ML20140E196
| ML20140E196 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/31/1985 |
| From: | Holm J, Jensen S, Ward G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17216A419 | List: |
| References | |
| XN-NF-85-117, XN-NF-85-117-S01, XN-NF-85-117-S1, NUDOCS 8602030201 | |
| Download: ML20140E196 (20) | |
Text
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XN-NF-85-117 SUPPLEMENT 1 ST. LUCIE UNIT 1 REVISED LOCA-ECCS ANALYSIS WITH 15/
STEAM GENERATOR TUBE PtUGGING q BREAK SPECTRUM AND EXPOSURE RESULTS DECEMBER 1985 RICHLAND, WA 99352 l
EXXON \UC_ EAR COV 3A\Y \C.
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XN-NF-85-117 Supplement 1 Issue Date: 12/6/85 ST. LUCIE UNIT 1 REVISED LOCA-ECCS ANALYSIS WITH 15% STEAM GENERATOR TUBE PLUGGING BREAK SPECTRUM AND EXPOSURE RESULTS Contributors: R. C. Gottula J. D. Kahn C. E. Slater (ITI)
Prepared by: /. -
// 27 (-(
'S. E. Jensin BWR Safety Analysis Approved by: f,/ Il!37 fpf J/} S F5Tm, Manager PWR Safety Analysis Concur: $ NN G. N. Ward, Manager s)LkPs-Reload Licensing Approved by: u nOW H. E. Williamson, Manager Licensing and Safety Engineering Concur: ,
H/ Svt- (.u {/f iz-Gy.C JJN. Morgan,ifanager <f gbstomer Services Engineering Approved by: _
/A/f/h' G. L. Ritter, Manager Fuel Engineering and Technical Services .
f NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of comoliance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:
A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
1 XN- NF- F00, 766
l l
i XN-NF-85-117 Supplement 1 1
TABLE OF CONTENTS Section Page Number Number 1
1.0 INTRODUCTION
AND
SUMMARY
6 2.0 LOCA-ECCS ANALYSIS.......................... ..............
7 2.1 BREAK SPECTRUM RESULTS................................
2.2 EXPOSURE ANALYSIS RESULTS............................. 8
3.0 CONCLUSION
S.................... ........................... 12
4.0 REFERENCES
......... ....................... .. .... ..... . 14
ii XN-NF-85-117 Supplement 1 LIST OF TABLES Table Title Page Break Spectrum PCT Results ........................... 4 1.1 1.2 Exposure Study PCT Results ........................... 5 2.1 LOCA Break Spectrum Event Times ...................... 9 2.2 LOCA Break Spectrum Results ................ ......... 10 2.3 LOCA Exposure Study Results 0.8DECLG Break ....... .. 11
l 1
XN-NF-85-ll7 Supplement 1 ST. LUCIE UNIT 1 REVISED LOCA-ECCS ANALYSIS WITH 15% STEAM GENERATOR TUBE PLUGGING BREAK SPECTRUM AND EXPOSURE RESULTS
1.0 INTRODUCTION
AND
SUMMARY
In November 1985, Exxon Nuclear Company (ENC) reported revised LOCA-ECCS analyses for St. Lucie Unit I which corrected errors in previous analyses and increased the level of analyzed steam generator tube plugging to 15%(1). The referenced document reported the first phase of the analysis and provided axially dependent linear heat generation rate (LHGR) limits for St. Lucie Unit 1 operating with ENC 14x14 fuel for cycle 7 and future cycles. The LHGR limits were established by LOCA-ECCS analyses calculations assuming worst case exposure conditions for the limiting break determined from partial LOCA break spectrum resul ts . This supplement reports the second and final phase of the analysis and provides the completed LOCA-ECCS break spectrum results which confivm the previously established limiting LOCA break and also provides LOCA analysis results at exposure conditions which confirm the worst case exposure assumption. Analysis for Co.mbustion Engineering (CE) fuel is reported in the St. Lucie Unit 1 FSAR. Use of the ENC LHGR limits for both the ENC and CE fuel is conservative for monitoring plant operation.
2 XN-NF-85-ll7 Supplement 1 All analyses were performed with the ENC EXEM/PWR ECCS Evaluation Model(2) according to Appendix K of 10 CFR 50(3). The approved WREM-1 reflood heat transfer correlation (4) was used. The St. Lucie Unit 1 plant was analyzed with the cycle 7 core configuration and the vessel thermal shield removed consistent with the Reference 1 analysis.
