sensitivity studies that assured the limiting LBLOCA and SBLO
[10 CFR 5046(a)(1)(i)j.
to the ECCS
. The results of the new LBLOCA and SBLOCA analyses conform acceptance criteria [10 CFR 50.46(b)].
tion of the
. This report provides NRC with notification of the change in the applica
[1 0 CFR evaluation models and their effect on the limiting ECCS analyses 50.46(a)(3)(ii)].
n of RSGs with allowance The new LBLOCA and SBLOCA analyses for the implementatio es-of-record) for for up to 10% SGTP constitute new licensing basis analyses (analys entry into Mode I from the RSG Waterford 3. These analyses became effective after 201 3. The new analyses will be refueling outage (RF1 8) which was completed in January future change s to or used as the reference analyses to evaluate the impact on PCT of ord 3.
errors in the 1999 EM and the S2M and their application to Waterf
Attachment to W3FI -2014-0024 Page 7 of 12 REFERENCES tance Criteria for I . Code of Federal Regulations, Title I 0, Part 50, Section 50.46, Accep rs.
Emergency Core Cooling Systems for Light Water Nuclear Power Reacto Evaluation Models.
- 3. CENPD-132P, Calculative Methods for the C-E Large Break August 1974.
Break LOCA CENPD-1 32P, Supplement I Calculational Methods for the C-E Large Evaluation Model, February 1975.
Large Break LOCA CENPD-1 32-P, Supplement 2-P, Calculational Methods for the C-E Evaluation Model, July 1975.
Large Break LOCA CENPD-1 32, Supplement 3-P-A, Calculative Methods for the C-E June 1985.
Evaluation Model for the Analysis of C-E and W Designed NSSS, Nuclear Power Large CENPD-1 32, Supplement 4-P-A, Calculative Methods for the CE Break LOCA Evaluation Model, March 2001.
Evaluation for
- 4. NRC Letter to Westinghouse dated June 27, 2007, Final Safety l Report (TR) CENPD-1 32 Westinghouse Electric Company (Westinghouse) Topica
[Combustion Supplement 4-P-A, Addendum I -P, Calculative Methods for the CE Improvement to Engineering) Nuclear Power Large Break LOCA Evaluation Model in/sec Core Reflood 1999 Large Break LOCA EM Steam Cooling Model for Less than I
- 5. O.D. Parr (NRC) to F.M. Stern (C-E), June 13, 1975.
- 6. O.D. Parr (NRC) to A.E. Scherer (C-E), December 9, 1975.
of Combustion
- 7. D.M. Crutchfield (NRC) to A.E. Scherer (C-E), Safety Evaluation tion Model and Accep tance for Referencing of Engineering ECCS Large Break Evalua Related Licensing Topical Reports, July 31 , I 986.
LOCA Evaluation Model,
- 8. CENPD-137P, Calculative Methods for the C-E Small Break August 1974.
Break LOCA CENPD-137, Supplement 1-P, Calculative Methods for the C-E Small Evaluation Model, January 1977.
CE Small Break CENPD-1 37, Supplement 2-P-A, Calculative Methods for the ABB LOCA Evaluation Model, April 1998.
l s CENPD-1 33,
- 9. K. Kniel (NRC) to A.E. Scherer (C-E), Evaluation of Topica Report ber 27, 1977.
Supplement 3-P and CENPD-137, Supplement 1-P, Septem ncing of the Topical
- 10. T. H. Essig (NRC) to I. C. Rickard (ABB), Acceptance for Refere ative Metho ds for the C-E Small Break Report CENPD-1 37(P), Supplement 2, Calcul LOCA Evaluation Model (TAC No. M95687), Decem ber 16, 1997.
Reference Report, I I WCAP-1 6500-P-A, Rev. 0, CE 1 6x1 6 Next Generation Fuel Core August 2007.
