ML20062K454

Re Ginna Nuclear Plant Cycle 12 SAR for Low Temp & Pressure

ML20062K454
Person / Time
Site:	Ginna
Issue date:	07/20/1982
From:	Copeland R, Krajicek J, Nielsen L SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17256B178	List: ML17256B177 ML17256B179 ML17256B180 ML20062K429 ML20062K454
References
XN-NF-82-57, NUDOCS 8208170206
Download: ML20062K454 (44)

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{{#Wiki_filter:I XN NF 82 57 lg I lI J!I R.E. GINNA NUCLEAR PLANT CYCLE 12 !I SAFETY ANALYSIS REPORT

g FOR LOW TEMPERATURE AND PRESSURE 1

l I JULY 1982 g I RICHLAND, WA 99352 l ERON NUCLEAR COMPANY,Inc. l l

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I XN-NF-82-57 Issue Date: 07/20/82 l R. E. GINNA NUCLEAR PLANT CYCLE 12 SAFETY ANALYSIS REPORT FOR LOW TEMPERATURE AND PRESSURE Drafted: [ ~ M!67 - I [.'. Nielsen, Unit Managdr PWR Neutronics Drafted: gkjA g /jgz. i R. A'. C pela'd, Design Coordinator Drafted: 7/z/sz I 6J.5.[rajicek I /# Drafted: ~W.I.A t'~ Reviewed: je ( #t Tt F.8.SkigO, Manager PWR Ne Prepared: b lldI h [ 2 - R'.B. Stout, Manage [r Neutronics and Fue anagement I ff ~ wL y/;//L Prepared; JI5. Morgan,ganager/ Licensing and/ Safety Engineering Approved: M(hy-y Jc/c3 t G. A. Sofer M r Fuel En inebring and Te[chnical Services Approved: 7[v/h G. J.'Busselman, Manager Fuel Design Concur: .[L Mftc)l/[i, ~ 7/ h H.157 Williamson, Manager Proposals & Customer Services Engineering ERON NUCLEAR COMPANY,Inc.

I l i XN-NF-82-57 I TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

AND

SUMMARY

1 2.0 OPERATING HISTORY OF THE REFERENCE CYCLE............... 3 3.0 CYCLE 12, REVISED TEMPERATURE AND PRESSURE OPERATIONS COMPARED TO OPERATIONS AT NORMAL TEMPERATURE AND PRESSURE CONDITIONS...... 6 4.0 GENERAL DESCRIPTION......................... 11 5.0 FUEL SYSTEM DESIGN......................... 16 5.1 FUEL DESCR IPTION........................ 16 5.2 DES IGN CR ITER I A........................ 16 6.0 NUCLEAR DESIGN........................... 20 6.1 PHYSICS CHARACTERISTICS.................... 21 6.1.1 Power Distribution Considerations.......... 21 6.1.2 Control Rod Reactivity Requirements......... 22 6.1.3 Moderator Temperature Coefficient Considerations... 23 6.2 ANALYTICAL METHODOLOGY..................... 23 7.0 THERMAL HYDRAULIC DESIGN AND PERFORMANCE.............. 30 l 8.0 ACCIDENT AND TRANSIENT ANALYSIS................... 31 8.1 ECCS ANALYSIS......................... 31 g 5 8.2 PLANT TRANSIENT ANALYSIS.................... 31 8.3 R0D EJECTION ANALYSIS FOR R. E. GINNA CYCLE 12......... 32

9.0 REFERENCES

35 I lll i

I ii XN-NF-82-57 I LIST OF TABLES I Table Page 4.1 R. E. Ginna Cycle 12 Fuel Assembly Design Parameters......... 13 5.1 R. E. Ginna Fuel Mechanical Design........... 17 5.2 R. E. Ginna Exposure and Flux History for the Pin With Maximum Discharge Exposure...................... 19 6.1 R. E. Ginna Neutronics Characteristics of Cycle 12 Compared With Cycle 11 Data..................... 25 6.2 R. E. Ginna Control Rod Shutdown Margins and Requirements for Cycle 12............................. 26 8.1 R. E. Ginna Unit 1 Kinetics Parameters................ 33 8.2 Ejected Rod Worth and Peaking Factors................ 34 LIST OF FIGURES Figure Page 2.1 R. E. Ginna Cycle 11 Critical Boron Concentration vs. Exposure, AR0............................ 4 2.2 R. E. Ginna Cycle 11 Power Distribution INCORE vs. PDQ Prediction, 3,600 MWD /MT....................... 5 3.1 R. E. Ginna Cycle 12 Power Distribution INCORE vs. PDQ Prediction, Octant Averaged, 555 MWD /MT............... 8 3.2 R. E. Ginna Cycle 12 Assembly Power Distribution, NTPXTG vs. LTPXTG Calculation, 1,000 MWD /MT, HFP, AR0............ 9 3.3 R. E. Ginna Cycle 12 Critical Boron vs. Exposure, AR0........ 10 4.1 R. E. Ginna Cycle 12 Loading Pattern................. 14 4.2 R. E. Ginna 80Cl2 Quarter Core Exposure Distribution and 15 Region ID..............................

I

I iii XN-NF-82-57

I i

LIST OF FIGURES (CONTINUED) Figure Page I 6.1 R. E. Ginna Cycle 12, Critical Baron vs. Exposure LTP, ARO...... 27 l 6.2 R. E. Ginna XTG Assembly Relative Power, Cycle 12, LTP, O MWD /MT, AR0, HFP, (E0C12 = 7,100 MWD /MT).............. 28 iE 6.3 R. E. Ginna Cycle 12 XTG Depletion, LTP, HFP,

E 8,700 MWD /MT, 41 ppm.........................