The break spectrum analysis consisted of calculations for a spectrum of seven large cold leg break LOCA's: guillotine breaks with discharge coefficients of 1.0, 0.8, 0.6, and 0.4, and split breaks with break areas equal to 1.0, 0.8, and 0.4 times the double-ended pipe area. The peak cladding temperature (PCT) results of the final break spectrum calculations are given in Table 1.1. The results show that the previously identified double-ended cold leg guillotine break with a discharge coefficient of 0.8 (0.8DECLG) is the worst or limiting break LOCA. This is consistent with the Reference 1 results.
The exposure analysis reported in Reference 1 assumed a worst case exposure at the point of maximum stored energy (1.8 MWD /kg peak rod average burnup). To confirm the worst case assumption, limiting break LOCA calculations were performed for two additional exposure cases: 14.0 MWD /kg and 49.1 MWD /kg peak rod average burnups. Peak cladding temperature (PCT) results for the three exposure calculations are given in Table 1.2. The maximum stored energy exposure, as reported in Reference 1, gives the highest PCT and bounds exposure conditions over the life of the ENC 14x14 fuel.
3 XN-NF-85-ll7 Supplemer.t 1 As expected, the results of the analyses contained in this supplement confirm the analysis presented in Reference 1, and support operation of St. Lucie Unit I with up to 15". steam generator tube plugging. Operation of the St. Lucie Unit I reactor with ENC 14x14 fuel at or below the LHGR limits of Reference 1 satisfies the criteria specified by 10 CFR 50.46 of the U. S. Code of Federal Regulations,(3) and assures that the emergency core cooling system for the St. Lucie Unit I reactor will meet the U. S. NRC acceptance criteria for loss-of-Coolant Accident breaks up to and including the double-ended severance of a reactor coolant pipe.
I
4 XN-NF-85-117 Supplement 1 Table 1.1 Break Spectrum PCT Results Large Break LOCA Guillotine Breaks Di scharge Break Peak C1ad Coefficient Area ( ft2) Temperature (OF) 1.0 10.014 2108 0.8 10.014 2188 0.6 10.014 2064 0.4 10.014 1983 Split Breaks Discharge Break Peak C1ad Coefficient Area (ft2) Temperature (OF) 1.0 10.014 1961 1.0 8.011 1911 1.0 4.006 1746 Axial profile 15.0 kw/ft at 0.6 x/1 Maximum Stored Energy Exposure
5 XN -NF-85-117 Supplement 1 Table 1.2 Exposure Study PCT Results 0.80ECLG Break Peak Rod Average Burnup (MWD /kg) 1.8 14.0 49.1 Peak C1 adding Temperature ( F) 2188 2077 1730 Axial Profile 15.0 kw/ft at 0.6 x/1
6 XN-NF-85-ll7 Supplement 1 2.0 LOCA-ECCS ANALYSIS The LOCA break spectrum and exposure calculations used the EXEM/PWR ECCS evaluation model(2) as described in Reference 1. St. Lucie Unit 1 system input with the vessel thermal shield removed was identical to the 0.8DECLG case reported in Reference 1 except for the required changes to represent the various breaks in the spectrum. Exposure cases were identical to the 0.8DECLG case except for the RODEX2 values calculated for the desired exposure points. All the break spectrum and exposure calculations were performed using the axial power profile with a peak at 15.0 kw/ft at 0.6 of core height as reported in Reference 1. An average steam generator tube plugging of 15% was assumed with the broken loop having 17% of the steam generator tubes plugged and the intact loop having 13% steam generator tube plugging. This asymmetric steam generator tube plugging is consistent with the Reference 1 analysis.
7 XN-NF-85-117 Supplement 1 2.1 BREAK SPECTRUM ANALYSIS The analysis reported in Reference 1 identified the double-ended guillotine break of the cold leg or reactor vessel inlet line with a discharge coefficient of 0.8 (0.8DECLG) as the worst or limiting break LOCA. Identification of the limiting break was baseJ on break spectrum results run only through the blowdown portion of the LOCA analysis. The preliminary results reported previously used different axial power profiles for the guillotine breaks than for the split breaks. To provide a complete spectrum of large break LOCA results which comply with NRC 10 CFR 50.46 criteria, six additional large break LOCA calculations were performed to completion. The additional six breaks were run using axial power profile and exposure input consistent with the limiting 0.8DECLG results reported in Reference 1. The axial power profile having a maximum LHGR of 15.0 kw/f t at an elevation of 0.6 of the core height was used.
Initial fuel conditions used for the break spectrum were the maximum stored energy exposure input from the Reference 1 limiting 0.8DECLG large break LOCA case.
