ML20083N248
| ML20083N248 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/07/1983 |
| From: | Holm J, Lindquist T, Nutt W SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17213B024 | List: |
| References | |
| XN-NF-82-81, XN-NF-82-81-R01, XN-NF-82-81-R1, NUDOCS 8302010599 | |
| Download: ML20083N248 (68) | |
Text
{{#Wiki_filter:_. --- _ XN NF 82 81 REVISION 1 l' i 1! ii - ST. LUCIE UNIT 1 CYCLE 6 l SAFETY ANALYSIS REPORT i , l[ l l l
, JANUARY 1983 i .
l! . . ll l; RICHLAND, WA 99352 i ll ERON NUCLEAR COMPANY,Inc. PDR
1 XN-NF-82-81 , f ' Revision 1 i
. Issue Date: 1/7/83 ST. LUCIE UNIT 1 CYCLE 6 SAFETY ANALYSIS REPORT 1
Contributors: K. A. Bryan l J. S. Holm i T. R. Lindquist W. T. Nutt Written by: -
!. n Te- f l'r PJ J4 5. H'olm Engineer PWR Neutr its
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Written by: / G Jm 63 T. R. Linaquisy, Engineer Plant Transier/t Analysjs Written by: // /5&_. / l W. 7. 'Nutt, Enginey't/ ; , Pla t Tra sient Analysis l 4 Revie'..ed by: a [ M~ f A+3 $$ ) F. S. Skogent,J Manager PWR Neutronics Reviewed by: 9' / m_. . , . ( /./ r; H. E. Williamson, Manager-Neutronics FelMapagement Reviewed by: h hWM! R. S. 5 tout, Manager 6 ra n 93 Licensing & afety Engineering Concurred by: Jr f.,_ //t /hr
- J. ~. Morgan, Wanager '/ .
P oposals & Cus omer S rvices Engineering f i Approved by: U d' .pn a f t l G. A. Sofer, Manager l Fuel Engi eering & Technical Services Approved by: d,sc ., , /[+[f) ] l G. J. Susselman, Manager Fuel Design
/ mar l EQON NUCLEAR COMPANY,Inc.
i t t I i t NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICS REGARDING CONTENTS AND USE OF THIS DOCUMENT 4 PLEASE READ CAREFULLY This technical -eport was derived through research and development programs sponsorw: by Exxon Nuclear Company, Inc. It is being sub- t mitted by Exxon Nuclear to the USNRC as part of a technical contri- i bution to facilitare safety analyses by licensees of the USNRC which utilize Exxon Nu:: ear fabricated reload fuel or other technkal services provided by Exxcn Nuclear for licht water power reactors and it is true and correct to - :e best of Exxon Nuclear's knowledge, information, and belief. The imormation contained herein may be used by the USNRC in its review df nis report, and by limnsees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration i of compliance with :he USNRC's regulations. Without derogating from the foregoing, neither Exxon Nuclear nor any person acting .,n its behalf: , A. Makes any warranty, express or implied, with respect to the accurr. y, completeness, or usefulness of the infor-mation ccntained in tv. document, or that the use of any infor: ation, apparatus, method, or promss disclosed in this dc:ument 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, sp
- paratus, r--thod, or process disclosed in this document. fe m
9 XN- NF- F00, 766 t O
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- i- XN-NF-82-81 Revision 1 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.0 OPERATING HISTORY'0F THE REFERENCE CYCLE . . . . . . . . . . . . . . . 4 4.0 GENERAL DESCRIPTION. . . . . . . . . . . . . . . . . . . . . . . . . . 7 5.0 FUEL SYSTEM DESIGN . ........................12 6.0 NUCLEAR CORE DESIGN. . . . . . . . . . . . . . . . . . . . . . . . . 13 6.1 PHYSICS CHARACTERISTICS . . . . . . . . . . . . . . . . . . . . .* 14 6.1.1 Power Distribution Considerations. . . . . . . . . . . . 15 6.1.2 Control Red Reactivity Requirements. . . . . . . . . . . 15. 6.1.3 Moderator Temperature Coefficient Considerations . . . . 16 6.2 ANALYTICAL METHODOLOGY. . . . . . . . . . . . . . . . . . . . . 17
- 7.0 THERMAL-HYDRAULIC DESIGN ANALYSIS. . . . . . . . . . . . . . . . . . 23 7.1 DESIGN BASES AND CRITERIA . . . . . . . . . . . . . . . . . . . 23 7.2
SUMMARY
OF THERMAL-HYDRAULIC DESIGN ANALYSIS RESULTS. . . . . . 23 7.3 HYDRAULIC CHARACTERIZATION. . . . . . . . . . . . . . . . . . . 24 7.4 CORE FLOW DISTRIBUTION ANALYSIS . . . . . . . . . . . . . . . . 2S \ . l 7.5 MDNBR SUBCHANNEL ANALYSIS . . . . . . . . . . . . . . . . . . . . 26 7.6 R0D B0W . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 l 7.7 FUEL CENTERLINE TEMPERATURE . . . . . . . . . . . . . . . . . . 29 I
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XN-NF-82-81 Revision 1 l TABLE OF CONTENTS (CONTINUED) Section Page ' 8.0 ACCIDENT AND TRANSIENT ANALYSES. . . . . . . . . . . . . . . . . . . 37 8.1 PLANT TRANSIENT ANALYSIS. . . . . . . . . . . . . . . . . . . . 37 8.2 ECCS ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . 37 8.3 R0D EJECTION ANALYSIS . . . . . . . . . . . . . . . . . . . . . 37 9.0
SUMMARY
OF OPERATING LIMITS. . . . . . . . . . . . . . . . . . . . . 41 9.1 REACTOR PROTECTION SYSTEM . . . . . . . . . . . . . . . . . . . 41 9.2 SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS . . . . . . . . . . . . 41 9.3 LIMITING SAFETY SYSTEM SETTINGS . . . . . . . . . . . . . . . . 42 9.3.1 -l Local Power Distribution Control . . . . . . . . . . . . 42 . 3 9.3.2 Thermal Margin / Low Pressure. . . . . . . . . . . . . . . 43 , i 9.3.3 Additional Trip Functions. . . . . . . . . . . . . . . . 44
- 9.4 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . 44 9.4.1 DNB Monitoring . . . . . . . . . . . . . . . . . . . . . 44
( 9.4.2 Linear Heat Rate Monitoring. . . . . . . . . . . . . . . 44 l-9.5 SETPOINT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 45 I l 9.5.1 Limiting Safety System Settings. . . . . . . . . . . . . 45 1
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9.5.2 Limiting Conditions for Operation. . . . . . . . . . . . 47
10.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 I
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-lii- XN-NF-82-81 Revision 1 LIST OF TABLES Table Page 4.1 St. Lucie Unit 1 Principal Characteristics for Nuclear Analysis of Cycle 6 Fuel . . . . . . . . . . . . . . . . . . . . . . . 9 k 6.1 St. Lucie Unit 1 Neutronics Characteristics of Cycle 5 Compared with Cycl'e 6 Data . . . . . . . . . . . . . . . . . . . . . 18 6.2 Comparison of Shutdown Margin Cycles 5 and 6 . . . . . . . . . . . . 19 7.1 Assembly Component Loss Coefficients . . . . . . . . . . . . . . . . 30 1
7.2 Thermal-Hydraulic Design Data. . . . . . . . . . . . . . . . . . . . 31 8.1 St. Lucie Unit 1 Cycle 6, Ejected Rod Analysis, HFP. . . . . . . . . 39 8.2 St. Lucie Unit 1 Cycle 6, Ejected Rod Analysis, HZP. . . . . . . . . 40 9.1 Additional LSSS Trip Functions . . . . . . . . . . . . . . . . . . . 49 , I 9.2 Uncertainties Applied for LPD Trip Calculation . . . . . . . . . . . 50 9.3 Uncertainties Applied to Only the DNBR Calculation . . . . . . . . . 51 9.4 Uncertainties Applied for the LCO Based on LPD . . . . . . . . . . . 52 9.5 Uncertainties Applied to LC0 Based on DNB. . . . . . . . . . . . . . 53 l l
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-iv- XN-NF-82-81 Revision 1 LIST Of FIGURES Figure Page 3.1 St. Lucie Unit 1 Cycle 5 Power Distribution Comparison, 5,000 MWD /MT (3,670 EFPH), HFP, ARO. . . . . . . . . . . . . . . . . . 5 3.2 St. Lucie Unit 1 Cycle 5, Critical Boron Concentration vs.
