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{{#Wiki_filter:0 KN-N-79-103 P E (lllklk PlljlllLEkk PILklll'NllllLE JIB MFET7 kkkL7SIIS PEPBP'll'IIYM Iw3IIXEB QXIIBE bhSSEliNBILIIES DECEMBER 1979 RICHLAND, NA 99352 E I'I I-I I XN-NF-79-103 0: IR/R14 79 R.E.GINNA NUCLEAR PLANT CYCLE 10.SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES Prepared: G.J.Buss man, Manager Neutronics and Fuel Management Approved: G.A.Sofe nager Nuclear Fue s Engineering Concurred:
{{#Wiki_filter:0 KN-N 103 P E (lllklk PlljlllLEkk PILklll'NllllLEJIB MFET7 kkkL7SIIS       PEPBP'll'IIYM Iw3IIXEB QXIIBE bhSSEliNBILIIES DECEMBER 1979 RICHLAND, NA           99352
J.N.Morgan, Manage Licensing and Safety Engineering Concurred:
 
L.J.Federico, Manager Nuclear Fuels Project l'~/7 E)j(ON NUCLEAR COMPANY, Inc.
E I
NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Ir PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc.It is being Sub.mitted by Exxon Nuclear to the USNRC as part of a technical contri.bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear.fabricated reioarl fuel or other teclmical services provided by Exxon Nuclear for lieht water power reactors an<I it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief.The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNQC whicII are customers ot Exxon Nuclear in their demonstration of compliance wIth the USNRC's regulations.
  '
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf: A.Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained,in this document, or that the use ot any information, apparatus, method, or process disclosed in this document will not Itffringe privately owned rights;or 8.Assumes any liabilities with respect to the use of, or for dan'ages resulting from the use of, any information, ap.paratus, method, or process disclosed in this document.XN-NF-FQO, 766 XN-NF-79-103 TABLE OF CONTENTS'ection
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0: XN-NF-79-103 IR/R14 79 R. E. GINNA NUCLEAR PLANT CYCLE 10.
SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES Prepared:
G. J. Buss   man, Manager Neutronics and Fuel Management Approved:
G. A . Sofe     nager Nuclear Fue   s Engineering Concurred:
J. N. Morgan, Manage Licensing and Safety Engineering Concurred:                               l'~/    7 L. J. Federico, Manager Nuclear Fuels Project E)j(ON NUCLEAR COMPANY, Inc.
 
NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Ir PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being Sub.
mitted by Exxon Nuclear to the USNRC as part of a technical contri.
bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear. fabricated reioarl fuel or other teclmical services provided by Exxon Nuclear for lieht water power reactors an<I it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNQC whicII are customers ot Exxon Nuclear in their demonstration of compliance wIth the USNRC's regulations.
Without derogating from the foregoing,   neither Exxon Nuclear nor any person acting on its behalf:
A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained,in this document, or that the use ot any information, apparatus, method, or process disclosed in this document will not Itffringe privately owned rights; or
: 8. Assumes any liabilities with respect to the use of, or for dan'ages resulting from the use of, any information, ap.
paratus, method, or process disclosed in this document.
XN- NF- FQO, 766
 
XN-NF-79-103 TABLE OF CONTENTS'ection
                                                                        ~Pa  e


==1.0 INTRODUCTION==
==1.0 INTRODUCTION==
AND  
AND  


==SUMMARY==
==SUMMARY==
.2.0 OPERATING HISTORY OF THE REFERENCE CYCLE.3.0 GENERAL DESCRIPTION 4.0 FUEL SYSTEM DESIGN.~Pa e 2 5 9 5.0 NUCLEAR DESIGN.................'........
.
10 5.1 PHYSICS CHARACTERISTICS 5.1.1 POWER DISTRIBUTION CONSIDERATIONS.
2.0 OPERATING HISTORY OF THE REFERENCE CYCLE .                               2 3.0 GENERAL DESCRIPTION                                                     5 4.0 FUEL SYSTEM DESIGN .                                                     9 5.0 NUCLEAR DESIGN.................'........                               10 5.1 PHYSICS CHARACTERISTICS 5.1.1     POWER DISTRIBUTION CONSIDERATIONS.
5.1.2 CONTROL ROD REACTIVITY REQUIREMENTS.
5.1.2     CONTROL ROD REACTIVITY REQUIREMENTS.                     12 5.1.3     MODERATOR TEMPERATURE COEFFICIENT CONSIDERATIONS .. 13 5.2 ANALYTICAL METHODOLOGY.                                           13 6.0 THERMAL HYDRAULIC   DESIGN....................                         20 7.0 ACCIDENT AND TRANSIENT ANALYSIS.                             ~ ~        21 7.1 PLANT TRANSIENT AND ECCS ANALYSES FOR R. E. GINNA .               21 7.2 ROD EJECTION ANALYSIS FOR R. E. GINNA CYCLE 10.                   22
12 5.1.3 MODERATOR TEMPERATURE COEFFICIENT CONSIDERATIONS
..13 5.2 ANALYTICAL METHODOLOGY.
13 6.0 THERMAL HYDRAULIC DESIGN....................
20 7.0 ACCIDENT AND TRANSIENT ANALYSIS.7.1 PLANT TRANSIENT AND ECCS ANALYSES FOR R.E.GINNA.7.2 ROD EJECTION ANALYSIS FOR R.E.GINNA CYCLE 10.