Calculated LOCA event times for the break spectrum cases are given in Table 2.1. The final LOCA-ECCS results are shown in Tables 1.1 and 2.2. As expected, the O'.8DECLG case which yielded the highest fuel temperatures at the end of blowdown, also gives the highest peak cladding temperature (PCT). The final break spectrum results confirm the limiting break LOCA and the LHGR limits presented in Reference 1.
8 XN-NF-85-ll7 Supplement 1 2.2 EXPOSURE ANALYSIS RESULTS To bound all exposure conditions for cycle 7 and beyond, the ENC LOCA-ECCS analyses reported in Reference 1 was performed at exposure conditions corresponding ta maximum fuel temperature or maximum stored energy as predicted by RODEX2. The calculated exposure having maximum stored energy was near beginning-of-life at 1.8 MWD /kg peak rod average burnup.
To confirm the worst case exposure point, ENC performed limiting break LOCA calculations for two additional frel exposures,14.0 MWD /kg and 49.1 MWD /kg peak rod average burnups. Table 2.3 gives the LOCA-ECCS results for the three exposure cases run. The maximum stored energy exposure (BOL) case has significantly higher initial fuel temperatures than the higher exposure cases and also results in the highest PCT. Thus, use of limits based on the BOL maximum stored energy bounds the fuel conditions over the expected exposure range. The LHGR limits given in Reference 1 are confirmed over the expected exposure range of ENC 14x14 fuel.
I Table 2.1 LOCA Break Spectrisn Event Times Event Ti_me (Sec) 1.0DECLG 0.8DECLG 0.6DECLG 0.4DECLG 1.00ECLS 0.8DECLS 0.4DECLS Event CD=1.0 CD=0.8 CD=0.6 CD=0.4 10.0 FT2 8.0 FT2 4.0 FT2 Start 0.0 0.0 0.0 0.0 0.0 0.0 0.0 l
Break is Fully Open 0.05 0.05 0.05 0.05 0.05 0.05 0.05 Safety Injection Signal 0.82 0.88 1.0 1.21 0.68 0.70 1.07 Pressurizer Empties 9.6 9.6 9.6 9.7 9.6 9.6 9.7 Accunulator Injection, Broken Loop 10.0 12.5 15.5 21.1 10.3 11.7 19.5 Accumulator Injection, Single Intact Loop 16.0 16.6 18.1 22.7 12.3 12.7 19.7 e
Accunulator Injeci on, Double Intact Loop 16.0 16.6 18.1 22.7 12.3 12.7 19.7 End-o f-Bypass 20.95 21.25 22.91 28.07 16.69 17.11 24.18 Safety Injection Flow, SIS 30.82 30.88 31.0 31.21 30.68 30.70 31.07 Start of Reflood 39.41 39.71 41.25 46.26 35.31 35.72 42.26 Acctanulators Empty, Broken Loop 61.05 62.85 65.81 71.45 60.34 61.56 69.98 Accumulators Empty, Single Intact Loop 65.75 66.35 68.16 73.15 62.00 62.36 70.23 mx E?
Accumulators Empty, Double intact Loop 65.85 66.35 68.06 73.15 62.00 62.36 70.23 c, y Peak Clad Temperature is Reached 186.1 183.1 181.5 177.6 176.1 172.0 130.1
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=
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Table 2.2 LOCA Break Spectrum Results l 1.00ECLG 0.80ECLG 0.6DECLG 0.4DECLG 1.0DECLS 0.8DECLS 0.40ECLS Analysis Results CD =1. 0 CD=0.8 CD=0.6 C0=0.4 10.0 FT2 8.0 FT2 4.0 FT2 Peak Clad Temperature (PCT), F 2108 2188 2064 1983 1961 1911 1746 Time of PCT, sec. 186.1 183.1 181.5 177.6 176.1 172.0 130.1 Peak Clad Temperature Location, ft. 9.72 9.47 9.72 9.72 9.72 9.72 9.97 Local Zr/H20 Reaction (max.), %* 3.40 4.53 2.99 2.26 2.19 1.91 1.92 Local Zr/H2O Location, f t, from bottom 9.47 9.47 9.72 9.72 9.72 9.72 9.47 Total Hydrogen Generation, % of total Zr Reacted <1.0 <1.0 <1.0 <1.0 <1.0 < 1.0 <1.0 E
Hot Rod Burst Time, sec. 39.8 32.0 43.9 54.1 56.2 62.76 67.83 Hot Rod Burst Location, ft. 7.72 7.72 7.72 7.72 7.72 7.72 9.47 Analysis In_p,ut License Core Power, Hwt 2700 Power Used in Analysis, Mwt 2754 Local Peaking Factor, Min. 1.1 Radial times Local Peaking (FaH) 1.84 Axial Power Profile 15.0 kw/ft at 0.6 x/1 Exposure Conditions Maximim Stored Energy (BOL) on x 5?