Exposure, AR0, HFP, 3-D XTG and Measured . . . . . . . . . . . . . . . 6 8 4.1 St. Lucie Unit 1 Cycle 6, Core Loading Pattern . . . . . . . . . . . 10 i 4.2 St. Lucie Unit 1 Cycle 6 Loading Pattern with 656 B 4C Rods. . . . . 11 6.1 St. Lucie Unit 1 Cycle 6 Boron Rundown Curve, ARO, HFP, EOC5 = 13,215 MWD /MT . . . . . . . . . . . . . . . . . . . . . . . . 20 6.2 St. Lucie Unit 1 Cycle 6, Relative Power Distribution, 100 MWD /MTU,1,013 ppm, 2iOO MWt, ARO. . . . . . . . . . . . . . . . 21 f 6.3 St. Lucie Unit 1 Cycle 6, Relative Power Distribution, . 15,492 MWD /MTU, 23 ppm, 2700 MWt, ARO. . . . . . . . . . . . . . . . 22 ! 7.1 Comparison of ENC and CE Component loss Coefficients . . . . . . . . 32 7.2 Cycle 6 1/8 Core Model . . . . . . . . . . . . . . . . . . . . . . . 33 7.3 Cycle 5 1/8 Core Model . . . . . . . . . . . . . . . . . . . . . . . 34 7.4 ENC 1/8 Assembly Subchannel Model. . . . . . . . . . . . .-. . . . . 35 7.5 CE 1/8 Assembly Subchannel Model . . . . . . . . . . . . . . . . . . 36 9.1 St. Lucie Uni t 1 LPD Barn. . . . . . . . . . . . . . . . . . . . . . 54 9.2 St. Lucie Unit 1 TM/LP Trip function A1. . . . . . . . . . . . . . . : 55 9.3 St. Lucie Unit 1 TM/LP Trip Function QR1 . . . . . . . . . . . . . . 56 9.4 St. Lucie Unit 1 Limi+,ing Condition for Operation DNB Barn . . . . . 57 9.E St. Lu:ie Unit 1 Limiting Condition for Operation LHR Barn . . . . . 58 I
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f 8 XN-NF-82-81 Revision 1 ST. LUCIE UNIT 1 CYCLE 6 SAFETY ANALYSIS REPORT
1.0 INTRODUCTION
The results of the. Safety Analysis for Cycle 6 of the St. Lucie Unit I nuclear plant are presented in this report. The topics 2.ddressed include operating history of the reference cycle, power distribution considerations, control rod reactivity requirements, temperature coeffi-cient considerations, thermal-hydraulic design analysis, control rod ejection accident analysis, and setpoint analysis. The Cycle 6 design includes 84 Exxon Nuclear Company (ENC) Reload Batch XN-1 assemblies enriched to 3.67 w/o U-235. The Cycle 6 design utilizes 656 B 4C-Al 0 burnable absorber rods, each containing 23.8 mg 23 B-10 per inch. The reload consists of 28 assemblies with no burnable absorber rods, 8 assemblies with 4 burnable absorber rods, 36 assemblies with 12 burnable absorber rods, and 12 assemblies with 16 burnable absorber rods. 1
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l XN-NF-82-81
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SUMMARY
The St. Lucie Unit 1 nuclear plant is scheduled to operate in Cycle i 6 b'eginning in April of 1983 with 84 fresh assemblies supplied by Exxon Nuclear Company (ENC) (Reload Batch XN-1 or H). The composition of the core during Cycle 6 will be 84 ENC assemblies in Batch H and a total of 133 Combustion Engineering (CE) assemblies; 64 in Batch G, 68 in Batch F and 1 in Batch E. The characteristics of the fuel and the reloaded core are in confor-mance with existing Technical Specification limits or with proposed revised Technical Specification limits supported by ENC analyses regarding shutdown margin provisions and temperature coefficient limits. The ENC fuel design is presented in References 1 and 2. The Plant Transient Analysis is presented in Reference 3 and LOCA-ECCS Break Spectrum Analysis is provided in Reference 4. The results of the Control Rod Ejection Analysis are provided herein and are derived in part from the generic analysis described in Reference 5. The neutronic characteristics of Cycle 6 are similar to those of Cycle 5. The range of kinetics coefficients reflected in the Plant Transient Analysis bounds those expected in Cycle 6. The excess' shutdown - margin at E0C HZP is calculated to be 1500 pcm. A postulated control rod ejection event is conservatively calculated to result in an energy deposition of less than 155 cal /gm.
XN-NF-82-81 Revision 1 The peaking factors F and F at BOC HFP equilibrium xenon conditions are 1.511 08 and 1.551 08, respectively and occur in an ENC assembly. The peaking factors Fr and F, at B0C for Combustion Engineering assemblies i are 1.391 07 and 1.51+.08, respectively. The maximum values for the peaking factors Fr and F xy occur at E0C in an ENC assembly and are 1.531 08 and 1.561 08, respectively. The St. Lucie Unit 1 Cycle 6 Technical Specification limit on F and F xy is 1.70. 4 t l 9
XN-NF-82-81 Revision 1 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE St. Lucie Unit 1 Cycle 5 has been chosen as the reference cycle with respect to Cycle 6 due to the close resemblance of the neutronic characteristics between these two cycles. The Cycle 5 operations began in December 1981 with s.hutdown scheduled for February 1983. .The core has accrued about 10,000 MWD /MT as of the end of November 1982. The Cycle 5 core consisted of 217 Combustion Engineering assemblies. The measured power peaking factors at hot-full-power equilibrium xenon conditions have remained below Technical Specification limits throughout Cycle 5. The linear heat rate including uncertainties of 7% for measurement, 3% for engineering, 1% for densification and 2% for power, has remained below 13.7 kw/ft for Cycle 5. The peaking factors F and F have both remained below 1.60 for Cycle 5. Cycle 5 has operated essentially free of control rods and Cycle 6 is anticipated to operate in a similar manner. A Cycle 5 assembly power distribution as measured by the Comt;stion Engineering code CECOR is shown in Figure 3.1 at 5,000 MWD /MTU. Calculations with the XTGPWR(9) code are also shown in Figure 3.1 for comparison l with the measured data. The measured critical boron concentration versus - exposure for Cycle 5 is shown in Figure 3.2. The critical boron curve predicted by ENC is plotted in Figure 3.2 for comparison with the measured data. l
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- XN-NF-82-81 Revision 1 i
K H L J G F E D C B A i 1 E2 2 F2 3 Fil 4 El 5 F1 6 El 7 F1 8 F2 11 .870 1.035 1.272 1.015 1.212 .982 1.219 .997 1kf lkb dOh k 28f hh db - 9f .857 10 F2 11 G4 12 F2 13 G3 14 El 15 El 17 .843 9 1.035 1.266 1.075 1.262 .984 1.250 F1 16 1.004 1.099 G3
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j7gg4 {9gg 2 17 g0 {y0 g }390 19 F1 20 F2 21 F1 22 D1 23 F1 24 F2 25 F2 26 G1 .392 1 1.272 1.077 1.301 .966 1.253 1.005 .957 1.095 .51 7 1.220 1.082 1.261 .994 1.230 1.032 .981 1.111 i -4.09 .46 -3.07 2.90 -1.84 2.69 2.51 1.46 ! 27 El 28 G3 29 D1 30 F1 31 El 32 G3 33 F2 34 G1 1 1.015 1.265 .965 1.185 .956 1.231 .950 .884 6 1.021 1.228 .998 1.174 1.002 1.204 .981 .913 .
.59 -2.92 3.42 .93 4.81 -2.19 3.26 3.28 !
35 F1 36 El 37 F1 38 El 39 F1 40 F2 41 G1 42 E2 1.212 .991 1.257 .969 1.2 61 .979 1.080 .373 - - 5 . 1.178 1.015 1.235 1.005 1.227 .999 1.073 .396 ' 9 A1 2.42 -1.75 3.72 -2.70 2.04 .65 6.17 -- 43 El 44 F1 45 F2 46 G3 47 F2 48 G2 49 E2
.982 1.265 1.013 1.240 .985 1.068 .397 4
1.001 1.236 1.044 1.207 .998 1.046 .441 1.93 -2.29 3.06 -2.66 1.32 -2.06 11.08 50 F1 51 El 52 F2 53 F2 54 G1 55 E2 r 1.219 1.012 .965 .966 1.087 .411 CECOR 3 1.165 1.012 .985 -}
.983 1.074 .441 XJG'
-4.43 ----- 2.07 1.76 -1.20 7.30 (- -1)x100 56 F2 57 G3 58 G1 59 G1 60 E2 2
.997 1.108 1.096 .884 .356 ,.
.978 1.070 1.114 .915 .397
- 1. 41 1 41 1.64 1.R1 11 R2
.86h .38b 1 .843 .397 -
-1.98 3.12 ,
k Figure 3.1 St. Lucie Unit 1, Cycle 5 Power Distribution } Comparison, 5,000 MWD /MT (3,670 EFPH), HFP, AR0 i
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0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 Cycle Exposure (GWD/MT) Figure 3.2 St. Lucie Unit 1, Cycle 5 Critical Boron Concentration vs. Exposure, ARD, HFP, 3-D XTG and Measured
XN-NF-82-81 Revision 1 1 4.0 GENERAL DESCRIPTION The St. Lucie Unit 1 reactor consists of 217 assemblies, each having ) a 14x14 fuel rod array. The assemblies are composed of up to 176 fuel rods, a number of B4 C-A1 0 burnable absorber rods in place of fuel rods , 23 (0 to 16 rods), and 5 control rod guide tubes or 4 control rod guide tubes and one instrument tube. The fuel rods consist of slightly enriched UO 2 pellets inserted into zircalcy tubes. The control rod guide tubes and instrument tubes are also made of zircaloy. Each ENC assembly contains nine zircaloy spacers with Inconel springs. i The projected Cycle 6 loading pattern is shown in Figure 4.1 with assemblies identified by their assembly number and Cycle 5 core location. 23 burnable absorber pins along The initial enrichment, number of B C-Al 4 j with other descriptive parameters are shown in Table 4.1 for the fuel assemblies in Cycle 6. The calculated B0C 6 exposures, based on an E0C 5 exposure of 13,215 MWD /MTV, are shown in a quarter core representation in Figure 4.2 along with the fuel type identification. The core consists of 84 fresh ENC assemblies at an average enrichment of 3.67 w/o U-235 and 133 exposed Combustion Engineering assemblies. A low radial leakage fuel management plan has been developed and results in scatter-loading - of the fresh fuel throughout the core with the fresh assemblies loaded in the core interior containing B C-Al 0 burnable absorber rods. The 4 23 m-- - - - - . - - - - - - - - _ _ _ _ - - - - - - - - _ _ . - - - - - - - - - - - - . . . - - - - _ - - - - - _ - - - - - _ _ - - - _ - - - - - - .