==8.0 REFERENCES==
==8.0 REFERENCES==
  .                                                          25
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XN-NF-79-103 LIST OF TABLES Tabl e                                                                       ~Pa   e 3.1   R. E. GINNA CYCLE 10 FUEL ASSEMBLY DESIGN PARAMETERS     -.....           6 5.1 WITH CYCLE 9 DATA....................
~~~.l~~~t~~t XN-NF-79-103 LIST OF TABLES Tabl e~Pa e 3.1 R.E.GINNA CYCLE 10 FUEL ASSEMBLY DESIGN PARAMETERS
R. E. GINNA NEUTRONICS CHARACTERISTICS OF CYCLE 10 COMPARED 15 5.2   R. E. GINNA CONTROL ROD SHUTDOWN MARGINS AND REQUIREMENTS FOR CYCLE 1 0 ~ ~ ~ ~ ~ ~ ~   ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~       16 II 7.1   R. E. GINNA KINETIC PARAMETERS .                                   ~ ~   23 7.2   EJECTED ROD WORTH AND PEAKING FACTORS.     . . . . . . . . . . . . . 24
-.....6 5.1 R.E.GINNA NEUTRONICS CHARACTERISTICS OF CYCLE 10 COMPARED WITH CYCLE 9 DATA....................
15 5.2 R.E.GINNA CONTROL ROD SHUTDOWN MARGINS AND REQUIREMENTS FOR CYCLE 1 0~~~~~~~~~~~~~~~~~~~~~~~~II 7.1 R.E.GINNA KINETIC PARAMETERS
.16~~23 7.2 EJECTED ROD WORTH AND PEAKING FACTORS..............24  
\~~~~~~~~~l~~i I~~~~~
XN-NF-79-103 LIST OF FIGURES~Fi ure 2.1 R.E.GINNA CYCLE 9 CRITICAL BORON CURVE, PREDICTED VS.MEASURED e~~~~~~~~~~~~~~~~~~~~~~~~2.2 R.E.GINNA POWER DISTRIBUTION COMPARISON TO MAP IX-24 5)505 MWD/MT e~~~~~~~~~~~~~~~~~~~~~~3.1 R.E.GINNA CYCLE 10 LOADING PATTERN.3.2 R.E.GINNA BOC10 QUARTER CORE EXPOSURE DISTRIBUTION AND REGION ID@~~~~~~~~~~~~~~~~~~~~~~~~5.1 R.E.GINNA CYCLE 10 ARO CRITICAL BORON CONCENTRATION VS.EXPOSURE~~~~~~~~~~~~~~~~~~~~~~~~~5.2 R.E.GINNA CYCLE 10 POWER DISTRIBUTION HFP, 0 MWD/MT 1)254 PPM 5.3 R.E.GINNA CYCLE 10 POWER DISTRIBUTION HFP, 9,500 MWD/MT PPM~~~~~~~~~~~~~~~~~~~~~~~~~~7~Pa e~~3 4 7 8 17 18 18 t 1~~~~~~I~l i~~~~l XN-NF-79-103 R.E.GINNA NUCLEAR PLANT CYCLE 10 SAFETY ANALYSIS REPORT k gl


==1.0 INTRODUCTION==
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XN-NF-79-103 LIST OF FIGURES
~Fi ure                                                                      ~Pa  e 2.1  R. E. GINNA CYCLE 9 CRITICAL BORON CURVE, PREDICTED VS.
MEASURED  e ~  ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~ ~ ~  ~ ~    3 2.2  R. E. GINNA  POWER  DISTRIBUTION COMPARISON TO  MAP  IX-24 5)505 MWD/MT  e  ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~ ~ ~          4 3.1  R. E. GINNA CYCLE 10 LOADING PATTERN .                                      7 3.2  R. E. GINNA BOC10 QUARTER CORE EXPOSURE DISTRIBUTION AND REGION ID@  ~  ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~ ~ ~          8 5.1   R. E. GINNA CYCLE 10 ARO CRITICAL BORON CONCENTRATION VS.
EXPOSURE  ~ ~  ~  ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~ ~ ~          17 5.2  R. E. GINNA CYCLE 10 POWER DISTRIBUTION HFP, 0 MWD/MT 1)254 PPM                                                                  18 5.3  R. E. GINNA CYCLE 10 POWER DISTRIBUTION HFP,      9,500  MWD/MT 7 PPM  ~ ~ ~  ~  ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~ ~  ~ ~ ~ ~  ~ ~ ~ ~ ~ ~          18
 
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XN-NF-79-103 R. E. GINNA NUCLEAR PLANT CYCLE 10 SAFETY ANALYSIS REPORT


==1.0  INTRODUCTION==
AND  
AND  


==SUMMARY==
==SUMMARY==
The R.E.Ginna Nuclear plant will operate in Cycle 10 beginning in early 1980 with three regions of fuel supplied by Exxon Nuclear Company (ENC).The loading will consist of 32 ENC assemblies in Region 12 and 4 Westinghouse mix oxide (MOX)assemblies.
 
The remainder of the core contains 40 once-burnt and 32 twice-burnt ENC assemblies and 13 exposed Westinghouse supplied assemblies.
The R. E. Ginna       Nuclear plant   will operate in     Cycle 10 beginning in early   1980 with three regions of fuel supplied         by Exxon Nuclear Company (ENC). The loading   will consist of     32 ENC assemblies     in Region   12 and 4 Westinghouse mix oxide (MOX) assemblies.             The remainder   of the core contains 40 once-burnt and     32   twice-burnt ENC   assemblies   and 13 exposed   Westinghouse supplied assemblies.
The characteristics of the fuel and of the reloaded core result in conformance with existing Technical Specification limits regarding shutdown margin provisions and thermal limits.This document provides the neutronic analysis for the plant during Cycle 10 operation and the control rod ejection analysis.The ENC fuel design.is unchanged from the fuel design used in the"Cycle 8 and 9 ENC fuel reloads.The previous Plant Transient Analysis'(2)remains valid for Cycle 10.The ECCS analysis is applicable to Cycle 10 operation.
The characteristics of the fuel         and of the reloaded core result in conformance   with existing Technical Specification limits regarding shutdown margin provisions and thermal         limits. This     document provides the neutronic analysis for the plant during Cycle           10 operation and the control rod ejection k
The consequences of the rod ejection accident for Cycle 10 are slightly less severe than those calculated for Cycles 8 and 9.The (4)(5)introduction of the 4 MOX assemblies into the reactor core leads to small changes in the core average kinetic parameters resulting in minimal effects to the previous analyses performed for Cycles 8'and 9 (1,2,4)(1,3,5)  
analysis. The ENC fuel design   . is unchanged   from the fuel design used in the"Cycle   8 and 9 ENC     fuel reloads. The previous Plant Transient Analysis
~~~~~l~~~}i~~~~~~~
                                                                                                  '(2) remains   valid for Cycle     10. The ECCS     analysis       is applicable to Cycle       10 gl operation. The consequences     of the rod ejection accident for Cycle           10 are for Cycles 8 (4) and    (5) .
XN-NF-79-103 2.0 OPERATING HISTORY OF THE REFERENCE CYCLE R.E.Ginna Cycle 9 has been chosen as the reference cycle with respect to Cycle 10 due to the close resemblance of the neutronic characteristics between these two cycles.The Cycle 9 operation began on April 3, 1979, and as of November 31, 1979 the core had accrued about 6,714 MWD/MT.The Cycle 9 loading included 40 fresh ENC fuel assemblies with 32 exposed ENC assemblies and 49 exposed Westinghouse assemblies.
slightly less    severe than those calculated                                9          The introduction of the       4 MOX assemblies   into the reactor core leads to small changes   in the core average kinetic parameters resulting in minimal effects to the previous analyses performed for Cycles 8 (1,2,4)       '      and 9 (1,3,5)
The measured power peaking factors at hot-full-power, equilibrium xenon conditions, have remained considerably below the Technical Specification ,limits throughout Cycle 9.The total nuclear peaking factors, F , and the radial nuclear pin peaking factor, F H, have remained below 1.75 and 1.45, respectively.
 