29 iI .I

I

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I iv XN-NF-82-57 I R. E. GINNA NUCLEAR PLANT CYCLE 12 SAFETY ANALYSIS REPORT FOR LOW TEMPERATURE AND PRESSURE PROLOGUE I In the fall of 1982, Rochester Gas and Electric Company (RG&E) is scheduling a reduction in both coolant temperature and system pressure for the R. E. Ginna plant. This report addresses the safety analysis evaluation for Cycle 12 at low temperature and pressure conditions. Reports supplied previously for Cycle 12 included the neutronic analyses associated with the Preliminary and Final Scheduled Delivery Date Notices, (PWR:023:80 ar.d PWR:003:81), the Cycle 12 Fuel Cycle Design Analyses, XN-NF-81-66 (P), the Cycle 12 Safety Analysis Report at normal conditions, XN-NF-81-94, and the Cycle 12 Startup and Operations Report, XN-NF-82-41 (P). INCORE computer decks required for Cycle 12 operations will be supplied as the cycle progresses. I I I I I I

E g 1 XN-NF-82-57 I R. E. GINNA NUCLEAR PLANT CYCLE 12 SAFETY ANALYSIS REPORT FOR LOW TEMPERATURE AND PRESSURE

1.0 INTRODUCTION

AND

SUMMARY

The R. E. Ginna nuclear plant began Cycle 12 operation in May of 1982 at normal temperature and pressure (NTP) conditions (T = 573.5 F 3g and pressure = 2,250 psia). In the fall of 1982, Rochester Gas and Electric (RG&E) is scheduling a 15 F reduction in temperature and a 250 psia reduction in pressure for the plant. This report addresses the safety analysis evaluation for operation of Cycle 12 at the lower tempera-ture and pressure conditions (LTP). The analysis bounds Cycle 12 operation between cycle exposures of 0 MWD /MT and 9,400 MWD /MT. The R.E. Ginna core contains four (4) regions of fuel supplied by Exxon Nuclear Company (ENC). The Cycle 12 fresh fuel loading consists of 12 ENC assemblies from Region 13 (Batch XN-4 with zircaloy guide tubes) and twelve (12) ENC assemblies from Region 14 (Batch XN-5 with zircaloy guide tubes). The remainder of the core contains 24 once-burnt assemblies with zircaloy guide tubes and 4 once-burnt, 32 twice-burnt, and 33 thrice-burnt ENC assemblies with stainless steel guide tubes in addition to four (4) twice-burnt Westinghouse mixed oxide (MOX) assemblies. The characteristics of the fuel and of the reloaded core result in i 1 conformance with existing Technical Specification limits regarding shutdown I margin provisions and thermal limits. This document provides the neutronic l I

I 2 XN-NF-82-57 i dnd Control rod ejection analyses for the plant during Cycle 12 operation II) with the lower temperature and pressure conditions. The ENC fuel design for Batch XN-5 is very similar to the design of Batch XN-4. The dimensions are unchanged, however, tighter tolerances on the fuel and clad allow Batch XN-5 to go to higher burnup than the previous ENC reload batches. The Cycle 12 analysis for low temperature and pressure conditions is determined to be bounded by the previous Plant Transient Analysis (2,3,4) Likewise, the ECCS analysis (5,6) is also applicable to Cycle 12 operation for low temperature and pressure conditions. The consequences of the rod ejection accident for Cycle 12 at LTP are similar to those cal-I) IO) I9),11(10), and 12(11,12) (NTP). culated for Cycles 8 ,9 , 10 I l I I I i I I I I

I 3 XN-NF-82-57 I 2.0 OPERATING HISTORY OF THE REFERENCE CVCLE R. E. Ginna Cycle 11 has been chosen as the reference cycle with respect to Cycle 12 due to the close resemblance of the neutronic charac-ter. tics between these two cycles. Cycle 11 operation began on June 19, 1981 and was terminated during the February 1982 outage. The core had accrued a Cycle 11 burnup of 7,100 MWD /MT. The Cycle 11 loading included 28 fresh ENC fuel assemblies. Remaining assemblies in the core consisted of 89 exposed ENC assemblies and 4 exposed Westinghouse mixed oxide assemblies. The measured power peaking factors at hot full power, equilibrium xenon conditions, remained considerably below the Technical Specification N limits throughout Cycle 11. The total nuclear peaking factor, F, and g the radial nuclear pin peaking factor F H, remained below 1.75 and 1.48, respectively. Cycle 11 operation was typically rod free with the D control bank positioned in the range of 215 to 218 steps, 225 steps being fully withdrawn. It is anticipated the control bank insertions throughout Cycle 12 will be similar to those in Cycle 11. The critical boron concentration as calculated by ENC for Cycle 11 agreed to within about 35 ppm compared to the observed values (see Figure 2.1). Also the predicted power distributions typically agreed to within +3 percent of the measured values (see Figure 2.2 for typical comparison). I I I

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I 5 XN-NF-82-57 I .906 1.008 1.214 1.072 1.028 .948 .705* .890 .991 1.190 1.048 1.020 .943 .695 +1.8 +1.7 +2.0 +2.3 +0.8 +0.5 +1.4 1.005 1.238 1.036 1.091 1.271 1.148 .639 .991 1.229 1.022 1.080 1.270 1.133 .624 +1.4 +0.7 +1.4 +1.0 +0.1 +0.9 +2.4 1.210 1.032 .946 1.303 1.101 .962 1.190 1.026 .952 1.307 1.097 .954 +1.7 l+0.6 -0.7 -0.3 +0.4 +0.8 1.065 1.090 1.291 .968 1.156 .436 1.048 i 1.086 1.310 .976 1.159 .427 I +1.6 +0.4 -1.5 -0.8 -0.3 -0.3 1.034 l1.273 1.086 1.145 .780 1.020 1.273 1.099 1.159 .788 I +1.4 +0.0 -1.2 -1.2 -1.1 .938 1.132 .948 .433 Measured I .943 1.140 .955 .433 Calculated PDQ -0.6 -0.7 -0.7 -1.1 m-c I x 100 c .680* .613 Peaking .697 .624 Maximum alcula n Me surement l -2.5 -1.7 f 1.596 1.599 g

• Mixed Oxide Fuel F

1.431 1.423 g l 1.310 1.303 i R l Figure 2.2 R. F. Ginna Cycle 11 Power Distribution INCORE vs. PDQ Prediction, 3,600 MWD /MT I I