XN-NF-82-81
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exposed fuel is also scatter-loaded in the center in a manner to control the power peaking. The B4 C-Al23 0 burnable absorber rods contain 23.8 mg B-10 per inch. A total of 656 burnable absorber rods are utilized; 8 assemblies with 4 buenable absorber rods, 36 assemblies with 12 burnable absorber rods and 12 assemblies with 16 burnable absorber rods. 1
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Table 4.1 St. Lucie Unit 1 Principal Characteristics for . Nuclear Analysis of Cycle 6 Fuel E2 F1 F2 Gl G2 G3 G4 H1 112 H3 H4 Enrichment (w/o) 2.75 3.65 3.03 3.65 3.65 3.20 3.03 3.67 3.67 3.67 3.67 No. 84C Pins 0 0 12 0 4 8 8 0 4 12 16 Nominal Density 94.75% 94.75% 94.75% 94.75% 94.75% 94.75% 94.75% 94% 94% 94% 94% Pellet 00 (in.) .377 .377 .377 .377 .377 .377 .377 .370 .370 .370 .370 Clad OD (in.) .440 .440 .440 .440 .440 .440 .440 .440 .440 .440 .440 Diametral Gap (in.) .0075 .0075 .0075 .0075 .0075 .0075 .0075 .008 .008 .008 .008 Clad Thickness (in.) .028 .028 .028 .028 .028 .028 .028 .031 .031 .031 .031 Rod Pitch, (in.) .580 .580 .580 .580 .580 .580 .580 .580 .580 .580 .580
' Spacer Material Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Bi- Bi- Bi- Bi- ,
Metal. Metal. Metal. Metal. ? Fuel Supplier CE - CE CE CE CE CE CE ENC ENC ENC ENC Fuel Stack lleight Nominal (in.) 136.7 136.7 136.7 136.7 136.7 136.7 136.7 136.7 136.7 136.7 136.7 No. of Assemblies 1 40 28 32 4 24 4 28 8 36 12 Re ionwise loading MTU) .389 15.56 10.13 12.44 1.52 8.92' 1.49 10.67 2.98 12.78 4.16 Exposure (MWD /MT) 80C6 25992 26526 28321 12799 13820 15897 16752 0 0 0 0 EOC6 3/539 41834 35561 29888 32206 31449 31445 14379 15796 20033 19651 Incremental = >< Exposure 11547 15308 7240 17089 18386 15552 14693 14379 15796 20033 19651
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XN-NF-82-81 Revision 1 X W V T S R PN ML KJ HG F' D C Y E B Af l ll ll ll ll F101 H128 H101 F114 l 21 3,gg p,gg , F119 H116 H120 G121 H337 G105 H105 H109 F105 i 20 T-18 S-18 F-18 E F145 H236 G008 G110 H477 G001 H473 G118 G013 H229 F126 19 R-19 8-06 N-06 X J-06 X-06 G-19 I F137 H124 G123 H345 F026 G005 F104 G031 F020 H346 G103 H121 F117 i i 18 W-15 X-13 R-17 M-01 L-02 K-01 G-17 B-13 C-15 F125 H232 G108 H357 F002 H353 F015 H365 F019 H354 F003 H358 G107 H233 F111 17 i V-17 N-20 L-07 T-17 G-15 L-03 J-20 D-17 H112 G002 H349 F013 H369 F011 G014 G035 G027 F007 H370 F008 H350 G021 H113 16 - , F-02 C-11 E-11 W-05 D-04 C-05 S-06 R-11 5-02 ; l H108 G120 F036 H361 F006 H341 F030 H481 F040 H342 F023 H362 F033 G116 H117 - i 15 F-13 T-15 F-06 N-18 J-18 L-05 E-15 S-13 - pg, , F118 I 14 - W-16 G113 H480 G006 F038 G024 F027 G022 G203 G023 F031 G009 F018 G012 H474 G104, C;; H127 V-16 A-12 R-15 E-19 V-13 R-20 N-13 B-15 0-13 T-19 E-17 0-16
] Y-12 H125 H340 G010 F113 H368 G036 H484 G202 E105 G204 H482 G034 H366 F140 G018 H338 I. .
11 - R-02 B-11 0-18 N-09 E-02 J-13 V-04 X-11 G-20 l
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G119 H476 G030 F022 G016 F021 G017 G201 G025 F029 G015 F035 G032 H478 G106 9- V-06 A-10 T-05 E-03 V-09 X-07 J-09 G-02 0-09 T-03 G-07 Y-10 0-06 F121 pg; . 8 -- W-06 H119 G115 C-Oo I F034 H364 F009 h344 F039 H483 F037 H343 F005 H363 F025 G109 H106 7- F-09 T-07 L-17 N-04 J-04 S-16 E-07 S-09 . H115 G029 H352 F017 H372 F001 G028 G033 G020 F004 H371 F014 H351 G019 H110 6 F-20 G-11 F-16 W-17 V-18 C-17 T-11 W-11 S-20 _ F103 H235 G102 H360 F012 H356 F024 H367 F010 H355 F016 H359 G101 H230 F102 1 5 V-05 N-02 L-19 R-07 E-05 L-15 J-02 0-05 t s
, F115 H123 G112 H348 F032 G011 'F106 G003 F028 H347 G124 H122 F136 W-07 X-09 R-05 M-21 L-20 K-21 G-05 B-09 C-07
{ F141 H231 G004 G117 H479 G026 H475 Gill G007 H234 F143 l 3 R-03 B-16 N-16 B-07 J-16 X-16 G-03 F107 Hill H107 G122 H339 G114 K118 H114 F139 ! 2 T-04 S-04 F-04 E-04 F116 H126 H103 F124 Assemoly Number ) 1 S-03 F-03 , Cycle 5 Position } l Figure 4.1 St. Lucie Unit 1, Cycle 6 j Cere Loading Pattern
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XN-NF-82-81 K H Revision 1 L J G F E D C B A i 11 26.0 16.8 0.0 13.8 0.0 25.8 14.3 0.0 0 0.0 9 16.8 14.3 25.2 13.8 25.9 11.4 0.0 16.3 8 26.7 .) 7 0.0 25.3 0.0 27.9 0.0 25.7 16.8 0.0 6 13.8 13.8 27.9 0.0 27.6 0.0 11.7 0.0 5 0.0 25.3 0.0 28.6 0.0 14.5 0.0 29.8 4 25.8 11.4 25.8 0.0 14 .5 0.0 29.7 ,, 3 14.3 0.0 16.8 11.7 0.0 29.8 BOC6 Assembly Exposure GWD/MTU 2 0.0 16.4 0.0 0.0 29.8 ENC Number of Fuel Types Enrichment 84C Rods 1 0.0 26.7 l H1 3.67 w/o 0 - H2 3.67 w/o 4 H3 3.67 w/o 12 H4 3.67 w/o 16 Figure 4.2 St. Lucie Unit 1 Cycle 6, Loading Pattern with 656 B4C Rods
XN-NF-82-81 Revision 1 5.0 FUEL SYSTEM DESIGN A description of the Exxon Nuclear supplied fuel design and design methods is contained in References 1 and 2. This fuel has been specifically designed to be compatible with the resident fuel supplied by Combustion Engineering. 6 0 k I i
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XN-NF-82-81 Revision 1 6.0 NUCLEAR CORE DESIGN The neutronic characteristics of the projected Cycle 6 core are quite similar to those of the Cycle 5 core (see Section 6.1). The nuclear design bases for the Cycle 6 core are as follows:
- 1. The design shall permit operation within the Technical Specifi-cations for St. Lucie Unit 1 nuclear plant.
- 2. The length of Cycle 6 shall be determined on~the basis of an assumed Cycle 5 energy of 13,215 MWD /MT exposure. .