Cycle 9 operation has typically been rod'free with the D control bank positioned in the range of 218 to 222 steps, 225 steps being fully withdrawn.
~
It is anticipated that similbr control bank insertions will be seen in Cycle 10.The critical boron concentration as calculated by ENC for Cycle 9 has agreed to within about 8 ppm compared to the observed values (see Figure 2.1).Also the predicted power distributions have typically agreed to within+3 percent of the, measured values (see Figure 2.2 for typical com-parison).  
~
-~~--~~-~~=-:LL~=*~-:..t=~I-~~-~t 44 C-l:-f~I~~t*'I t j=~~~++4 I~~=~~~~-~~~~~-~~-~*t-Ut'e.~----I.fl GMO'NT.}'~-:.-::F.-T''re
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XN-NF-79-103 2.0   OPERATING HISTORY OF THE REFERENCE CYCLE R. E. Ginna Cycle 9 has been chosen     as the reference cycle with respect to Cycle   10 due to the close resemblance of the neutronic characteristics between these two   cycles. The Cycle 9   operation began   on April 3, 1979,   and as of November 31, 1979 the core had accrued about 6,714 MWD/MT.             The Cycle 9 loading included 40 fresh   ENC fuel assemblies with     32 exposed   ENC assemblies and 49 exposed   Westinghouse assemblies.
The measured   power peaking factors at hot-full-power, equilibrium         xenon conditions, have remained considerably below the Technical Specification
,limits throughout Cycle 9. The total nuclear peaking factors,       F , and the radial nuclear pin peaking factor,     F H, have remained below 1.75 and 1.45, respectively. Cycle 9 operation has   typically   been rod'free with the     D control bank positioned in the range of       218 to 222 steps,   225 steps   being fully withdrawn. It is anticipated that similbr control           bank insertions will be seen   in Cycle 10.
The critical boron concentration as calculated by       ENC for Cycle 9 has agreed to   within about 8 ppm compared   to the observed values (see Figure 2.1). Also the predicted power distributions have       typically   agreed to within +3 percent of the, measured values (see Figure 2.2 for typical com-parison).
 