I 6 XN-NF-82-57 I 3.0 CYCLE 12, REVISED TEMPERATURE AND PRESSURE OPERATIONS COMPARED TO OPERATIONS AT NORMAL TEMPERATURE AND PRESSURE CONDITIONS The R. E. Ginna nuclear plant is currently operating in Cycle 12 at normal temperature and pressure conditions. Comparisons between early Cycle 12 neutronic calculations to measured data at normal temperature and pressure conditions are presented in this section. In addition, calculated data at the revised temperature and pressure conditions for early Cycle 12 operations are compared against the calculated and measured data at normal conditions. The comparisons show a slight increase in assembly power (~2%) in the edge assemblies and an increase of ~20 ppm of dissolved boron with the revised conditions. Cycle 12 operation began on May 25, 1982, and as of June 21, 1982, the core had accrued about 700 MWD /MT of burnup. The Cycle 12 loading includes twenty-four (24) fresh ENC fuel assemblies. Remaining assemblies in the core consist of ninety-three (93) exposed ENC assemblies and four (4) exposed Westinghouse mixed oxide assemblies. The measured power peaking factors at hot full power, equilibrium xenon conditions are considerably below the Technical Specification limits N in Cycle 12. The total nuclear peaking factor, F and the radial nuclear pin peaking factor, F g, are measured to be below 1.71 and 1.46, res-pectively. The calculated values of F" and F at normal temperature g and pressure conditions (NTP) are 1.67 and 1.43, respectively. At the revised low temperature and pressure conditions (LTP) F" and F are g H I I

I 7

XN-NF-82-57 calculated to be 1.71 and 1.44, respectively. These values are very close. The calculated assembly power distribution at NTP and the measured assembly power distribution at NTP is shown in Figure 3.1 at a cycle burnup of 555 MWD /MT. A comparison of the calculated assembly powers between the NTP and LTP core with the ENC core simulator model at 1,000 MWD /MT is shown in Figure 3.2. The power distributions are very similar. The critical boron concentration at NTP as calculated by ENC for Cycle 12 is agreeing to within 16 ppm at 700 MWD /MT. At startup the hot zero power measured critical boron was within 5 ppm of the calculated value. The calculated boron rundown curve at NTP, LTP, and the measured values at NTP are shown in Figure 3.3. I

I I

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I 8 XN-NF-82-57 G F E D C B A .95 1.20 .88 1.12* .99 1.26 .87 .93 1.1/ .87 1.13 1.00 1.27 .87 7 2.11 2.50 1.14 .89 -1.01 .79 0.0 1.21 1.12 .93 .97 1.28 1.04 .68

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.91 .97 1.29 1.03 .67 8 1 2.48 2.68 2.15 0.0 .78 .96 1.47 .88 .92 1.29 1.25 1.17 .93 .87 .91 1.26 1.25 1.17 .93 9 J.14 1.09 2.33 0.0 0.0 0.0 1.12* I .97 1.25 1.15 1.09 .42 1.13 .98 1.25 1.15 1.09 .42 10 .89 -1.03 0.0 0.0 0.0 0.0 .99 1.28 1.16 1.09 .74 Measured 1.00 1.30 1.17 1.10 .74 Calculated 11 -1.01 -1.56 .86 .92 0.0 x 100 1.28 1.00 .87 .42 i 1.27 !1.02 .93 .42 12 I .78 -2.00 -6.90 0.0 Peaking Maximum .87 .67 Calculation Measurement 13 lI .87 .67 .0 0.0 0.0 Q 1.43 1.46

• Mixed 0xide Fuel AH F

1.30 1.28 R 3 Figure 3.1 R. E. Ginna Cycle 12 Power Distribution 5 INCORE vs. PDQ Prediction, Octant Averaged, 555 MWD /MT l I I I

I 9 XN-NF-82-57 I G F E D C B A .978 1.219 .914 1.125* .988 1.235 .885 .972 1.211 .906 1.115 .992 1.249 .899 7 .613 .656 .875 .889 .405 -1.134 -1.582 I 1.235 1.122 .930 .992 1.267 .998 .683 f1.227 1.103 .924 .989 1.274 .997 .692 8 I .648 1.693 .645 .302 .552 .100 -1.318 .915 .931 1.259 1.212 1.132 .973 .907 .924 1.257 1.202 1.129 .976 9 .874 .752 .159 .825 .265 .308 1.125*l .992 1.212 1.107 1.087 .419 1.115 .989 1.203 1.103 1.098 .421 10 f .889 .302 .743 .361 -1.012 .477 .986 1.265 1.132 1.087 .785 .990 1.272 1.128 1.098 .789 11 .406 .553 .353 .947 .510 ~ 1.203 .993 .972 .418 NTP Assembly Power 1.217 .993 .975 .421 LTP Assembly Power 12 -1.164 0 .309 .718 N-L x 100 N .882 .680 Peaking 13 .896 .689 l NTPXTG LTPXTG -1.587 !-1.323 F 1.59 1.60 q

• Mixed 0xide Assembly F

1.42 1.43 AH F 1.11 1.13 7 I Figure 3.2 R. E. Ginna Cycle 12 Assembly Power Distribution, NTPXTG vs. LTPXTG Calculation, 1000 MWO/MT, HFP, AR0 I I I

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I 11 XN-NF-82-57 I 4.0 GENERAL DESCRIPTION The R. F., Ginna reactor consists of 121 assemblies, each having a 14x14 fuel rod array. Each assembly contains 179 fuel rods, 16 RCC guide tubes, and one (1) instrumentation tube. The fuel rods consist of slightly enriched UO2 pellets inserted into zircaloy tubes. The RCC guide tubes and the instrumentation tube in Batches XN-2 and XN-3 are made of SS-304L. Composition of the RCC guide tubes and instrument tubes in Batches XN-4 and XN-5 are zircaloy. Each ENC assembly contains nine (9) zircaloy spacers with Inconel springs; eight (8) of the spacers are located within the active fuel region. Four (4) of the 121 assemblies contain mixed oxide (Pu02 plus UO ) bearing fuel rods. The M0X assemblies 2 consist of three (3) enrichment zones of Pu0 utilizing natural U0 as 2 2 the dilu~ent. The Cycle 12 loading pattern is shown in Figure 4.1 with the assemblies identified by their Fabrication ID's and Region ID's. The initial enrichments of the various regions are listed in Table 4.1. 80C12 exposures, based on an EOC11 exposure of 7,100 MWD /MT, along with Region ID's are shown in Figure 4.2. The core consists of 12 fresh ENC XN-4 assemblies at 3.20 w/o U-235 with zircaloy guide tubes and twelve (12) fresh ENC XN-5 assemblies at 3.30 w/o U-235 with zircaloy guide tubes. A total of twenty four (24) fresh assemblies are loaded in Cycle

12. Remaining assemblies in the core consist of 93 exposed ENC assemblies and four (4) Westinghouse mixed oxide assemblies.