- At 200 seconds @ fo
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11 XN-NF-85-117 Supplement 1 Table 2.3 LOCA Exposure Study Results 0.8DECLG Break Exposure Study Results Peak Rod Average Burnup (MWD /kg) 1.8 14.0 49.1 Peak Node Fuel Average Temperature (OF) 2380 2097 1927 Hot Rod Burst
. Time ( sec) 32.0 40.6 76.8
. Elevation ( ft) 7.72 7.72 9.22
. Channel blockage Fraction .393 .504 .648 Peak Clad Temperature
. Time (sec) 183.1 183.6 181.6
. Elevation ( ft) 9.47 9.72 9.97
. Temperature (OF) 2188 2077 1730 Zr-Steam Reaction
. Local Maximum Elevation (ft) 9.47 9.72 9.72
. Local Maximum (%)* 4.53 3.18 1.22**
. Core Maximum (%) 1.0 1.0 1.0 Analysis Input License Core Power, Mwt 2700 Power Used in Analysis, Mwt 2754 Local Peaking Factor, Min. 1.1 Radial times Local Peaking (F 3H) 1.84 Axial Power Profile 15.0 kw/ ft at 0.6 x/1
- At 200 sec **At 250 sec
12 XN-NF-85-ll7 Supplement 1
3.0 CONCLUSION
S A revised LOCA-ECCS analysis has been performed for St. Lucie Unit I with ENC 14x14 fuel using the EXEM/PWR ECCS Evaluation model in conformance with Appendix K of 10 CFR 50. The revised analysis corrects an error in previous analyses and increases the analyzed level of steam generator tube plugging to 15%. The limiting LOCA break was identified as the large double-ended guillotine break of the reactor vessel inlet pipe with a discharge coefficient of 0.8. Axially dependent LHGR limits were determined based on the limiting break LOCA which assure compliance with NRC 10 CFR 50.46 criteria. The analyses were performed for bounding exposure conditons and apply for St. Lucie Unit I cycle 7 and beyond using ENC supplied fuel.
The results of the limiting break LOCA-ECCS analysis were presented in Reference 1. Axially dependent LHGR limits were determined which assure conformance with NRC criteria. The required LOCA-ECCS large break spectrum results and exposure analysis results which confirm the limiting LOCA break and the worst case exposure are contained in this supplement.
The limits calculated for ENC fuel are equal to or conservative with respect to those currently in place for CE fuel. Additionally the CE fuel will operate with LHGRs significantly below those of ENC fuel due to the CE fuel exposures relative to the ENC fuel exposures. It is therefore conservative to monitor the CE fuel to the limits established for the ENC fuel.
13 XN-NF-85-ll7 Supplement 1 Operation of the St. Lucie 1 reactor with ENC 14x14 fuel within the limits given in Reference 1 assures that the St. Lucie 1 emergency core cooling system will meet the acceptance criteria as required by 10 CFR 50.46. That is:
- 1. The calculated peak fuel element clad temperature does not exceed the 2200 F limit.
- 2. The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the core.
- 3. The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching.
- 4. The system long term cooling capabilities provided for previous cores remains applicable to cores containing ENC reload fuel.
l 14 XN-NF-85-117 Supplement 1
5.0 REFERENCES
(1) "St. Lucie Unit 1 Revised LOCA-ECCS Analysis with 15% Steam Generator Tube Plugging," XN-NF-85-ll7, Exxon Nuclear Company, November 1985.
(2) "St. Lucie Unit 1 LOCA Analysis Using the EXEM/PWR ECCS Model,"
XN-NF-82-98, Supp. 1, Revision 1, Exxon Nuclear Company, January 1983.
(3) " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and 10 CFR 50 Appendix K.
(4) " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model,"
XN-75-41 and supplements, Exxon Nuclear Company, 1975.
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15 XN-NF-85-ll7 Supplement 1 Issue Date 12/6/85 ST. LUCIE UNIT 1 REVISED LOCA-ECCS ANALYSIS WITH 15% STEAM GENERATOR TUBE PLUGGING BREAK SPECTRUM AND EXPOSURE RESULTS Distribution M. J. Ades K. A. Bryan R.' A. Copeland N. F. Fausz R. C. Gottula T. J. Helbling J. S. Holm J. W. Hulsman S. E. Jensen J. D. Kahn W. V. Kayser J. N. Morgan G. L. Ritter C. E. Slater T. Tahvili G. N. Ward H. E. Williamson L. J. Federico FPL/T. J. Helbling (25)
Document Control (5)
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