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XN-NF-79-103
XN-NF-79-103
.968.995-2.71 1.110 1.129-1.68.920.938-1.92.949..965-1.66 1.189 1.178'93.996.963 3.43.794.782 1.53 1.108 1.131-2.03.986 1.013-2.67 1.029 1.048-1.81 1.099 1.103-.36 1.189 1.168 1.80 1.074 1.051 2.19.663.645 2.79.918.947-3.06 1.030 1.051-2.00.996 1.006-.99 1.203 1.191 1.01 1.079 1.055 2.27;977.974.31.953.968-1.55 1.098 1.105-.63 1.191 1.192-.08 1.036 1.021 1.47 1.178 1.149 2.52.714.711.42 10 1;195 1.179 l.36 1.188 1.169 1.63 1.072 1.055 1.61 1.173 1.149 2.09.804.798.75.985.963 2.28.772.782-1.28 1.071 1.051 1.90.651 ,.645.93.977.975.21.715.712.42 Measured Assembly Power Calculated (XTGPWR)x 100 c 12 13 Calculated Measured%Difference N Fq 1.528 1.564 2.33 F~H 1.351 F 1.105 1.337 1.154-1.07 4.42 Figure 2.2 R.E.Ginna Power Distribution Comparison, To Map IX-24, HFP, 5,505 MWD/MT I-l XN-NF-79-103 3.0 GENERAL DESCRIPTION The R.E.Ginna reactor consists of 121 assemblies, each having a 14xl4 fuel rod array.Each assembly contains 179 fuel rods, 16 RCC guide tubes, and 1 instrumentation tube.The fuel rods consist of slightly enriched U02 pellets inserted into zircaloy tubes.The RCC guide tubes and the instr umen-tation tube are made of SS-304L.Each ENC assembly contains nine zircaloy spacers with Inconel springs;eight of the spacers are located within the active fuel region.Four of the 121 assemblies contain Mixed Oxide (Pu02 plus U02)bearing fuel rods.The MOX assemblies consist of three enrichment zones of Pu02 utilizing natural U02 as the diluent.The projected Cycle 10 loading pattern is shown in Figure 3.1 with the assemblies identified by their Fabrication ID's and Region ID's.The initial enrichments of the various regions are listed in Table 3.1.BOC10 exposures, based on an EOC9 exposure of 9,570 MHD/MT, along with Region ID's are shown in Figure 3.2.The core consists of 32 fresh ENC assemblies at 3.45 w/o and 4 fresh Westinghouse MOX assemblies loaded on the periphery with 72 ENC and 13 Westinghouse exposed assemblies scatter-loaded in the center portion of the core.Pertinent fuel assembly parameters for the Cycle 10 fuel are depicted in Table 3.1.The transuranic elements, including Am-241, have been\accounted for up to the time of the anticipated reactor startup.
  .968     1.110    .920        .949  1.189          .996        .794
XN-NF-79-103 Table 3.1 R.E.Ginna Cycle 10 Fuel Assembly Design Parameters 10 Region 12 MOX Enrichment, wtX U-235 Number of Assemblies Pellet Densi'ty, X TD Pellet-to-Clad Diametrical Gap, Mil Fuel Stack Height, inch 7.5 7.5 141.4 142.0 3,103 3,100 13 32 95,0 94,0 40 94.0 32 4 94.0 95.0 7.5 7.5 7.5 142.0 142.0 141.4 3.200 3.450 2.626*Region Average Burnup at BOC10, MWD/MT Nominal Assembly Weight, KgU 24,339 17,885 8,335 0 0 392.56 373.78 373,78 373.78 395.91***wtX Pu (based on assembly average)*" in Kg HM XNrNF-79-103 K J I H.G F E 0 C B A 12 HOX 12 12 12 N09'12 L14 12 L09 M14 K03 L31 M39 12 12 L01 H01 12 12 M02 L06 N17 L19 M28 L26 N36 L05 N12 12 12 L02 M33 L21 M23 K05 N30 L24 M20 L12 12 12 MOX L32 M40 L27 12 K09 M25 M31 K13 N07 K20 N08 K28 K19 N06 N22 L18 H13 K18 M27 K27 L13 12 12 MOX 12 L15 M15 L20 M24 K26 M05 K17 H29 L25 M38 L30 12 12 Ll 0 M18 L22 N32 K25 M21 L23 M35 L04 12 12 Ml 0 L07 12 M03 N34 L28 N26 L03 M37 K14 L17 H16 M19 L08 M04 Lll Mll 12 12 12 12 L29 12 L16 12 12 12 MOX 12 Fabri cation or New Fuel Region Identification Figure 3.1 R.E.Ginna Cycle 10 Loading Pattern XN-NF-79-103 D C B 24,736 7,517 24,117 7,809 24,708 0 12 MOX 7;522 24,093 11,341 17,578 9,544 10 19,816 10 12 24,117 11,342 17,506 6,926 16,061 10 10 12 7,809 17,574 6,928 10 18,677 10 6,393 0 12]n 24,708 9,549 16,061 6,198 10 12 12 19,809 10 0 a 12 12 BOC10 Exposure MWD/MT Region ID*12 0 MOX 12 13*See Table 3.1 for Region definitions f Figure 3.2 R.E.Ginna BOC10 quarter Core Exposure Distribution and Region ID XN-NF-79-103 4.0 FUEL.SYSTEM DESIGN A description of the Exxon Nuclear supplied fuel design and design methods is contained in Reference 1.This fuel has been specifically designed to be compatible to the resident fuel supplied by Westinghouse.
  .995    1.129   .938      ..965    1.178          .963        .782
I 10 XN-NF-79-103 5.0 NUCLEAR DESIGN The neutronic charact ristics of the projected Cycle 10 core are quite similar to those of the Cycle 9 core (see Section 5.1).The nuclear design bases for the Cycle 10 core are as follows: 1)The design shall permit operation within the Technical Specifications for the R.E.Ginna plant.2)The length of Cycle 10 shall be determined on the basis of an assumed Cycle 9 length of 9,570 MWD/MT.3)The Cycle 10 loading pattern shall be optimized to achieve power distributions and control rod reactivity worths according to the following constraints:
-2.71      -1. 68 -1.92        -1.66          '93 3.43         1.53 1.108       .986 1.029        1. 099  1.189          1.074        .663 1.131      1.013  1. 048      1.103  1.168         1.051        .645
a)The peak F~shall not exceed 2.32 and the peak F H shall not exceed 1.66 (including uncertainties) in any single fuel rod.through the cycle under nominal full power operation condi-tions.b)The scram worth of all rods minus the most reactive shall exceed BOC and EOC shutdown requirements.
-2.03      -2.67  -1. 81        -.36    1. 80          2.19         2.79
4)The Cycle 10 core shall have a negative power coefficient.
  .918     1. 030   .996       1. 203 1.079          ;977
5)The MOX assemblies shall be located in a region of the reactor core as to minimize the effects on shutdown margin provisions and thermal limits.The neutronic design methods utilized to ensure the above requirements are consistent with those described in References 6, 7, and 8.
  .947    1. 051 1.006        1.191  1.055           .974
XN-NF-79-103 5.1 PHYSICS CHARACTERISTICS The neutronic characteristics of the Cycle 10 core are compared with those of Cycle 9 and are presented in Table 5.1.The data presented in the table indicate the'eutronic similarity between Cycles 9 and 10.The Cycle 10 loading pattern is applicable for Cycle 9 lengths of+700 MWD/MT and-800 MWD/MT about the nominal length of 9,570 MWD/MT.The calculated boron letdown curve for Cycle 10 is shown in Figure 5.1.The curve indicates a BOC10, no xenon, critical boron concentration of 1,254 ppm.At 150 MWD/MT, equilibrium xenon, the critical boron concentration is 921 ppm.The Cycle 10 length is projected to be 9,500+300 MWD/MT with 7 ppm of boron at EOC.5.1.1 Power Distribution Considerations P Representative predicted power maps for Cycle 10 are shown in Figures 5.2 and 5.3 for BOC and EOC conditions, respectively.