Methods for the CE 12.CENPD-132-P-A Supplement 4-P-A Addendum 1-P-A, Calculative 1999 Large Break Nuclear Power Large Break LOCA Evaluation Model, Improvement to I in/sec Core Refloo d, August 2007.
LOCA EM Steam Cooling Model for Less Than
Attachment to W3FI -2014-0024 Page 8 of 12 ZIRLOTM Cladding Material in CE Nuclear 1 3. CENPD-404-P-A, Rev. 0, Implementation of Power Fuel Assembly Designs, November 2001.
Addendum I to WCAP-1 2610-
- 14. WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A, ZIRLO TM, July 2006.
P-A and CENPD-404-P-A Optimized Diboride Burnable Absorber
- 15. WCAP-16072-P-A, Rev. 0, Implementation ofZirconium t 2004.
Coatings in CE Nuclear Power Fuel Assembly Designs, Augus se), Final Safety
- 16. Letter from H. K. Nieh (NRC) to J. A. Gresham (Westinghou ny Topica l Report WCAP-1 6500-P, Evaluation for Westinghouse Electric Compa Refere nce Report (TAC No.
Revision 0, CE 16x16 Next Generation Fuel Core MD0560), July 30, 2007.
NPF-38-271 to Support Next
- 17. Entergy letter to the NRC, License Amendment Request (ADAM S No.: ML072180042)
Generation Fuel, August 2, 2007 (W3F1-2007-0037).
Performance Analysis to
- 18. Entergy letter to the NRC, Emergency Core Cooling System 0038). (ADAMS No.:
Support Next Generation Fuel, August 9, 2007 (W3FI-2007-ML072250389) nghouse), Final Safety I 9. Letter from H. N. Berkow (NRC) to J. A. Gresham (Westi Report WCAP -12610 -P-A and CENPD-404-P-A, Evaluation for Addendum 1 to Topical Optimized ZIRLOTM (TAC No. MB8041), June 10, 2005.
to Allow the Use of Optimized
- 20. Entergy letter to the NRC, License Amendment Request ZIRLOTM Fuel Rod Cladding, April 24, 2007 (W3FI-2007-0020). (ADAMS No.:
MLO71 160348) ions (EOI), Waterford Steam 21 Letter from N. Kalyanam (NRC) to Vice President, Operat in 10 CFR 50.46 and Electric Station, Unit 3 Exemption from Specific Requirements ized ZIRLOTM Fuel Rod from Appendix K to 10 CFR Part 50, to Allow the Use of Optim S No. : ML080380002)
Cladding Material (TAC No. MD5426), March 1 1 2008. (ADAM ance Analysis Submitted
- 22. Entergy letter to the NRC, Supplement to the ECCS Perform Cooling Model in Support of Next Generation Fuel I 999 EM Optional Steam (ADAM S No.: ML072820400)
Justification, October 4, 2007 (W3FI-2007-0045).
dson (Westinghouse) dated
- 23. Letter from Mr. S. A. Richards (NRC) to Mr. P. W. Richar D-1 32, Supplement 4, December 1 5, 2000, Safety Evaluation of Topical Report CENP Large Break LOCA Revision I Calculative Methods for the CE Nuclear Power Evaluation Model (TAC No. MA5660).
dson (Westinghouse) dated
- 24. Letter from Mr. S. A. Richards (NRC) to Mr. P. W. Richar D-404-P, Revision 0, September 12, 2001 Safety Evaluation of Topical Report CENP TM Material Cladding in CE Nuclea r Power Fuel Assembly Implementation of ZIRLO Designs (TAC No. MB1035).
(Westinghouse), dated May
- 25. Letter from Mr. H. N. Berkow (NRC) to Mr. J. A. Gresham 6, 2004, Final Safety Evaluation for Topical Report WCAP -16072-P, Revision 00, er Coatin gs in CE Nuclear Implementation of Zirconium Diboride Burnable Absorb Power Fuel Assembly Designs, (TAC No. MB8721).