The Cycle 12 core I I

!I l 12 XN-NF-82-57 .I. i I loading has eight (8) Region 11 (XN-2) and 24 fresh assemblies located on the core periphery. This loading pattern satisfies the Technical l l Specification criteria with respect to power peaking while maintaining an adequate shutdown margin. '1iI !,I I I !I i

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m M M M M M M M M M M M M M M M M M M Table 4.1 R. E. Ginna Cycle 12 Fuel Assembly Design Parameters Region 11 12 M0X 12 13 13 14 Enrichment, wt% U-235 3.200 3.450 2.626 3.45 3.20 3.20 3.30 Number of Assemblies 33 32 4 4 24 12 12 Pellet Density, %TD 94.0 94.0 95.0 94.0 94.0 94.0 94.0 Pellet-to-Clad Diametral Gap, Mil 7.5 7.5 7.5 7.5 7.5 7.5 7.5 Fuel Stack Height, in. 142.0 142.0 141.4 142.0 142.0 142.0 142.0 Region Average Burnup t; at BOC12, MWD /MT 26,572 16,857 11,663 5,868 6,673 0 0 Nominal Assembly Weight, KgU 373.78 373.78 395.91** 373.78 373.78 373.78 373.78 Guide Tube Composition SS304L SS304L SS304L SS304L Zr Zr Zr Fuel Supplier ENC ENC W ENC ENC ENC ENC Wt% Pu (Based on Assembly Average) x* In Kg HM e i

14 XN-NF-82-57 I l l M L K J H G F E D C B A P30 P26 P34

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lg 13 13 13 1 M16 Q08 N31 N05 N30 Q04 M37 2 I 11 14 12 12 12 14 11 Q10 P07 N23 P15 M26 P12 N22 PO4 Q12 14 13 12 13 11 13 12 13 14 I MC9 NP3 M01 NO3 N12 P06 MIS lN15 M33 F02 N09 4 12 11 M0X 11 12 12 13 11 11 13 12 002 42 7 N04 j P23 l M17 M07 M36 P22 N14 N26 005 14 12 12 l 13 l 11 11 11 13 12 12 14 P35 N35 P14

10 2 M33 N17 P20 N20 M20 M12 P10 N34 P32 6

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13 12 13 11 11 12 13 12 11 11 13 12 13 QC6 N23 015 Pl7 M34 MOS M19 P24 NO2 N25 QO1 9 14 12 12 13 11 11 11 13 12 12 14 l N01 M03 NP l M11 N13 Nll P01 M40 M13 POS N10 I 10 11 13 12 l 12 11 M0x 11 12 12 13 11 011 P03 N24 Pil M23 P16 N21 P03 009 14 13 12 13 11 13 12 13 14 .139 ' 003 N32 N06 N29 Q07 M14 11 14 12 12 12 14 11 P33' P25 P29 Assembly ID 13 13 13 13 Region ** ENC Reload Fuel by Fractions IDXXX See Table 4.1 for Region Definitions Figure 4.1 R. E. Ginna Cycle 12 Loading Pattern I I

I ~ 0 15 XN-NF-82-57 1 G F E D C B A 29,470 8,862 25,988 11,663 27,132 4,817 0.0 7 11 13 11 M0X 11 12 13 7,710 16,558 26,785 24,397 7,054 18,506 0.0 8 13 12 11 11 13 12 13 25,966 26,787 8,297 14,357 16,170 0.0 9 l 11 11 13 12 12 14 I 11,664 24,379 14,356 20,254 4,670 28,184 0 M0X 11 12 12 13 11 27,155 7,058 16,158 4,669 0.0 11 11 13 12 13 14 6,920 18,499 0.0 28,195 BOC12 Exposure, MWD /MT I 12 12 12 14 11 Region ID I 13 0.0 0.0 i 13 13 I l Figure 4.2 R. E. Ginna B0C12 Quarter Core Exposure Distribution and Region ID l l lI

16 XN-NF-82-57 I 5.0 FUEL SYSTEM DESIGN The previous R. E. Ginna fuel supplied by Exxon are described in references 13, 14 and 15. The first three reloads used stainless steel guide tubes, and the fourth reload used Zircaloy-4 guide tubes. By using as-built dimensions and actual operating history power cycles, the burnup of this fuel was extended to 40,000 MWD /MT. The fifth and sixth ENC reloads for R. E. Ginna are a high burnup design which allows the peak assembly to reach 42,000 MWD /MT. The design report describing the mechanical analyses is given in reference 16. 5.1 FUEL DESCRIPTION The XN-5 and XN-6 reloads (Cycles 12 and 13) are very similar in design to the XN-4 reload. Table 5.1 gives a comparison of the XN-4, XN-5, and XN-6 reloads. The basic dimensions and design are the same. The differences that allow Batches XN-5 and XN-6 to go to higher burnups include: o Control of the cladding contractile strain ratio o Smoother cladding ID surface o Resinter restrictions on the fuel. 5.2 DESIGN CRITERIA Reference 16 gives the mechanical design criteria and the analysis results. The criteria were satisfied for a peak assembly average exposure of 42,000 MWD /MTU and for a peak rod average exposure of 45,000 MWD /MTU. Table 5.3 gives the power history used for the mechanical design.

b 17 XN-NF-82-57 I Table 5.1 R. E. Ginna Fuel Mechanical Design I Fuel Rod XN-4 XN-5, XN-6 Fuel Material U0 Sintered U0 Sintered 2 2 Pellets Pellets Fuel Enrichment, w/o 3.20 3.30, 3.45 Pellet Diameter, (in) 0.3565 0.3565 Dish Volume Per Pellet, (Total %) 1.0 1.0 I Pellet Density, (% TO) 94.0 94.0 Cladding Material Zircaloy-4 Zircaloy-4 Cladding ID, (in) 0.364 0.364 Cladding OD, (in)(l4) 0.424 0.424 Diametral Gap, Cold I Nominal,(in) 0.0075 0.0075 Active Length, (in) 142.0 142.0 Total Rod Length, (in) 149.10 149.10 Number of Active Fuel Rods per UO Bundle 179 179 2 Fuel Rod Array, Square 14x14 14x14 Fuel Rod Pitch, (in) 0.556 0.556 I Spacer Type Zircaloy-4 with Zircaloy-4 with 718 Inconel Springs 718 Inconel Springs Number Per Assembly 9 9 Number Within Active Fuel 8 8 Control Rod Guide Tube Material Zircaloy-4 Zircaloy-4