-3.06      -2.00  -.99        1. 01  2.27             .31
The power distributions were obtained from a three-dimensional model with moderator I, density and Doppler feedback effects incorporated.
  .953     1.098  1. 191      1. 036  1.178          .714
For the projected Cycle 10 loading pattern the calculated BOC nuclear power peaking factors, F~, N N N F, and Fz, are,l.745, 1.433, and 1.201, respectively.
  .968    1.105  1.192       1. 021  1.149          .711                10
At EOC conditions the corresponding values are 1.517, 1.358, and 1.098.The Technical Specifi-cation limits relative to F~and F>H, with the measurement uncertainties N N backed out, are 2.15 and 1.60.Additionally the predicted axial F distri-butions are well below the axially dependent Technical Specification limits on F~.The BOC F value of 1.745 compares with the measured Cycle 9 value N in Table 5.1 of 1.758.
-1.55      -.63  -.08        1.47    2.52             .42 1;195     1.188  1. 072      1.173    .804 1.179      1.169  1.055       1.149    .798
12 XN-NF 79-,103 The control of the core 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", September 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 value of 2.30.The Technical Specification limit for R.E.Ginna has a peak of 2.32 and an axial dependence identical to that supported by the procedures.
: l. 36      1.63  1.61        2.09     .75
The physics characteristics of the Ginna Cycle 10 core are similar to those utilized in the PDC supporting analysis.The Ginna Technical Specification limits on F can therefore be protected by operation under the PDC procedures as stated in XN-76-40.5.1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 10 are compared with Cycle 9 data in Table 5.2.The ENC Plant Transient Simulation (PTS)Analysis indicates that the minimum required shutdown margin is 1,800 pcm based upon the steamline break accident analyzed for ENC fuel at the EOC conditions.
.985     1. 071  .977        .715    Measured Assembly Power
A value of 1,900 pcm is used at EOC in the evaluation of the shutdown margin to be consistent with'the Technical Specifications.
.963    1.051    .975        .712    Calculated      (XTGPWR)          12 2.28      1.90     .21          .42 x 100 c
The Cycle 10 analysis indicates excess shutdown margins of 1,414 pcm at the BOC and 344 pcm at the EOC.The Cycle 9 analysis indicates excess shut-down margins for that cycle of 1,795 pcm at the BOC and 393 pcm at the EOC.The slightly lower Cycle 10 excess shutdown margins, when compared to the Cycle 9 values, are due to slightly lower calculated rod worths.
.772      .651
13 XN-NF-79 103 The control-rod groups and insertion limits for Cycle 10 will remain unchanged from Cycle 9.With these limits the'nominal worth of the control bank, D-bank, inserted to the insertion limits'at HFP is 122 pcm at, BOC and'70 pcm at EOC.The control rod shutdown requirements in Table 1 5.2 allow for a HFP D-bank insertion equivalent to 300 pcm for both BOC and EOC.5.1.3 Moderator Tem erature Coefficient Considerations The reference Cycle 10 design calculations indicate that the moderator temperature coefficient is negative at all times during the cycle as shown in Table 5.1.This meets the Technical Specification requirement that the moderator temperature coefficient be negative at all times during power operation and the design criteria that the power coefficient be nega-tive.The least negative moderator temperature coefficient occurs at BOC HZP and is-2.0+2pcm/
.782    ,  .645
F.This compares with the BOC9 HZP value of-2.0 pcm/F.5.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 10 core analyses are described in References 6, 7, and 8.These methods have been verified for both U02 and Pu02-U02 lattices.In summary, the reference neutronic design analysis of the reload core was performed using the XTG (Reference 9)reactor simulator system-.The input exposure data were based on quarter core depletion calcu-lations performed from Cycle 5 to Cycle 9 using the XTG code.The BOC5 exposure distribution was obtained from plant data.The fuel shuffling between c'ycles was accounted for in the calculations.
-1.28        .93                                                          13 Calculated   Measured         %Difference N
14 XN-NF-79-103 Predicted values of F~, Fx , and F were studied, with the XTG 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.
Fq       1. 528       1. 564           2.33 F~H     1.351         1. 337          -1. 07 F        1.105        1.154            4. 42 Figure 2.2 R. E. Ginna Power Distribution Comparison, To Map IX-24, HFP, 5,505 MWD/MT
15 XN-NF-79-103 Table 5.1 R.E.Ginna Neutronics Characteristics of Cycle 10 Compared with Cycle 9 Data BOC C cle 9 EOC C cle 10 BOC EOC (Critical Boron HFP, ARO, Equilibrium Xenon (ppm)HZP, ARO No Xenon (ppm)Moderator Temperature Coefficient HFP, (pcm/oF)HZP, (pcm/oF)Doppler Coefficient, (pcm/F)Boron Worth, (pcm/ppm)HFP HZP Total Nuclear Peaking Factor Fq, HFP 961')12(1,41 0(2)-8.12-8.58-8.72 1.758(-7.6(2)-30.4-2.0-21.5-1.25 to-2.0 921 1,414-8.1-2.0-30.4-21.6-7.95 ,-8.62 1.745 1.517-,1.35-1.84 Delayed Neutron Fraction.0061.0051.0058.0052 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm)Excess Shutdown Margin (pcm)Moderator Pressure Coefficient (pcm/psi)5,751)5,821.1,795()393(0.35 5';341 5,696 1,414 344 0.35 (1)Extrapolated from measured data (2)Measured Data (3)70/Power Map (4)Reference 5 16 XN-NF-79-103 Table 5.2, R.E.Ginna Control Rod Shutdown Margins and Requirements for Cycle 10 BOC C cle 9**EOC C cle 10 BOC EOC Control Rod Worth HZP , cm All Rods Inserted (ARI)ARI less most reactive (N-1)N-1 less lOX allowance L(N-1)*9l Reactivit Insertion cm 5,176 5,239 6,407 6,634 5,751 5,821 5,949 6,420 5,341 5,696 4,807 5,125 Moderator plus Doppler Flux Redistribution Void Sum of the above three Rod Insertion Allowance Total Requirements 1,431 600 50 2,081 300 2,381 1,996 600 50 2,'646 300 2,946 1,443 1,932 600 600 50 50 2,093 2,582 300 300 2,393 2,882 Shutdown.Margin (N-l)*.9-Total Requirements Required Shutdown Margin*Excess Shutdown Margin 2,795 2,293 1,000 1,900 1,795 393 1,414 344 2,414 2,244 10000 1,900*Technical Specification 3.10"*Calculated values from Reference 5  
 