- 3. The Cycle 6 loading pattern shall be designed to achieve power distributions and control rod reactivity worths according ,
to the following constraints: ,, a) The peak LHR shall not exceed 15 kw/ft and the peaking factors Frand F x shall not exceed 1.70 in any single fuel rod through the cycle under nominal full power operating conditions. b) The scram worth of all rods minus the most reactive rod shall exceed BOC and E0C shutdown requirements. The neutronic design methods utilized to ensure the above requirements are consistent with those described in References 6, 7 and 8. The Cycle 6 loading contains 656 fresh Al 23 0 -B4C burnable absorber rods distributed among 56 of the 84 fresh ENC supplied assemblies. The reload consists of assem: lies at 3.67 w/o U-235, 28 assemblies with no i l i
XN-NF-82-81 Revision 1 burnable absorber rods, 8 assemblies with 4 burnable absorber rods, 36 assemblies with 12 burnable absorber rods and 12 assemblies with 16 burnable absorber rods. The A123 0 ' 4C burnable absorber rods each contain 23.8 mg/in of B-10. The core loading pattern has been designed to achieve a desirable power distribution while maximizing the benefit of assemblies with burnable absorbers to reduce the beginning of cycle l (B0C) boron concentration. 6.1 PHYSICS CHARACTERISTICS The neutronics characteristics of the Cycle 6 core are compared ,. I with those of Cycle 5 and are presented in Table 6.1. The data presented ' in the table indicates the neutronic similarity between Cycles 5 and 6. The reactivity coefficients of the Cycle 6 core are bounded by the i coefficients used in the safety analysis. The safety analysis for Cycle 6 is applicable for Cycle 5 burnups of +500 MWD /MT and -500 MWD /MT , about the nominal burnup of 13,215 MWD /MT. The boron letdown curve for Cycle 6 is shown in Figure 6.1. < The 80C6 HZP xenon free critical boron concentration is calculated to be 1,432 ppm. At 100 MWD /MT, equilibrium xenon, HFP, the critical boron concentration is 1,013 ppm. The Cycle 6 length is projected to - be 15,492 MWD /MT +300 MWD /MT at a core power of 2,700 MWt with 23 ppm I soluble boron remaining. ;
,\
a
XN-NF-82-81 Revision 1 6.1.1 Power Distribution Considerations Representative calculated power maps for Cycle 6 are shown in Figures 6.2 and 6.3 for BOC (100 MWD /MTU, equilibrium xenon), and E0C canditions, respectively. The power distributions were obtained from a three-dimensional XTGPWR(9) model with moderator density and Doppler feedback effects incorporated. As shown in Figure 6.2 for the design Cycle 6 loading pattern, the calculated BOC hot full power (HFP) equilibrium nuclear peaking factors Fr and F are 1.51 and 1.55, respectively. The peak linear he. ' rate (LHR) at B0C HFP is calculated to be 11.1 kw/f t. At E0C conditions the corresponding values for F , F and LHR are 1.53, r 1.56 and 10.5 kw/ft, respectively. The Technical Specification limit on F and F is 1.70. The B0C HFP equilibrium xenon LHR including uncertainties of 7% II9) for measurement, 3% for engineering, 1% for densification and 2% for thermal power is 12.6 kw/ft and is comparable to the measured value for Cycle 5 BOC HFP equilibrium xenon of 13.7 kw/ft. At E0C HFP equilibrium xenon conditions the maximum LHR is calculated to be 11.9 kw/ft including uncertainties. The Technical Specification limit on , I LHR is 15 kw/ft. 6.1.2 Control Rod Reactivity Requirements Shutdown margin evaluations for Cycle 5 and 6 are compared in Table 6.2. The Cycle 5 calculations are taken from the - Cycle 5 Safety Analysis Report by Combustion Engineering and the Cycle . __ y ,. p ., : .; . -- Ly. 4
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. .c XN-NF-82-81 Revision 1 6 calculations were performed by ENC. The comparisons assume a Technical Specification for the required shutdown margin of 5000 pcm. The Cycle i
5 excess shutdown margin is 200 pcm while for Cycle 6 it is 100 pcm. The ENC plant transient analysis (3) supports a technical specification change in the required shutdown margin from 5000 pcm to 3600 pcm. The excess shutdown margin at E0C6 HZP with a required shutdown margin of 3600 pcm is 1500 pcm. The control rod groups and insertion limits for Cycle 6 will remain unchanged from Cycle 5. With these limits the nominal worth of the control barh, Bank 7, inserted to the insertion limitt at HFP is 120 pcm at B0C and 190 pcm at E0C.
^^
6.1.3 Moderator Temperature Coefficient Considerations The current Technical Specifications require that ; the moderator temperature coefficient be less than +5 pcm/0F at or i below 70% of rated thermal power, less than 2 pcm/0F above 70% power i and greater than -22 pcm/0F at 100% of rated thermal power. The ENC . plant transient analysis accounts for revised Technical Specifications on the moderator temperature coefficients of less than or equal to +7 L pcm/0F at HZP, less than or equal to +2 pcm/0F at HFP and greater than - t or equal to -28 pcm/0F at HFP. Tne BOC HIP, AR0 moderator temperature l i I
- ):
XN-NF-82-81 Revision 1 coefficient is calculated to be 4.8 pcm/0F and will meet the revised Technical Specification limit at HZP conditions. The moderator temperature coefficient at 100% rated power, BOC, equilibrium xenon conditions is calculated to be -1.3 pcm/0F. The moderater temperature coefficient at EOC, HFP, equilibrium xenon conditions is calculated to be -21 pcm/0F. 6.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 6 core analysis are described in References 6, 7 and 8. In summary, the reference neutronic design analysis of the reload core was performed using the XTGPWR(9) reactor simulator code. The input isotopics data were based on quarter core depletion calculations performed for Cycle 6 using the XTGPWR code. *
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The fuel shuffling between cycles was accounted for in the calculations. Calculated values of LHR, F , and F, were determined with the XTGPWR reactor model. The calculational thermal-hydraulic feedback and axial exposure distribution effects on power shapes, rod worths, and cycle lifetime are explicitly included in the analysis. e 6
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XN-NF-82-81 Revision 1 Table 6.1 St. L ucie Unit 1 Neutronics Characteristics of Cycle 5 Compared with Cycle 6 Data Cycle 5 Cycle 6 ) BOC E0C BOC EOC Critical Boron (ppm) ' HFP, AF.0, Equilibrium xenon 980(a)(d) 20(c) 1013(d) 23 HZP, AR0, No xenon 1396(a) 1432 -- i Moderator Temperature Coefficient , pcm/0F HFP .4(a) --
-1.3(d) -20.7 I
HZP 5.8(a) -- 4.8 -14.4 )
^
Doppler Coefficient (pcm/0F) -1.3(b) -1.5(b) -1.2 -1.4 t 1 Boron Worth (pcm/ ppm) HFP -- --
-8.5(d) -10.2 HZP -9.9(a) ,-
-8.8 -10.7 i
LHR (including uncertainties) 13.7(a) -- 12.6 11.9 } j Delayed Neutron Fraction .0061(b) .0052(b) .0058 .0050 1 i Control Rod Worth, ARI-1, - HZP (pcm) -- 7300(b) 6490 7120 I l Excess Shutdown Margin, HZP ' (pcm) -- 200(b)(e) -- 100(8) (a) Measured data (b) Cycle 5 Safety Analysis Report } (c) ENC calculated l l (d) 100 MWD /MTU (e) A required snutdown margin of 5000 pcm is utilized for the comparison.
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XN-NF-82-81 Revision 1 Table 6.2 Comparison of Shutdown Margin Cycles 5 and 6 EOC5* E0C6 HZP HZP Control Rod Worth (pcm) ARI 10300 9070 N-1 7300 7120 Power Dependent Insertion Limit (PDIL) 1500 1450 , [(N-1) - PDIL] * .9 5200 5100 Shutdown Margin (pcm) , Required Shutdown Margin 5000** 5000** . Excess Shutdown Margin 200 100
- From the Cycle 5 Safety Analysis Report.
** A revision to the Technical Specifications for St. Lucie Unit 1 Cycle 6 will be made to alter the required shutdown margin to 3600 pcm.
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0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 Cycle Exposure (GWD/MT) Figure 6.1 St. Lucie Unit 1, Cycle 6 Boron Rundown Curve, - 1 AR0, HFP, E0C5 = 13,215 MWD /MT ;
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l XN-NF-82-81 l K H l i L J G F E D C B A l l 11 .73 .91 1.12 1.24 1.26 .93 1.15 1.08 1 10 .80 ! l 9
.91 1.07 1.02 1.21 1.06 1.22 1.12 .89 l 8 .34 7 1.12 1.02 1.20 1.00 1.22 1.04 1.05 1.16 l
6 1.24 1.21 1.00 1.17 1.00 1.24 1.19 .97 5 1.26 1.07 1.21 .98 1.20 1.04 1.06 .35 4 -
.93 1.23 1.04 1.23 1.04 1.10 .38 ..
3 1.15 1.13 1.05 1.18 1.06 .38 2 1.08 .89 1.16 .97 .35 Peak Assembly = 1.26 (11E) 1 .80 .34 Peak F R = 1.51 (11E) , Peak F xy = 1.55 (11E) Peak LHR = 11.1 kw/ft (11E) Figure 6.2 St. Lucie Unit 1 Cycle 6, Relative Power Distribution, - 100 MWO/MTU, 1,013 ppm, 2700 MWt, AR0
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XN-NF-82-81 K H Revision 1 : L J G F E D C B A 11 .81 .96 1.23 1.16 1.25 .93 1.16 1.19 10 .83 9 .96 1.06 1.02 1.12 1.03 1.15 1.24 .95 j 8 .40 7 1.23 1.02 1.26 1.03 1.27 1.03 1.01 1.01 r 1.16 1.12 1.03 1.29 1.06 1.27 1.06 .84 5 1.25 1.03 1.27 1.0S 1.28 1.04 .37 J9 4 .93 1.15 1.03 1.26 1.04 1.02 .42 ,, I 3 1.16 1.24 1.01 1.06 .99 .42 2 1 19 .95 1.01 .84 .37 i
.83 .40 Peak Assembly = 1.29 (6F) ,
1 Peak F = 1.53 (6F) ' R Peak F = 1.55 (6F) xy l Peak LHR = 10.5 kw/ft (6F) Figure 6.3 St. Lucie Unit 1 Cycle 6, Relative Power Distribution, 15,492 MWD /MTU, 23 ppm, 2700 MWt, AR0 f 1
XN-NF-82-81
. Revision 1 7.0 THERMAL-HYDRAULIC DESIGN ANALYSIS This section identifies thermal-hydraulic design criteria for ENC reload fuel at St. Lucie Unit 1. The hydraulic compatibility of ENC fuel with existing CE fuel at St. Lucie Unit 1 is quantified in terms of hydraulic loss coefficients. The relative DNB performance of the two fuel types is compared. Expected peak fuel rod power levels relative to the power levels required for centerline melt are also examined.