I 18 XN-NF-82-57 Table 5.1 (Continued) Fuel Assembly XN-4 XN-5, XN-6 Control Rod Guide Tube Dimensions (Upper), (in) 0.541 OD x 0.507 ID 0.541 OD x 0.507 ID Control Rod Guide Tube Dimensions (Lower),(in) First Step 0.479 OD x 0.445 10 0.479 OD x 0.445 ID Second Step 0.475 00 x 0.441 ID 0.475 OD x 0.441 ID Instrumentation Tube Material Zircaloy-4 Zircaloy-4 Instrumentation Tube Dimensions,(in) 0.424 00 x 0.346 10 0.424 OD x 0.346 ID Spacer Outside Dimensions, (in) 7.763 x 7.763 7.763 x 7.763 Fuel Assembly Pitch, (in) 7.803 x 7.803 7.803 x 7.803 Length Between Tie Plates, (in) 150.665 150.665 Total Assembly Length, (in) 160.130 160.130 l l I

19 XN-NF-82-57 I Table 5.2 R. E. Ginna Exposure and Flux History For The Pin With Maximum Discharge Exposure Cumulative I Pin Flux Cycle Irradiation Cumulative Peak Pin Pin > 1 MeV Exposure Time Assy. Exp. Exposure LHGR 13 2 Cycle GWD/MT EFPH GWD/MT GWD/MT kw/ft 10 n/cm sec 12 0 0 0 0 8.64 7.92 1 716 1.133 1.438 8.44 7.81 2 1431 2.261 2.863 8.43 7.83 4 2862 4.521 5.713 8.35 7.83 6 4293 6.778 8.535 8.24 7.79 I 8 5724 9.019 11.320 8.15 7.77 8.8 6296 9.909 12.421 8.12 7.76 9.2 6583 10.354 12.971 8.11 7.76 13 0 0 10.354 12.971 7.63 8.18 1 716 11.568 14.252 7.56 8.12 2 1431 12.778 15.530 7.54 8.12 I 4 2862 15.194 18.080 7.51 8.11 6 4293 17.595 20.617 7.46 8.08 8 5724 19.983 23.137 7.42 8.06 9.2 6583 21.407 24.642 7.39 8.04 14 0 0 21.407 24.642 6.93 8.26 1 716 22.683 25.831 7.06 8.41 I 2 1431 23.841 27.022 6.98 8.32 4 2862 26.131 29.380 6.89 8.21 6 4293 28.392 31.709 6.87 8.19 I 8 5724 30.648 34.032 6.87 8.18 9.2 6583 32.000 35.426 6.86 8.17 I 15 0 0 32.000 35.426 5.61 7.48 1 716 32.977 36.395 5.79 7.67 2 1431 33.961 37.372 5.77 7.64 4 2862 35.923 39.319 5.77 7.62 I 6 4293 37.887 41.267 5.82 7.65 8 5724 39.866 43.230 5.87 7.68 9.2 6583 41.019 44.400 5.90 7.69 a 7298 41.983 45.377 5.79 7.67 I I I

20 XN-NF-82-57 I 6.0 NUCLEAR DESIGN The neutronics characteristics of the Cycle 12 core are quite similar to those of the Cycle 11 core (see Section 6.1). The nuclear design bases for the Cycle 12 core are as follows: 1. The design shall permit operation within the Technical Specifi-cations for the R. E. Ginna plant. 2. The length of Cycle 12 shall be determined on the basis of an E0Cll length of 7,100 MWD /MT. 3. The Cycle 12 loading pattern shall be optimized to achieve power distributions and control rod reactivity worths according to the following constraints: a) The peak F shall not exceed 2.32 and the peak F shall g AH not exceed 1.66 (including uncertainties) in any single fuel rod through the cycle under nominal full power operation conditions with either NTP or LTP conditions. b) The scram worth of all rods minus the most reactivity ( shall exceed the 80C and E0C shutdown requirements. 4. The Cycle 12 core shall have a negative power coefficient. The neutronic design methods utilized to ensure the above requirements are consistent with those described in References 17, 18, and 19. l I I l

I 21 XN-NF-82-57 I 6.1 PHYSICS CHARACTERISTICS The neutronic characteristics of the Cycle 12 core at LTP are compared with those of Cycle 11 and Cycle 12 at NTP in Table 6.1. The data presented in the table indicate the neutronic similarity between Cycle 11 and Cycle 12 at NTP and Cycle 12 at LTP. The Cycle 12 loading pattern is applicable for Cycle 12 operation at LTP and NTP and is based on the actual Cycle 11 shutdown burnup of 7,100 MWD /MT. The calculated boron letdown curve for Cycle 12 at LTP is shown in Figure 6.1. The curve indicates a 80C12, no xenon, critical boron concentration of 1,198 ppm. At 100 MWD /MT, equilibrium xenon, the critical boron concentration is 873 ppm. The Cycle 12 length is projected to be I 8,800+300 MWD /MT with 31 ppm of boron at E0C. 6.1.1 Power Distribution Ccnsiderations Representative predicted power maps for Cycle 12 at LTP are shown in Figures 6.2 and 6.3 for 80C and E0C conditions, respectively. The power distributions were obtained from a three-dimensional core simulator model(20) with moderator density and Doppler feedback effects incorporated. For the projected Cycle 12 loading pattern at LTP the calculated 80C N N nuclear power peaking factors, F, p H, and F, are 1.711, 1.436, and 1.195, respectively. At E0C conaitions the corresponding values are 1.524, 1.370, and 1.104 respectively. The Technical Specification limits relative to F" and F H, with the measurement uncertainties backed out, i q are 2.15, and 1.60, respectively. Additionally the predicted axial F"q distributions are well below the axially de. pendent Technical Specification I I