~-~.~~'-~~*~-~*~I~>>.~=~~~=~~-~-t~=~~4 h~g~~=~=Wt WI.~'~~\-~*~4~t~~~->>*~~~~=-~=~-~-~-~~-Figure 5tl R.E.Ginna Cycle 10 ARO Critical Boron Concentration vs.Exposure r 4 el~~~~~~~~~~~~~~~~~~~~~~~I~~J Ll I I~~
I
I~~~~0~~I~I~I~~~~~~~~~~
-
20 XN-NF-79-103 6.0 THERMAL HYDRAULIC DESIGN The thermal and hydraulic considerations in the Region 12 design are unchanged from those presented in Reference 4 for Region 10 fuel.
l
l~I j 21 XN-NF-79-103 7.0 ACCIDENT AND TRANSIENT ANALYSIS 7.1 PLANT TRANSIENT AND ECCS ANALYSES fOR R.E.GINNA The ECCS analysis provided in Reference 3 is applicable to all ENC fuel residing in the core during Cycle 10 operation.
 
The Plant Transient Analysis reported in XN-NF-77-40 for the R.(2)E.Ginna plant was intended to cover all anticipated ranges of values for all significant fuel dependent plant parameters for Cycle 8 and for all future 7 reloads.Table 7.1 presents a comparison of the kinetic parameters used in the Plant Transient Analysis and the parameters calculated specifically for Cycle 10.Due to the introduction of the 4 MOX assemblies the reactivity worth of the-boric acid used by the HPSIS (High Pressure Safety Injection System)and the BOC delayed neutron fraction have been calculated to be outside the range reported in the XN-NF-77-40 analysis.The analysis was reviewed and it was found that the change in boric acid worth affects the smal,l and large steamline break transients and that the delayed neutron fraction most affects the fast uncontrolled rod withdrawal transient.
XN-NF-79-103 3.0   GENERAL DESCRIPTION The R. E. Ginna   reactor consists of       121 assemblies,     each having a 14xl4 fuel rod array.     Each assembly     contains 179 fuel rods,       16 RCC   guide tubes, and 1 instrumentation tube.       The fuel rods consist of slightly enriched         U02 pellets inserted into zircaloy tubes.             The RCC guide tubes and the instr umen-tation tube are     made of SS-304L. Each ENC assembly     contains nine zircaloy spacers with Inconel springs; eight           of the spacers are located within the active fuel region.       Four   of the 121 assemblies   contain Mixed Oxide (Pu02 plus U02) bearing fuel rods.         The MOX assemblies     consist of three enrichment zones   of Pu02 utilizing natural       U02 as   the diluent.
The enveloping data for both steamline breaks are the EOC data and for the fast uncontrolled rod withdrawal are BOC data.The impact of the Cycle 10 parameters (see Table 7.1)have been evaluated for each of the transients.
The projected Cycle   10   loading pattern is     shown   in Figure 3.1 with the assemblies   identified   by their Fabrication ID's       and Region     ID's. The initial enrichments   of the various regions are listed in Table 3.1.               BOC10 exposures, based on an EOC9 exposure         of 9,570 MHD/MT,   along with Region ID's are shown in Figure 3.2.     The core   consists of     32 fresh   ENC assemblies   at 3.45 w/o   and 4 fresh Westinghouse     MOX assemblies     loaded on the periphery with 72         ENC and 13 Westinghouse     exposed assemblies       scatter-loaded     in the center portion of the core. Pertinent fuel assembly parameters for the Cycle               10 fuel are depicted in Table 3.1.       The   transuranic elements, including Am-241, have             been
The results of the evaluation for the transients were found to be nearly equivalent to the previous results and that the figure of merit for the transients were not violated, i.e.for the small steamline break the system does not go critical, for the large steamline break the MDNBR is greater than the 1.30 limit and for the uncontrolled rod withdrawal the MDNBR margin is not altered.
                                \
22 XN-NF-79-103 7.2 ROD EJECTION ANALYSIS FOR R.E.GINNA CYCLE 10 A Control Rod Ejection Accident is defined as the mechanical fail-ure 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 mechanical 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-78-53 is still applicable to Cycle 10 operation.
accounted   for up to the time of the anticipated reactor startup.
The location of the 4 MOX assemblies introduces minimal effects on ejected rod worths and hot pellet peaking factors.The e'jected rod worths and hot pellet peaking factors are cal'culated using the XTG code.No credit was taken for the powei flattening effects, of Doppler or moderator feedback in the calculation of ejected rod worths'r peaking'factors.The calculations made for Cycle 10 using XTG were two-dimensional (x-y)with appropriate axial buckling correc-tion term's.The total'eaking factors (F~)were determined as the product of the radial peaking facto'r (as calculated using XTG)and a conservative axial peaking factor;The pellet energy deposition resulting from an ejected rod was evaluated to be less than the r'esults reported in References 4 and 5.The rod ejection accident was found to result in energy deposition of less than 280 cal/gm st'ated in Regulatory Guide 1.77'and provides a greater energy deposition marg'in than that determined by Reference 4.The results of the control rod ejection transient for this case are presented in Table 7.2 along with results'from References 4 and 5.
 