7.1 DESIGN BASES AND CRITERIA The primary thermal-hydraulic design basis . for Exxon Nuclear Company reload fuel is that fuel rod integrity should be maintained during normal operation and anticipated operational occurrences. Specific criteria are: (1) Avoidance of boiling transition for the limiting fuel rod in the core with at least a 95% probability at a 95% ccnfidence level. (2) Fuel centerline temperatures should be below the melting point of the fuel pellets. ! Observance of these criteria during anticipated operational trans- ! ients is considered conservative relative to the requirement that anticipated operational transients not produce fuel rod failures or loss of functional capability. - The margin to boiling transition for ENC and CE fuel is assessed with ENC's XNB critical heat flux correlation (10). 7.2
SUMMARY
OF THERMAL-HYDRAULIC DESIGN ANALYSIS RESULTS The overall hydraulic loss coefficient of ENC reload fuel is found to be less than 4.0% greater than the loss coefficient of existing CE fuel. Thus, insertion of ENC fuel into the St. Lucie Unit 1 reactor will not l j significantly imoact orimary coolant flow. I
. e 24 XN-NF-82-81 Revision 1 Evaluation of thermal margin for both fuel types indicates MDNBRs are about 6.0% less for ENC fuel during Cycle 6* than for CE fuel evaluated for Cycle 5*" ' operation under conditions representative of a severe opera.
tional transient event. The bulk of this MDNBR reduction for ENC fuel is ) associated with its higher spacer loss coefficient compared to CE fuel. An evaluation of future cycles in which a larger fraction of the core will be ENC fuel indicates slight increases in assembly flow for both ENC and CE fuel, and e A thereby increases limiting assembly MDNBR for both fuel types. Thus MDNBR I results for the present Cycle 6 analysis will be bounding of MDNBRs in future ' cycles. 7.3 HYDRAULIC CHARACTERIZATION Table 7.1 shows component loss coefficients for both fuel types. - J These loss coefficients are based on pressure drop tests performed in ENC's - I Portable Loop Hydraulic Test Facility. The single phase loss coefficients in ; I Table 7.1 are referenced to the respective bare rod regions in each of the two fuel types. The upper and lower tie plate loss coefficients include l reversible losses due to area change and losses due to simulated upper and l I lower core support plates. l l Figure 7.1 provides a diagrammatic comparison of ENC and CE fuel , i ! j pressure loss coefficients for normal full-power operation. The overall , assembly loss coefficient of ENC fuel differs by less than 4.0% from that of > l CE fuel at St. Lucie Unit 1. Since the core loss is about one third of the j
- Cycle 6 represents the first ENC reload core consisting of 84 ENC Assemblies and 133 CE assemblies. .
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** Cycle 5 is a reference core consisting solely of CE fuel l <
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p O XN-NF-82-81 Revision 1 total loop loss, the impact of an all-ENC core on primary loop flow is less than 0.5%. 7.4 CORE FLOW DISTRIBUTION ANALYSIS The core flow distribution analysis is performed to assess cross-flow between assemblies in the core for use in subsequent MDNBR subchannel I analyses. The core flow distribution analysis is particularly important for mixed fuel loadings where hydraulically different fuel types are coresident in the core. In the analysis each fuel assembly in an octant of the St. Lucie Unit 1 core is modeled as a hydraulic channel. The calculations are performed with the XCOBRA-IIIC computer code (16). Crossflow between adjacent assem-blies in the open lattice core is directly modeled. The single p~ nase loss coefficients given in Table 7.1 are used to hydraulically characterize the ENC and CE fuel types. To establish limiting assembly mass, energy and momentum crossflows for subsequent MDNBR subchannel analyses, two separate core flow distribution calculations are made. In the first, an ENC fuel assembly is assumed to be limiting during Cycle 6 with a significantly higher power level relative to the remainder of the core. In the second calculation, a CE fuel assembly is assumed to be limiting during Cycle 5 operation with a high assembly power level. For the present analysis of relative DNB, performance, a maximum Fr limit of 1.70 is assumed and imposed on the limiting assembly power for both the ENC and CE limiting cases. The two coreflow calculations thus establish two limiting assembly axially varying crossflow boundary conditions to be used in respective MDNBR evaluations for ENC ano CE fuel types. m- ~
. )
~26- XN-NF-82-81 Revision 1 Table 7.2 summarizes thermal hydraulic parameters used in the core flow distribution calculations and subsequent MDNBR analyses. The calculations are performed at a 130% overpower condition (1.30 x 2700 MWt) in order to maximize potential differences in limiting assembly MDNBR between ENC and CE f
fuel, and to provide MDNBRs which are representative of those that may be expected during a severe operational transient event (i.e. MDNBRs near the XNB 95:95 correlation limit). Figures 7.2 and 7.3 show the core loading for 1 the ENC fuel limiting and CE fuel limiting cases, respectively. In each case. ' 5% inlet flow maldistributic,3 is assumed for the limiting and surrounding assemblies. 7.5 MDNBR SUBCHANNEL ANALYSIS 1., The MDNBR subchannel analysis uses the XCOBRA-IIIC computer code to - evaluate thermal-hydraulic conditions in each of the subchannels of the ' limiting assembly. Crossflow between rod subchannels is determined in the ( calculations. Figures 7.4 and 7.5 show the 1/8 assembly subchannel model used for the ENC and CE MDNBR analyses, respective 1". The local fuel rod power distributions shown in Figures 7.4 and 7.5 are conservatively flat and are representative of the power distributions for each fuel type at their respective average discharge expost'95 , l l The MDNBR subchannal 'e t , is includes factors to account for , manufacturing uncertainties . a :,s fication effects. Specifically a 3% ( i engineering factor is applied to the limiting rod power. This factor is to account for fabrication tolerances on pellet diameter, density and enrich- ,! ment. Also included is the tolerance for claddirig diameter. These manuf acturing uncertainties potentially impact heat flux at the limiting DNB location in thc assembly. The 3% engineering factor is applied to both ENC < and CE fuel designs.
XN-NF-82-81
. Revision 1 Fuel densification can result in a shortening of the pellet column length and thus potentially causes an increase in the average linear heat generation rate. In-reactor shortening of the active fuel column length is conservatively evaluated from the following relationship for anisotropic densification:
i
- ' AL=Ao L 2 where AL = decrease in fuel column length L = fuel column length Ap = fractional density change With an upper bound on density change of 3.5%, the increase in heat flux would be 1.75%. The fuel column length decrease due to densification is -
compensated by an increase in column length due to thermal expansion. This -- length increase is approximately 1% of the active column length. Thus, the net effect due to axial densification and thermal expansion results in a .75% increase in heat flux. In the present analysis, a conservative 1.0% heat flux penalty has been applied to allow for densification. Interassembly crossflow between the limiting and adjacent fuel assemblies was accounted for in thermal margin calculations by imposing an axially varying crossflow boundary condition on the subchanr el model used to. determine MDNBR. Mass, energy and momentum crossficws were established from ! core flow distribution analyses for both ENC and CE limiting cases. Reference 17 describes in detail the application of ENC's thermal margin methodology for pressurized water reactors. . The XNB correlation with correction factors for non-uniform heat flux profile is used to determine the margin to boiling transition for both 1
f
. .L 1
-28 XN-NF-82-81 Revision 1 I ENC and CE fuel assemblies.
Conditions for the present thermal margin analysis are summarized in Table 7.2. The important factor is that the calculations have been performed for 130% overpower. The relative impact of ENC fuel assemblies on existing core thermal ' margins is about an 6.0% decrease in MDNBR relative to results for a limiting CE assenibly in Cycle 5. In the present analysis, the limiting ENC assembly is assumed to consist of 164 ac'.tve fuel rods, while the CE limiting assembly
\
is composed of 168 active rods. Due to a slight difference in the number of ( active Nel rods in Cycle 6 and Cycle 5, ENC rods have about a 1.0% higher heat flux which yields a proportionate decrease in MDNBR. The remaining 6% difference in MDNBR between ENC and CE fuel is due to hydraulic dissimilar- , ities between the two fuel designs. 1 i 7.6 ROD B0W In accordance with ENC rod bow methodology (ll,18), the magnitude of j rod bow for St. Lucie Unit l' fuel was estimated. The calculations indicate , that 50% closure of rod-to-rod gap occurs at an assembly exposure of about 85,000 MWD /MTM for' ENC's 14x14 design. Significant impact to MDNBR due to rod < bow does not occur until gap closures exceed 50%. Since the maximum design exposure for ENC fuel in the St. Lucie Unit 1 core is significantly less, rod ! bow does not limit MDNBR for ENC fuel. A further consequence of the small i amount of rod bow for ENC fuel is that total power peaking is not ; significantly impacted. - 1 1
1 1' . t XN-NF-82-81 Revision 1 7.7 FUEL CENTERLINE TEMPERATURE The power level required to produce centerline melt in zircaloy clad urania fuel rods is about 21 kW/ft. Loss-of-coolant accident consider-i ations for St. Lucie Unit I limit the steady state peak LHR to 15 kW/ft. The i 40% margin between 21 kW/ft and 15 kW/f t is large enough that fuel centerline melt is not a limiting factor for anticipated operational transients. ( 4 e S 4 e a O h- - _- ____m________.-_____.-_________ _________ __m.___ __ _ _ _ - - _ - _ _ _ _ . _ _ _ _ . . _ _
Table 7.1 Assembly Component loss Coefficients i ENC _CE Lower Tie Plate
- 5.197 Re .0.0356 3.55 Spacer 1.752 Re-0.067 1.019 Re-0.04 Upper Tie Plate
- 7.42 7.45 Bare Rod Friction 0.1987 Re-0.20 0.1987 Re-0.27
- Includes reversible losses due to change in area and losses due to core support hardware.