I 22 XN-NF-82-57 I N limits on F. The 80C F value of 1.711 compares with the measured Cycle q g 12 NTP value in Table 6.1 of 1.703. The control of the ccre power distribution is accomplished by following the procedures as discussed in the report, XN-76-40, " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors", Sep-tember 1976 and its addendum. The results reported in these documents demonstrate that the Power Distribution Control (PDC) procedures defined in the report will protect an axially dependent F limit with a peak g value of 2.30. The Technical Specification limit for R. E. Ginna has a peak of 2.32 and an axially dependence identical to that supported by the procedures. The physics characteristics of the Ginna Cycle 12 core at LTP are similar to those utilized in the PDC supporting analysis. The Ginna Technical Specification limits on F can therefore be protected by g operation under the PDC procedures as stated in XN-76-40. 6.1.2 Control Rod Reactivity Requirements Detailed calculations of shutdown margins for Cycle 12 at LTP are compared with Cycle 11 and Cycle 12 at NTP in Table 6.2. The ENC Plant Transient Simulation (PTS) Analysis indicates that the minimum required shutdown margin is 1600(4) pcm based upon the steamline break accident analyzed for ENC fuel at the E0C conditions. A value of 1,900 pcm is used at E0C in the evaluation of the shutdown margin to be consistent with the Technical Specifications. The Cycle 12 analysis at LTP indicates excess shutdown margins of 1,631 pcm at the B0C and 365 pcm at the E0C. The Cycle 12 NTP analysis (I ) indicates excess shutdown margins for that I

I l 23 XN-NF-82-57 I cycle of 1,497 pcm at the B0C and 288 pcm at the E0C. The slightly higher Cycle 12 LTP excess shutdown margins, when compared to the Cycle 12 NTP values, are due to slightly lower changes in the power defect due to the difference in the conditions in the moderator. The control rod groups and insertion limits for Cycle 12 will remain unchanged from Cycle 11. With these limits the nominal worth of the control bank, D-bank, inserted to the insertion limits at HFP is 211 pcm at B0C and 293 pcm at E0C. The control rod shutdown requirements in Table 6.2 allow for a HFP D-bank insertion equivalent to 300 pcm for both B0C and EOC. 6.1.3 Moderator Temperature Coefficient Considerations The reference Cycle 12 design calculations indicate that the moderator temperature coefficient is negative at all times during Cycle 12 at LTP as shown in Table 6.1. This meets the Technical Specifi-cation requirement that the moderator temperature coefficient be negative at all times during power operation and the design criteria that the power coefficient be negative. The least negative moderator temperature coefficient occurs at BOC HZP and is -0.4 pcm/ F. This compares with the 80C11 HZP value of -0.7 pcm/ F. 6.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 12 core analyses at LTP are described in References 17, 18, and 19. These methods have been verified for both 00 and Pu0 -UO 1 ttices. In summary, the reference neutronic design 2 2 2 analysis of the reload core was performed using the XTG (Reference 20) I

24 XN-NF-82-57 ,I reactor simulator system. The input exposure data were based on quarter 1 core depletion calculations performed from Cycle 5 to Cycle 11 using the XTG code. The 80C5 exposure distribution was obtained from plant data. The fuel shuffling between cycles was accounted for in the calculations. Predicted values of F, Fx, and F, were studied with the XTG q reactor model. The calculational thermal-hydraulic feedback and axial exposure distribution effects on pnwer shapes, rod worths, and cycle lifetime are explicitly included in the analysis. I I I I I I I I I

M M M M M M M M M M M M M M M M M Table 6.1 R. E. Ginna Neutronics Characteristics of Cycle 12 Compared with Cycle 11 Data Cycle 11 Cycle 12(5) NTP Cycle 12 LTP BOC E0C BOC EOC B0C E0C Critical Baron HFP, AR0, Eq. Xenon (ppm) 890(1) 20(1) 855(5) 0 873 31 HZP, ARD, No Xenon (ppm) 1,359(2) 1,330(2) 1,330(2) Moderator Temperature Coefficieint HFP, (pcm/0F) -8.2(4) -30.9(4) -7.1 -27.3 -5.65 -24.81 HZP,(pcm/0F) -0.7(2) -24.5(4) -0.3 -20.0 .44 -20.41 Doppler Coefficient, (pcm/0F) -1.39(4) -1.62(4) -1.35 -1.63 -1.40 -1.55 Boron Worth, (pcm/ ppm) HFP -8.2(4) -8.8(4) -7.8 -8.6 -7.9 -8.4 Total Nuclear Peaking Factor N F, HFP 1.750(3) 1.703(2) 1.51 1.711 1.524 0? Delayed Neutron Fraction .0059(6) .0052(4) .0059(5) .0052 .0059 .0052 Control Rod Worth of All Rods in Minus Most Reactive Rod, HZP,(pcm) 5775(4) 5967(4) 5599 5811 5665 5769 Excess Shutdown Margin (pcm) 1720(4) 463(4) 1497 288 1631 365 Moderator Pressure Coefficient (pcm/ psia) .09 0.35 .09 0.35 .09 0.35 x* (1) Extrapolated From Measured Data a (2) Measured Data 7 (3) 100% Power Map 150 MWD /MT (4) Reference 10 (5) Reference 21 (6) Startup Value

W W W M M M M M M M M M M M M M M M M Table 6.2 R. E. Ginna Control Rod Shutdown Margins and Requirements for Cycle 12 Cycle 11** Cycle 12+ NTP Cycle 12 LTP BOC EOC BOC E0C BOC EOC Control Rod Worth (HZP), pcm All Rods Inserted (ARI) 6,675 6,922 6,387 6,705 6,476 6,667 ARI Less Most Reactivity (N-1) 5,775 5,967 5,599 5,811 5,665 5,769 N-1 Less 10% Allowance ((N-1)*.9) 5,197 5,370 5,039 5,230 5,098 5,192 Reactivity Insertion, pcm Moderator plus Doppler 1,527 2,057 1,592 2,092 1,517 1,977 Flux Redistribution 600 600 600 600 600 600 Void 50 50 50 50 50 50 g Sum of The Above Three 2,177 2,707 2,242 2,742 2,167 2,627 Rod Insertion Allowance 300 300 300 300 300 300 Total Requirements 2,477 3,007 2,542 3,042 2,467 2,927 Shutdown Margin (N-1)*.9 - Total Requirements 2,720 2,363 2,497 2,188 2,631 2,265 Required Shutdown Margin