23 XN-NF-79-103 7.1 R.E.Ginna Kinetic Parameters Parameters
XN-NF-79-103 Table 3.1   R. E. Ginna Cycle 10 Fuel Assembly Design Parameters Region 10                 12         MOX Enrichment, wtX U-235                 3,103    3,100    3.200    3.450    2.626*
, Moderator Temperature Coefficient (pcm/oF)Moderator Pressure Coefficient (pcm/psia)
Number  of Assemblies                   13      32      40        32      4 Pellet Densi'ty, X TD               95,0      94,0    94.0      94.0    95.0 Pellet-to -Clad Diametrical Gap, Mil                       7.5       7.5     7.5      7.5       7.5 Fuel Stack Height, inch            141.4    142.0    142.0     142.0   141.4 Region Average Burnup at BOC10, MWD/MT             24,339   17,885   8,335         0         0 Nominal Assembly Weight, KgU      392.56   373.78   373,78   373.78   395.91**
Reference Cycle (1)BOC E C 0.0-35.0+.25+.35 Cycle 10 B C-8.1+.09 E C-30.4+.35 Moderator Density Coefficient (pcm/gm/cm3)
* wtX Pu (based on assembly average)
Doppler Coeffi:cient (pcm/F)-1.25-2.00-1.35-1.84 0.0+29635.0+6858.0+25740.0 Boron Worth Coefficient (pcm/ppm)Delayed Neutron Fraction-8.75-8.72.0061.0051-7.95.0058-8.62.0052 Reference 2 XN-NF-7.9.-103 Table 7.2 Ejected Rod Worth and Peaking Factors'~Before Ejection~C1 8()~C1 9()~C1 10(HFP HZP HFP HZP HFP HZP 2,25 2.82 2.24 2.62 2.15 2.59 F~After Ejection N 4.36'.30 2.96 5.59 g g4()6 01(Maximum Rod Worth from a Full Inserted Bank (X hp)0.470 0.640 0.362 0.553 0.280 0.435 Energy Deposition (cal/gm)171 37 (1)Includes a conservative estimate of F at HFP of 1.4 and at HZP of 1.8.(2)Reference 4, calculated with XTRAN.(3)Reference 5, calculated with XTGPWR.(4)Calculated with XTGPWR.
*" in   Kg HM
25 XN-NF-79-103
 