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XN-NF-82-81
. Revision 1 L
Table 7.2 Thermal-Hydraulic Design Data j Operating Conditions Rated Power (Core 2700 MWt ,i
- Fraction of Heat Generated in Fuel .975 Pressurizer Pressure 2250 psia
't Core Inlet Temperature 549.00F f Total Reactor Coolant Flow 139.4 x 106 -lb/hr i Active Coolant Flow 134.8 x 106 lb/hr Limiting Assembly Peaking Factors Axial 1.60 , Engineering 1.03 Fr 1.70 l Total Nuclear 2.80 ii l l - l e e I I
XN-NF-82-81 Revision 1 4 25.0 - - I i 21.05 20.27 ' 20.0 - -
,, KUTP=7.42 t 25 KUTP=7.45 a
I C
.2 15.0 - -
.2 is '
E3 K5 =9 x .724 ; m jj 10.0 - - KF = 3.86 K3=9x .E01 ! t KF = 3.86 i i 5.0 - - ' i l ( ' l l l l KLTP=3.25 KLTP=3.55 j ; 1 . 1 ENC C.E. . i i Re = 5.3 x 105 ! I 4 Figure 7.1 Comparison of ENC and C.E. - Component loss Coefficients l ) 1 1 i i
XN-NF-82-81 Revision 1 CE CE CE ENC CE CE ENC ENC ~ 0- - 8 -- 16 - - 4 -
-0 -
-0 - - 12 -
0.70 0.88 1.11 1.24 1.26 0.92 1.15 1.08 ENC CE CE CE CE CE ENC CE O 79 0 0 0 0 0 16 8 1.06 s 1.01 1.21 1.06 1.23 1.12 0.88 CE
' 12 ENC CE ENC CE CE ENC 0.34 12 0 12 0 8 0 1.20 s 1.00 1.22 1.04 1.05 1.17
' ENC CE ENC
- CE ENC '
12 0 12 0 0 1.18 s 1.00 1.25 1.20 0.98 ,
' ENC CE ENC CE 12 8 4 12 1.21 s 1.04 1.07 0.34
' ENC CE
- Limiting Assembly 0 12
- 1.11 0.37 x
5 xx Fuel Type x No. of B 4C Pins xxx Radial Assembly Power Factor . Figure 7.2 Cycle 6 1/8 Core Model N
d l l XN-NF-82-81 Revision 1 CE CE CE CE CE CE CE CE 0-- 12 --0 -
-U --0 --
0-- 0-- 12 0.70 s 0.88 1.11 1.24 1.26 0.92 1.15 1.08 CE s 0 1 CE CE CE CE CE CE CE 0.79 8 12 8 0 0 0 8 1.06 s 1.01 1.21 1.06 1.23 1.12 0.88 CE j
' CE CE CE CE CE CE 0.34 0 0 0 12 12 0 1.20 s 1.00 1.22 1.04 1705 1.17
' CE CE CE* CE CE -
0 0 8 12 0 1.18 x 1.00 1.25 1.20 0.98 -
' CE CE CE CE O 12 0 0 1.21 s 1.04 1.07 0.34
- Limiting Assembly
' CE CE 4 0 1.11 0.37
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xx Fuel Type { x No. of B 4C Pins i xxx Radial Assembly Power Factor - Figure 7.3 Cycle 5 1/8 Core Model 1 (
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XN-NF-82-81
- Revision 1 I
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l 3 l 1 i XN-NF-82 ., . Revision.1 j
- l 8.0 ACCIDENT AND TRANSIENT A.'!ALYSES t<
8.1 PLANT TRANSIENT ANALYSIS
- Plant transient analyses for the ENC fuel that is being placed
- . in St. Lucie Unit 1 this cycle are reported in Reference 3. Thermal i
!! margins from this analysis are applicable to mixed core loadings of ENC and the resident Combustion Engineering fuel.
- f 8.2 ECCS ANALYSIS
!f The LOCA-ECCS analysis for ENC fuel at St. Lucie Unit 1 is reported in Reference 4. This analysis is' applicable to ENC fuel for mixed core loadings of ENC and the resident Combustion Engineering fuel
- at St. Lucie Unit 1. The analysis remains valid so long as any NSSS modifications and changes in system operational parameters continue. to be bounded by the analysis.
8.3 R0D EJECTION ANALYSIS ~
- , A Control Rod Ejection Accident is defined as the mechanical-failure of a control rod mechanism cressure housing, resulting in the ejection of a Control Element Assembly (CEA) and drive shaft. The co,nse-quence of this mechanical failure is a rapid rnctivity insertion together I
with an adverse core power distribution, possibly leading to localized - fuel rod damage. The rod ejection accident has been evaluated with the procedures developed in the ENC Generic Rod Ejection Analysis (5). The ejected rod worths and hot pellet peaking factors were calculated using the XTGPWR(9)-
.w- ,-,g--..-m.__ ..-y-- . _ - _ . . _ - - - - . _ ,
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l
. 4 XN-NF-82-81 Revision 1 code. No credit was taken from the power flattening effects of Doppler or moderator feedback in the calculation of ejected rod worths or resultant peaking factors. The calculations made for Cycle 6 using a full core i XTGPWR model were two-dimensional. The total peaking factor, F , were determined as the product of radial peaking (as calculated using XTGPWR) and a conservative axial peaking factor. The pellet energy deposition s
resulting from an ejected rod was conservatively evaluated explicitly for B0C and E0C conditions. The HFP pellet energy deposited was calculated to be 151.3 cal /gm at BOC and 144.1 cal /gm at E0C. the HZP pellet energy deposition was calculated to be less than 25 cal /gm for both BOC and E0C conditions. The rod ejection accident was found to result in an l energy deposition of less than the 280 cal /gm limit as stated in Regulatory Guide 1.77. The significant parameters for the analyses, along with f the results, are summarized in Tables 8.1 and 8.2. A conservative core pressure surge associated with the CEA ejection is calculated in the ENC Generic Rod Ejection Analysis (5) to be 150 psia. Since the nominal core pressure for St. Lucie Unit 1 is i ! 2250 psic, this indicates a maximum pressure of 2400 psia. The primary ( ( coolant system pressure will not exceed the limit of 2750 psia. -- ! I l
. . . . _ . 1 -- __ _._ _. _
Table 8.1 St. Lucie Unit 1, Cycle 6, Ejected Rod Analysis,ilFP , B0C E0C Contribution (a) to Contribution (a) to Energy Deposition, Energy Deposition, Value (cal /gm) Value (cal /gm) A. Initial Fuel Enthalpy (cal /gm) 76.9 75.5 B. Generic Initial Fuel Enthalpy (cal /gm) 40.8 40.8 C. Deltr Initial Fuel Enthalpy (cal /gm) 36.1 36.1 .34.7 34.7 D. Maximum Control Rod Worth (pcm) 100 120 116 121 E. Doppler Coefficient (pcm/0F) -1.2 .95(b) -1.4 .89(b)
, F. Delayed Neutron Fraction, S .0058 1.02(b) .0050 .04(b) g G. Power Peaking Factor (c) 3.0 --
3.2 --
- 11. Power Peaking Factor Used (d) 4.0 --
4.0 -- Total Fuel Enthalpy cal /gm 151.3(c) 144,1(c) (a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy depo-sition contribution values and factors are derived from data calculated in the " Generic Analysis of the Control Rod Ejection Transient...." document. i (b) These values are multiplication factors applied to (C+D). (c) A conservative axial peaking factor of 1.50 is used in conjunction with a two dimensional fM calculation of the radial peaking factor for the Ejected Rod. The calculations are performed 5.a with the feedbacks turned off. Yl.7' El0 (d) ,The energy deposition due to maximum control rod worth is a function of the power peaking factor.~ & i (e) Total pellet energy deposition (cal /gm) calculated by the equaticn - Total (cal /gm) = (C+D) (E) (F) ; l l l
Table 8.2 St. Lucie Unit 1, Cycle 6, Ejected Rod Analysis, HZP . BOC E0C Contribution (a' .o Contribution (a) to Energy Deposition, Energy Deposition, Value (cal /gm) Velue (cal /gm) l A. Initial Fuel Enthalpy (cal /gm) 16.7 16.7 B. Generic Initial Fuel Enthalpy (cal /gm) 16.7 16.7 C. Delta Initial fuel Enthalpy (cal /gm) 0 0 0 0 D. Maximum Control Rod Worth (pcm) 214 20 292 20 E. Doppler Coefficient (pcm/0F) -1.0(e) 1.03(b) -1.0(e) 1.03(b) F. Delayed Neutron Fraction,8 .0058 1.05(b) .0050 1.20(b) , a G. Power Peaking Factor (c) 7.1 -- 8.8 --
?
H. Power Peaking Factor Used(d) 9.0 -- 9.0 -- Total Fuel Enthalpy cal /gm 21.6(f) 24.2(f) (a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy depo-sition contribution values and factors are derived from data calculated in the " Generic Analysis of the Control Rod Ejection Transient...." document. , (b) These values are multiplication factors applied to (C+D). (c) A conservative axial peaking factor of 2.0 is used in conjunction with a two dimensional calculation of the radial peaking factor for the Ejected Rod. The calculations are performed with the ':edbacks E E turned off. (d} The energy deposition due to maximum control rod worth is a function of the power peaking factor. a
~
(e) For this Dopplet coefficient a conservative value of -1.0 m s assumed at BOC and E0C. , (f) Total pellet energy deposition (cal /gm) calcuIated by the equation - Total-(cal /gm) = (C+D) (E) (F) . O
- i. .