• 1,000 1,900 1,000 1,900 1,000 1,900 Excess Shutdown Margin 1,720 463 1,497 288 1,631 365 g

En Technical Specification 3.10 7 Calculated Values From Reference 10 Calculated Values From Reference 12 +

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p_ i . _.,_. i,..,.f.. . _.g J... 2.. 2. l i 200 a-- M ---+ - -- _ [. ' - d 4...L !.i '! _. _ i.. j. p .i _._ ' - ).. i i.. I i k I....L I ._ j. l i i e i i X l i z i. . ~.. t.... g l i 2 0 } i I n 0 1 2 3 4 5 6 7 8 9 N iw Exposure (GWD/MT) ~ Figure 6.1 R. E. Ginna Cycle 12, Critical Boron vs. Exposure, LTP, AR0

lI i 28 XN-NF-82-57 G F E D C B A I l .932 1.177 .873 1.110* .990 1.283 .929 7 I 1.194 1.069 .892 .973 1.289 1.013 .708 8 I .874 .893 1.246 1.200 1.138 .996 9 g 1.109* .974 1.201 1.104 1.107 .414 10 3 I .988 1.286 1.138 1.107 .797 Assembly Power 11 I 1.248 1.008 .994 .414 12 i Maximum Peaking ( .926 .705 t F = 1.711 F = 1.436 AH

• Mixed 0xide Assembly F

= 1.195 Z Figure 6.2 R. E. Ginna XTG Assembly Relative Power, Cycle 12, LTP, O MWD /MT, ARO, HFP (E0C12 = 7,100 MWD /MT) I I I

I 29 XN-NF-82-57 G F E D C B A I 1.003 1.212 .951 1.116* .995 1.203 .909 7 i l 1.225 1.121 .962 1.004 1.226 .996 .723 8 l .952 .962 1.227 1.173 1.106 .972 I 1.116* ! 1.004 1.174 1.084 1.077 .456 10 I .993 1.225 1.106 1.078 .801 Assembly Power 11 I 1.179 .993 .972 .456 12 t i F = 1.524 i .908 .721 1 q 13 F = 1.370 I AH

• Mixed 0xide Assembly F

= 1.104 Z I Figure 6.3 R. E. Ginna Cycle 12 XTG Depletion, I LTP, HFP, 8,700 MWD /MT, 41 ppm I I I

1 I 30 XN-NF-82-57 I 7.0 THERMAL HYDRAULIC DESIGN AND PERFORMANCE I The basic thermal hydraulic design analyses for ENC reload fuel at R. E. Ginna are reported in References 2, 3, and 4. These analyses are applicable to Cycle 12 in which ENC reload batch XN-5 will be placed in the core. Rod bow analyses for ENC fuel at R. E. Ginna were reported in the Cycle 11 SAR, XN-NF-81-01(10) The rod bow analyses considered peak I assembly exposures of 41,000 MWD /MT. Current projections are that Batch XN-5 fuel may reach 42,000 MWD /MT. This is a small increase in burnup and is at an exposure when the fuel will be non-limiting due to fissile depletion. The increase in assembly exposure, 42,000 MWD /MT versus 41,000 MWD /MT, is sufficiently small that the MDNBR including rod bow penalty of 1.48 reported previously(10) for the limiting two (2) pump coastdown transient continues to apply. The total nuclear peaking augmentation factor including rod bow is calculated to be 1.087 for an assembly exposure of 42,000 MWD /MT. The Ginna Technical Specifications allow a total nuclear peaking augmentation factor of 1.082 for calculation for the ECCS safety limits. This factor is adequate to accommodate nuclear augmentation due to rod bow in a limiting assembly with exposure up to 35,800 MWD /MTM as calculated, using the methodology of Reference 22. Fuel assemblies with exposures in excess of this value are anticipated to be operating well below the LOCA limits due to the reduction of assembly reactivity. Therefore, no additional penalty due to rod bow needs to be applied for calculation of LOCA limits. I I

31 XN-NF-82-57 I 8.0 ACCIDENT AND TRANSIENT ANALYSIS 8.1 ECCS ANALYSIS ENC has reanalyzed the segment of the limiting large break (0.4 DECLG) calculation which is affected by the fan cooler capacity change for nominal primary coolant system temperature and pressure conditions (573.5 F T and 2,250 psia) and for reduced temperature and pressure avg conditions (527.5 F T and 2,000 psia). These calculated results are avg documented in XN-NF-82-26(0) The results show that the criteria specified by 10 CFR 50.46 are satisfied with this analysis which was performed in conformance with Appendix K of 10 CFR 50. This analysis supports operation of the R. E. Ginna plant with a peak Linear Heat Generation Rate (LHGR) I of 13.76 kw/ft at a total peaking factor (F ) of 2.32 and at a rated power of 1,520 MWt with ENC fuel. These results are applicable over the range of primary coolant system fluid conditions af 2,000 to 2,250 psia pressure and 527.5 F to 573.5 F T avg 8.2 PLANT TRANSIENT ANALYSIS I The kinetics parameters for Cycle 12, which has ENC P.eload Batch XN-5, are presented in Table 8.1. The kinetics parameters used in the plant transient analysis for reduced T, g and pressure reported in Reference 4 are also presented in Table 8.1. The analyses showed that operation at LTP is less limiting for plant transient considerations than operation at normal conditions. I I I

I 32 XN-NF-82-57 I 8.3 R0D EJECTION ANALYSIS FOR R. E. GINNA CYCLE 12, A Control Rod Ejection Accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a Rod Cluster Control Assembly (RCCA) and drive shaft. The consequence of this inechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel damage. The rod ejection accident analysis presented in the document XN-NF-77-53I7) is still applicable to Cycle 12 operations. The Cycle 12 fuel assembly loading configuration introduces minimal effects on ejected rod worths and hot pellet peaking factors. The ejected rod worths and hot pellet peaking factors are calculated using the XTG code. No credit was taken for the power flattening effects of Doppler or moderator feed-back in the calculation of ejected rod worths or peaking factors. The calculations made for Cycle 12 using XTG were two-dimensional (x-y) with appropriate axial buckling correction terms. The total peaking factors N (F ) were determined as the product of the radial peaking factor (as calculated using XTG) and a conservative axial peaking factor. The pellet energy deposition resulting from an ejected rod was evaluated using the " Generic Analysis of the Control Rod Ejection Transition for PWR."(23) The rod ejection accident was found to result in energy deposition of less than 280 cal / gam stated in Regulatory Guide 1.77. The results of the control rod ejection transient for this case are presented in Table 8.2 along with the results from References 7, 10, l 11, and 12. I