Attachment to W3FI -2014-0024 Page 9 of 12 Table I Waterlord 3 LBLOCA ECCS Performance Analysis Comparison of Important Input Parameters Current Analysis New Analysis Parameter 1999 EM 1999 EM LBLOCA Evaluation Model 3735 3735 Core Power Level, MWt (including power measurement uncertainty) 12.9 12.9 Peak Linear Heat Generation Rate (PLHGR) of the Hot Rod, kW/ft 1 2.0 12.0 PLHGR of the Average Rod in Assembly with Hot Rod, kW/ft 1 48.Oxl 06 1 48.Oxl 06 Reactor Coolant System Flow Rate, Ibm/hr I 44. 1 5x1 06 1 44. 1 5x1 06 Core Flow Rate, Ibm/hr 2250 2250 RCS Pressure, psia 533 533 Cold Leg Temperature, °F 598.7 598.7 Hot Leg Temperature, °F 1870 (OSG) 897 (RSG)
Number of Plugged Tubes per Steam Generator CE 16x16 NGF CE 16x16 NGF Fuel Assembly Design Optimized ZIRLOTM Optimized ZIRLOTM Fuel Rod Cladding Type
)
(a) Applicable to Cycle I 6 and later for Original Steam Generators (OSGs
Attachment to W3F1 -2014-0024 Page 10 of 12 Table 2 Waterford 3 LBLOCA ECCS Performance Analysis Comparison of Important Results Current Analysis New Analysis Parameter Limiting Cases for Peak Cladding Temperature I .0 DEG/PD 1 .0 DEG/PD Limiting Break Size 2166 2092 Peak Cladding Temperature, °F 16.9 12.8 Maximum Cladding Oxidation, %
<1 .0 <1.0 Maximum Core-Wide Cladding Oxidation, %
Limiting Cases for Maximum Cladding Oxidation I .0 DEG/PD 0.8 DEG/PD Limiting Break Size 2155 2091 Peak Cladding Temperature, °F 16.9 13.0 Maximum Cladding Oxidation, %
<1 .0 <1 .0 Maximum Core-Wide Cladding Oxidation, %
(a) OSGs with 1870 plugged tubes per steam generator (b) RSGs with 897 plugged tubes per steam generator (c) DEG/PD = Double-Ended Guillotine Break in Pump Discharge Leg
Attachment to W3FI -2014-0024 Page 11 of 12 Table 3 Waterford 3 SBLOCA ECCS Performance Analysis Comparison of Important Input Parameters Current Analysis New Analysis Parameter S2M S2M SBLOCA Evaluation Model 3735 3735 Core Power Level, MWt (including power measurement uncertainty) 13.2 13.2 Peak Linear Heat Generation Rate, kW/ft 6
148.0x 10 6 148.0x 10 RCS Flow Rate, Ibm/hr 6
144.15 x10 6 144.15 x10 Core Flow Rate, Ibm/hr 2250 2250 RCS Pressure, psia 552 552 Cold Leg Temperature, °F 615.5 615.5 HotLegTemperature,°F I 870 (OSG) 897 (RSG)
Number of Plugged Tubes per Steam Generator CE 16x16 NGF CE 16x16 NGF Fuel Assembly Design Optimized ZIRLOTM Optimized ZIRLOTM Fuel Rod Cladding Type (a) Applicable to Cycle I 6 and later for Original Steam Generators
Attachment to W3FI -2014-0024 Page 1 2 of I 2 Table 4 Waterford 3 SBLOCA ECCS Performance Analysis Comparison of Important Results Current Analysis New Analysis Parameter 0.055 0.05 Limiting Break Size, ft 2
1973 1925 Peak Cladding Temperature, °F 14.3 11.2 Maximum Cladding Oxidation, %
<0.80 <0.65 Maximum Core-Wide Cladding Oxidation, %
(a) OSGs wfth I 870 plugged tubes per steam generator (b) RSGs with 897 plugged tubes per steam generator