XN-NF-82-81 Revision 1 9.0
SUMMARY
OF OPERATING LIMITS Operating limits for the St. Lucie Unit I nuclear plant are summarized I below. Exxon Nuclear Company methods of analysis for determining or j verifying the operating limits are detailed in Subsection 9.5 and References 12, 13, and 14. Results of the ENC statistical analyses indicate that operating limits previously established for St. Lucie Unit 1 are adequate for Cycle 6 operation with ENC reload fuel present in the core. 9.1 REACTOR PROTECTION SYSTEM The reactor protection system (RPS) is designed to assure that the reactor is operated in a safe and conservative manner. The i input parameters for the RPS are denoted as limiting safety system settings (LSSS). The values or functional representation of the LSSSs are calculated to ensure adherence to the specified acceptable fuel 1 design limits (SAFDLs) during steady state and anticipated operational lf occurrences (A00s). The safe operation of the reactor is also maintained by restricting reactor operation to be in conformance with the limiting conditions for operation (LCOs) which are administratively applied at j the reactor plant. The LSSS and LC0 parametric values are presented in the following sections. . 9.2 SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS The specified acceptable fuel design limits (SAFDLs) are experimentally or analytically based limits on the fuel and cladding which preclude fuel damage. These limits may be exceeded neither during - 5
--mr -
--y .-
o . XN-NF-82-81 Revision 1 steady-state operation nor during A00s. The SAFDLs are used to establish-l the reactor setpoints to ensure safe operation of the reactor. The $ specific SAFDLs used to astablish the setpoints are: o The local power density (LPD) which coincides with fuel centerline melt. o The MDNBR corresponding to the accepted criterion which protects against the occurrence of DNB. The LPD limit for St. Lucie Unit I has been 21 kw/ft in prior , c3cles and this limit is being retained for Cycle 6. It is noted that v l ENC reload fuel for Cycle 6 dces not contain gadolinia-bearing fuel l rods and, therefore, will not exhibit any significant change in either 4 fuel operating temperature or fuel melt temperature relative to prior CE fuel. The ENC critical heat flux correlation, XNB, was used in the ENC thermal margin analysis with statistical parameters corresponding to an upper 95/95 value of 1.22 which is conservative relative to the , 95/95 limit for XNB. Observance of the limiting conditions for operation will protect against DNB with 95% probability at a 95% confidence level during an A00. - 9.3 LIMITING SAFETY SYSTEM SETTINGS f 9.3.1 Local Power Distribution Control The local power distribution (LPD) trip limit is the locus of the limiting values of core power level versus axial shape
v- . r . . w = e XN-NF-82-81 L; Revision 1 Y =
- index that will produce a reactor trip to prevent exceeding the 21 kw/ft LPD limit. The correlation between allowed core power level and b- -
peripheral axial shape index, ASI, was determined using ENC methods which take into c. count the total calculated nuclear peaking and the measurement and calculational uncertainties associated with power peaking. The LPD barn for operation at 2700 MW th is shown in Figure 9.1 as a locus of power and ASI pairs which is conservatively bou.nded by the ENC calculated power and ASI pairs. In this figure ASI is de~ fined as the difference between the bottom core power reading and the top core reading divided by the sum of the top and bottom readings. 9.3.2 Thermal Margin / Low Pressure The thermal margin / low pressure (TM/LP) trip protects l against the occurrence of DNB during steady state operations and for many, but not all, A00s. This reactor trip system monitors primary
/ system prer.sure, core inlet temperature, core power and ASI and a reactor trip occurs when primary system pressure falls below the computed limiting core pressure, Pvar. As with the LPD trip ENC has used its statistical setpoint methodology to verify the adequacy of the existing Tli/LP trip for Cycle 6. The ENC methodology for the TM/LP trip accounts for uncertaint.ies in core operating conditions, XNB (the ENC critical heat flux correlation) uncertainties and uncertainties in power peaking. The TM/LP existing trip function for operation at 2700 MWt, which was verified by ENC, is given by: -
l XN-NF-82-81 Revision 1 var = 2061*A1(ASI)*QR1(Q)+15.85* Tin-8950, P where Q is the higher of the thermal power and the nuclear flux power, T in is the inlet temperature in OF and Al and QR1 are shown in Figures 9.2 and 9.3, respectively. 9.3.3 Additional Trip Functions In addition to the LPD and TM/LP trip functions, other reactor system trips have been determined to provide adherence to reactor system design criteria. The setpoints for these trips, shown ! in Table 9.1, are unchanged from the Cycle 5 values. - 9.4 LIMITING CONDITIONS FOR OPERATION 9.4.1 DNB Monitoring , The validity of the existing LC0 for allowable core power as a function of ASI was verified to ensure adherence to thermal-hydraulic fuel design limits during a postulated CEA drop and loss-of- , flow operational occurrences. The ENC statistical analysis accounted for the effects of uncertainties associated in core operating parameters, the XNB critical heat flux correlation, and power peaking. The allowed core power as a function of ASI for~the existing LC0 is shown to be conservatively bounded by the present analysis in Figure 9.4. ' 9.4.2 Linear Heat Rate Monitoring In the event that the in-core detector system is , not in operation for an extended period of time, the peak linear heat i rate will be monitored through the use of a linear heat rate LCO. The
~
l l
i' h! O O XN-NF-82-81 Revision 1 i verification of this LC0 was performed in a fashion similar to that used in verifying the LPD limiting safety system setting (Section 9.3.1) 4 and the verification is shown in Figure 9.5. The Linear Heat Rate LC0 j limits core power so that an LPD of 15 kw/ft is not exceeded. This LPD I value is based on loss-of-coolant (LOCA) considerations. 9.5 SETPOINT ANALY5IS 9.5.1 Limiting Safety System Set'.inq: Local Power Distribution The letal power distribution (LPD) trip monitors core power and ASI in order to initiate a reactor scram which precludes exceeding fuel centerline melt conditions. In the analysis for this trip function 1374 axial power distribution cases were examined to establish bounding values of total power peaking, Fg , versus ASI. These cases were generated in a manner consistent with that discussed in Reference 12. ENC statistical methods were then employed to account for the uncertainties in the parameters that are given in Table 9.2. The peak linear heat rate in the core occurs at the position of the maximum total peaking factor, Fg. The maximum total peaking factor, F g, is r eferred to as the " hot spot" in the core. It is the ratio of the maximum linear heat generation rate in the cort to the average linear heat generation rate in the core. F gis, therefore, the product of [y (the ratio of the power of the peak fuel pin to the l
f XN-NF-82-81
, Revision 1 average fuel pin in the plane) times the core average axial power peaking factor, F 2, at the same axial position. The total peaking factor with uncertainties is, therefore: !'
Fg=Fxy xF'z {- Bounding values are utilized for Fxy, i.e., 1.70 for unrodded planes and 1.87 for rodded planes. The actual allowed power f'or each ASI was calculated , statistically incorporating the uncertainties listed in Table 9.2 as i described in References 13 and 14 to produce the series of points in Figure 9.1. The results in Figure 9.1 which bound the existing St. Lucie Unit 1 LPD trip and thus, verify the adequacy of the existing trip function. Thermal Margin / Low Pressure LSSS l i The TM/LP trip is designed to shut the reactor down ; should the reactor conditions (ASI, inlet temperature, core power and pressure) approach the point where DNB might occur during either normal operatioa or an A00. The present analysis uses ENC's XNB critical heat flux correlation and ENC's statistical setpoint methodology and is consist at with the NRC's Standard Review Plan in requiring DNB be avoided with 95% probability at a 95% confidence level. The uncertainties shown in Tables 9.2 and 9.3 were included in the verification of the TM/LP trip as described in References 13 and 14. The minimum excess margin of protection provided by the trip was 38 psi.
XN-NF-82-81 - Revision 1 i . 9.5.2 Limiting Conditions for Operation DNB Monitoring I The TM/LP trip system does not monitor reactor l coolant flow and does not consider changes in power peaking which do not significantly change ASI. Thus, the TM/LP trip generally does not provide DNB protection for the four pump coastdown and CEA drop A00s. The ENC analysis of these transients is given in References 14 and 15. The LC0 presented here administratively protects the DNB SAFDL for. these transients. The method used by ENC to establish the DNB LCO involved simulations of the CEA drop and the loss-of-flow transients using the core thermal hydraulic code, XCOBRA-IIIC(10) , to determine i
! the initial power, as a function of ASI, which provides protection from DNB with 95% probability. The uncertainties listed in Tables 9.4 and 9.5 were applied using the methodology described in Reference 13. The results of the statistical analyses are summarized by the points in Figure 9.4. The points bound and, thus, verify the adequacy of the
! existing DNB LC0 for St. Lucie Unit 1 which is shown in the same figure by the straight line segments. '
LPD Monitoring The plant technical specifications allow plant operation for limited periods of time if the in-core detectors are out
~
of service. In this situation, the LPD barn provides protection in
--._.~m ~. _
XN-NF-82-81 Revision 1 steady state operation against penetration of the 15 kw/ft LPD limit established by LOCA considerations. ENC statistical methodology for the LPD LCO is essentially the same as that for the LPD LSSS except: _ (1) The peak LPD limit is 15 kw/ft, and . (2) The uncertainties listed in Table 9.4 were used, as opposed to the values in Table 9.2. J I The allowed power versus ASI was statistically
~
analyzed to account for the appropriate uncertainties. The, points in Figure 9.5 represent the statistical calculation of the LHR curve as , described in Reference 13. The existing LC0 curve is shown by the straight line segments in Figure 9.5 and are conservative relative to the ENC calculated points.
XN-NF-82-81 Revision 1 Table 9.1 Additional LSSS Trip Functions Parameter Set Point Uncertainty Low steam generator pressure 600 psia 22 psi . Low steam generator water level 37% 10 in l Variable high power 9.61% of rated 2% (<10% of rated) Low reactor coolant flow 95% 2%
; High pressurizer pressure 2400 psia 22 psi Asymmetric steam generator pressure 135 psia 22 psi .