M M M M M M M M M M M M M M M M M M M Table 8.1 R. E. Ginna Unit 1 Kinetics Parameters Cycle 12 Values Used in Analysis LTP Nominal Rod Withdrawal Loss of Steamline Break Values And Loss of Load Primary Flow Transients From Parameters BOC E0C Transients, 80C Transients, BOC HZP, E0C Moderator Temperature Coefficient, (pcm/0F) -5.7 -24.8 0 0 Moderator Pressure Coefficient (pcm/ psia) +.09 +.35 0 0 Doppler Coefficient (pcm/0F) -1.4 -1.6 -1.0 -1.5 Boron Worth Coefficient w (pcm/ ppm) -7.9 -8.4 -7 Scram Worth (pcm) -3467 -4827 -1600 -1600 Shutdown Margin (pcm) 1000 1900 -1600 Delayed Neutron Fraction .0059 .0052 .0059 .0059 .0049 Not Applicable To This Transient See Figure 3.32 of Reference 4 See Figure 3.33 of Reference 4 5 M k

W W M M M M M M M M M M M M M M M M 1 Table 8.2 Ejected Rod Worth and Peaking Factors Cycle 8(2) Cycle 11(3)_ Cycle 12(3) NTP Cycle 12(3) LTP HFP HZP HFP HZP HFP HZP HFP HZP F"qBefore P'., ' ion 2.25 2.82 2.21(1) 3.12(1) 2.28(1) 2.71(1) 2.30(1) 2.76(1) N -v 4.36 5.30 2.88(1) 5.40(1) 3.00(1) 5.61(1) 3.11(1) 5.79(l) F After Ejc: Maximum Roc. + From a Fuli inse.r'.s.s Bank ( m ) 0.470 0.640 0.302 0.467 0.313 0.502 0.345 0.510 Energy Deposition (cal /gm) 171 37 16/(4) 29(4) 167(4) 28(4) 172(4) 30(4) (1) Includes a conservative estimate of F Of I* Z^ and at HZP of 1.8. (2) Reference 7, calculated with XTRAN. (3) Calculated with XTGPWR. (4) Reference 23, determined from the ENC generic analysis of the control rod ejection transient. 5, e

35 XN-NF-82-57 I

9.0 REFERENCES

I 1. XN-NF-77-52, "R. E. Ginna Reload Fuel Design", November 1977. I 2. XN-NF-77-40, " Plant Transient Analysis for R. E. Ginna, Unit 1 Nuclear Power Plant", Revision 1, July 1979. I MacDuff, R. B., " Plant Transient Analysis for The R. E. Ginna Unit 3. 1 Nuclear Power Plant", XN-NF-77-40, Supplement 1, March 1980. 4. XN-NF-82-45, " Plant Transient Analysis for Operation of the R. E. I Ginna Unit 1 Nuclear Power Plant at Reduced Pressure and Temperature", Revision 1, July 1982. 5. XN-NF-77-58, "ECCS Analysis for the R. E. Ginna Reactor with ENC WREM-II PWR Evaluation Model", December 1977. I XN-NF-82-26, "R. E. Ginna Revised LOCA ECCS Analyses for Nominal 6. and Reduced Temperature and Pressure Operation". 7. XN-NF-77-53, "R. E. Ginna Nuclear Plant Cycle 8 Safety Analysis I Report", December 1977. 8. XN-NF-78-50, "R. E. Ginna Cycle 9 Safety Analysis Report", December 1978. 9. XN-NF-79-103, "R. E. Ginna Cycle 10 Safety Analysis Report With Mixed 0xide Assemblies", December 1979.

10. XN-NF-81-01, "R. E. Ginna Nuclear Plant Cycle 11 Safety Analysis Report", Exxon Nuclear Company, February 1981.

11. XN-NF-81-94, "R. E. Ginna Nuclear Plant Cycle 12 S.afety Analysis Report", November 1981.

I

12. Letter, PWR:009:82, "R. E. Ginna: Impact of Reduced Cycle 11 Operation on the Cycle 12 Safety Analyses", March 23, 1982.

l 13. XN-NF-80-6, "R. E. Ginna Design Report, Zircaloy Guide Tube, Revision l 1", Exxon Nuclear Company, February 1981.

14. XN-NF-77-35, "R. E. Ginna Reload Fuel Mechanical and Thermal Hydraulic Design Report", September 1977.

l 15. XN-NF-81-10, "R. E. Ginna Design Report and Extended Burnup Analysis", l February 1981. 1 LI 1 !I

36 XN-NF-82-57 I XN-NF-600, "R. E. Ginna Design Report, XN-5 and XN-6 Design", 16. November 1981. 17. F. B. Skogen, " Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April 1977. 18. XN-75-27(A), Supplement 1 to Reference 17, April 1977. 19. XN-75-27(A), Supplement 2 to Reference 17, December 1977. I 20. XN-CC-28, Revision 3, "XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January 1975. 21. XN-NF-82-41 (P), "R. E. Ginna Cycle 12 Startup and Operations keport", May 1982. 22. T. L. Krysinski, J. L. Jaech and

t. A. Nielsen, " Computational Procedure for Evaluating Fuel Rod Bowing", XN-NF-75-32(NP),

Supplement 1, July 23, 1979. 23. XN-NF-78-44, "A Generic Analysis of the Control Rod Ejection l Transient for PWRs", January 1979. I I I I I

I XN-NF-82-57 l ISSUE DATE: 07/20/82 R. E. GINNA NUCLEAR PLANT CYCLE 12 SAFETY ANALYSIS REPORT FOR LOW TEMPERATURE AND PRESSURE DISTRIBUTION I GJ BUSSELMAN RA COPELAND / HE WILLIAMSON WV KAYSER / JE KRAJICEK JN MORGAN / GC COOKE LA NIELSEN / RL LUECK (2) WT NUTT . l GF OWSLEY lg FB SK0 GEN i RB STOUT GA SOFER RG&E / TJ HELBLING (80) DOCUMENT CONTROL (5) I I I .}}