High containment pressure 3.3 psig --
.t a
O e
--.m u u
d XN-NF-82-81
' Revision 1 -l 1.
Table 9.2 Uncertainties Applied for LPD Trip Calculation Socrce Value* Engineering tolerance jt0.03 Peaking uncertainty + 0.085 Trip processing & decalibration - 0.01 of rated
- 0.05 ASI uncertainty + 0.06 i
.. 'l l
I i The distributions are treated as normal and the uncertainty range represents + 2o values.
~
}
l I i
'l il XN-NF-82-81 Revision 1 Table 9.3 Uncertainties Applied to Only the DNBR Calculation
, Source Value*
i l' Pressure Measurement + 126.66 psi
; Trip bias -30 psi Inlet coolant temperature + 20F i XNB correlation (10) 1.261 t
0.843 Flow measurement ! 0.036 of rated 4
'I
- The distributions are treated as normal and the uncertainty range represents + 2e values. -
XN-NF-82-81
. Revision 1 Table 9.4 Uncertainties Applied for the LCO Based on LPD Source Value*
Engineering tolerance 1 0.03 Peaking uncertainty 1 0.085 Power measurement 2 0.02 of rated ; ASI uncertainty 1 0.05 1
- The distributions are treated as normal and the uncertainty range represents 1 2e. ,
Y
XN-NF-82-81 Revision 1
~ .t. <} Table 9.5 Uncertainties Applied to LC0 Based on DNB 'i' Source Value*
CEA Drop Lpss of Flow
- )
Pressure measurement j; 30 psi j; 22 psi
; Inlet coolant temperature + 280F + 20F XNB correlation (10) 1.261 1.261 0.841 0.841 Flow measurement j; 0.036 of rated j; 0.036 of rated ,
Scram delay -- j; 0.15 seconds ,, Trip setpoint -- j; 0.02 of rated flow Scram speed -- j; 8.35 inches /second Rod worth -- j;0.5% 1 Flow coastdown -- j; 0.098 i seconds l
- The distributions are treated as normal and the uncertainty range represents + 2a.
t Time for flow to drop from 90% to 80% of rated flow. 1 l l
o , k' Ea" "y$ E1EE" 0 7, 4 _. 1
) : .
,0 m.
3
,. ,0 0
0
. 1, 2
_- 2 o ,0 g X n k E r d D a B e l N D I 1 P
. L
,0 E 1 o P s
A t H i e n i t o r S U 7 n 1 e i a "o 1, ,00.A. L i c t 0 I u r L e b X c A t n U S f
" L A
o n o h i p ,0 1 R E 1 9 e i , T r H r t a 5 0 P u g n #' 0 g I i b m o g 1, 5 4 i t n s 2. PE R i F C 1 i
,0
- . x - E l - a i c - t s i t 3 a ,0 t S -
"d 0 ,
a 7, 4
"P
_ - 1 Fr . 0 _ 0 0 0 0 0 0 0 6 1 2 0 8 6 1 _ 1 1 1 3 5
- a$a(L at-d r oJ3_.-0 nmm)._. ( '
1
L b XN-NF-82-81 Revision 1
. e
-d m
-d e
- ! -d l n E
. -d 8 -i n b
_d "-
.E X h
= zg l1
- z >-
ou-
-dQ- ca z: e.
~
tn " o
.-g g ..
e mca l Ng x '.a
-dI "
l l N n 5
-d E i =
.?
-d4 m
-da .
l o e o m o m o m - n n n n - - o N N N N N N l (19) NOI1ONOJ NOIl038800 JIMI dVWI l e
, - - . _ . - - _ _ . -, - - n -
- 1. 2 - 120.0,1.2 9
a": , o 1-Z .97.2, 972 O 6-.
. o 4 y 0. 8 -
78.1,'.863'
.g n_ g,
- s z ?
i3 o [ O b lao O L3
= (Y- 0. 6 - . a:
8 o o Q_ e- e a: Y 0.1 - o_ E a
, s E
E 5-' a 0,. 23s $'y g 0.2 , , , , , , , , , , , , E? O 10 20 30 10 50 60 70 80 30 100 110 120 8$ POWER (PERGENT OF RATED) -s-d Figure 9.3 St. Lucie Unit 1 TM/LP Trip Function QR1 - h 1 _ _ . . _. .
- 1. 3 -
O CEA Drop-statistical combination of uncertainties O Loss of flow O
- 1. 2 s cn O z 8 0 0
1 1.1 - 0 0 0 ~ D D O E ta o .08,1 .15,1 ' 3 1- O
^ ^
g
. CL O !
ta 2 0 0. 9 - ., 3 ' J C Existing LC0 + o t O J 0. 8 - . l o.7 . , , , , , , , , ,. == ii l'
-0.30 -0.25 -0.20 -0.15 -0.10 -0.05 0.00 0.05 0.10 0.15 0.20 0.25 0.30 {g j
PERIPHERAL RXIRL SHRPE INDEX ..g Figure 9.4 St. Lucie Unit 1 Limiting Condition for Operation DNB Barn
^i 1
- a
110 - O Statistical Combination of Uncertainties O % Q O 100 - O D cp mg D[0 9 O cf3 cp D D Y U% D gO Eb O O o cP O 00 ((f3 b OOCD [D Dgo 90 -g 00
, O Ocb o cfD Q .05,.82 .15,.82 cf3 g 80 - E' L_1 2
o Q_
~
@ Existing LC0 2
o _J Q 60 - ( -.3,.58 .3,.58 i o
- O
._ )
50 -
$Y 10 h , , , , , , , , , , ,
[f
-0.30 -0.25 -0.20 -0.15 -0.10 -0.05 0.00 0.05 0.10 0.15 0.20 0.25 0.30s 8$
PERIPHERAL RXIRL SHAPE INDEX ~s Figure 9.5 St. Lucie Unit 1 Limi, ting Condition for Operation LHR Barn ,'
- n - - . e
-- '+ ~
w ~ --
c XN-NF-82-81 Revision 1
10.0 REFERENCES
- 1. XN-NF-82-09, " Generic Mechanical Design Report Exxon Nuclear 14x14 Fuel Assemblies for Combustion Engineering Reactors",
November 1982.
! 2. XN-NF 32-97, "St. Lucie Unit 1 Addendum to " Generic Mechanical Design Report Exxon Nuclear 14x14 Fuel Assemblies for Combustion Engineering Reactors", December 1982.
- 3. XN-NF-82-99, " Plant Transient Analysis for St. Lucie Unit 1 Reactor", Exxon Nuclear Company, to be issued.
- 4. XN-1F-82-98, St. Lucie Unit 1 LOCA Analysis Using EXEM/PWR ECCS model", Exxon Nuclear Company, to be issued.
- 5. XN-NF-78-44, "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors", Exxon Nuclear Company, January 1979. ,
- 6. XN-75-27, " Exxon Nuclear Neutronics Design Methods for Pressurized
- Water Reactors", Exxon Nuclear Company, June 1975.
- 7. XN-75-27, Supplement 1, September 1976.
- 8. XN-75-27, Supplement 2, December 1977.
- 9. XN-CC-28, Revision 5, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing", Exxon Nuclear Company, July 1979.
- 10. XN-NF-621(P), Rev. 1, "XNB DNB Correlation for PWR Fuel Designs",
Exxon Nuclear Company, April 1982.
- 11. XN-75-21(t), Supplement 4, " Computational Procedure for Evaluating Fuel Rod Bowing", Exxon Nuclear Company, May 1982. ,
- 12. XN-NF-507, Rev. 1, " ENC Setpoint Methodology fo- CE Reactors",
July 1980.
- 13. XN-NF-507, Supplement 1 (P), " ENC Setpoint Methodology for CE Reactors, Statistical Setpoint Methodology", Septen.ber 1982. ,
]
. . - e 1
XN-NF-82-81 Revision 1 I
- 14. XN-NF-507, Supplement 2 (P), " ENC Setpoint Methodology for CE )
Reactors, Sample Program", November 1982.
- 15. Xh-74-5, Revision 1, " Description of the Exxon Nuclear Plant Trinsient Simulation Model for Pressurized Water Reactors (PTSPWR),1975.
- 16. XN-NF-75-21(P), Revision 2, "XCOBRA-IIIC: A Computer Code to Determine the. Distribution of Coolant During Steady-State and Transient Core Operation" . Exxon Nuclear Company, September 1982. l 4
- 17. XN-NF-82-21(P), Revision 1 " Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configu-
- rations", Exxon Nuclear Company, September 1982. {
- 18. XN-75-21(NP), Supplements 1 through 3, " Computational Procedure ,.
for Evaluating Fuel Rod Bowing", Exxon Nuclear Company, July $ 1979. I
- 19. AN-NF 83-01, " Exxon Nuclear Analysis of Power Distribution Measurement Uncertainty for St. Lucie Unit 1", Exxon Nuclear .
- Company, to be issued.
, L' t k 1
- i l
l 4 k l r t t .
, ,- ,e XN-NF-82-81 REVISION 1 ISSUE DATE: 1/7/83
, ST.LUCIEUNIT1,CYCLFj _
SAFETY ANALYSIS REPORT DISTRIBUTION KA BRYAN
.; GJ BUSSELMAN GC COOKE RL FEUERBACHER TJ HELBLING JS HOLM SE JENSEN MR KILLGORE .
, JE KRAJICEK ,, TR LINDQUIST JN MORGAN WT NUTi FB SK0 GEN l GA 70FER RB if0VT GN WARD HE WILLIAMSON DOCUMENT CONTROL / 5 j FP&L (50)/GN WARD , i f}}