XNrNF-79-103 K       J   I H     .G   F     E       0     C     B     A 12   HOX   12 12   '12 L14   12   L31    12    12 12    N09  L09 M14   K03 M39     L01   H01   12 12 M02   L06   N17 L19 M28   L26   N36   L05   N12   12 12 L02   M33   L21 M23   K05 N30     L24   M20   L12   12 12   L32 M40   L27   M31 K13 N07   K19   N22   L18   H13   L13    12 MOX 12  K09    M25  K20 N08  K28  N06    K18   M27   K27   12   MOX 12   L15 M15   L20   M24 K26 M05   K17   H29   L25   M38   L30   12 12 Ll 0   M18   L22 N32   K25 M21     L23   M35   L04   12 12 Ml 0   L07   N34 L28   N26 L17    M19    L08  M04    12 12    M03  L03 M37   K14 H16     Lll   Mll   12 12   12 L29   12   L16     12   12 12   MOX   12     Fabri cation or New   Fuel Region   Identification Figure 3.1 R. E. Ginna   Cycle 10 Loading Pattern
 
XN-NF-79-103 D           C       B 24,736     7,517     24,117     7,809     24,708       0 12     MOX 7;522     24,093   11,341   17,578       9,544     19,816 10                     10      12 24,117     11,342   17,506     6,926       16,061 10                     10         12 7,809     17,574     6,928   18,677       6,393       0
                                                                        ]n 10                  10                    12 24,708     9,549   16,061     6,198 10                     12 19,809     0 a BOC10   Exposure MWD/MT 12 12        10        12        12      Region ID*
0 13 MOX       12
*See Table 3.1   for Region definitions f
Figure 3.2   R. E. Ginna BOC10   quarter Core Exposure Distribution and Region ID
 
XN-NF-79-103 4.0   FUEL. SYSTEM DESIGN A description of the Exxon Nuclear supplied fuel design and design methods   is contained in Reference 1. This fuel has been specifically designed to be compatible to the resident fuel supplied by Westinghouse.
 
I 10                           XN-NF-79-103 5.0 NUCLEAR DESIGN The neutronic charact     ristics of   the projected Cycle     10 core are quite similar to those of the Cycle       9 core (see Section 5.1).
The nuclear design bases for the Cycle         10   core are as follows:
: 1)   The design   shall permit operation within the Technical Specifications for the       R. E. Ginna   plant.
: 2)   The length of Cycle     10 shall be determined on the basis of an assumed   Cycle   9 length of 9,570     MWD/MT.
: 3)   The Cycle 10     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 H
shall not exceed 1.66   (including uncertainties) in         any single fuel rod
                .through the cycle under nominal         full power operation condi-tions.
b)   The scram worth     of all rods   minus the most     reactive shall exceed   BOC and EOC shutdown requirements.
: 4)   The Cycle 10 core     shall have a negative power       coefficient.
: 5)   The MOX   assemblies   shall be located in   a region of the reactor core as to minimize the effects on shutdown margin provisions and thermal   limits.
The neutronic design methods utilized to ensure the above requirements are consistent with those described in References             6, 7, and 8.
 
XN-NF-79-103 5.1     PHYSICS CHARACTERISTICS The neutronic characteristics of the Cycle           10 core   are compared with those of Cycle         9 and   are presented     in Table 5.1. The data presented     in the table indicate         the'eutronic similarity between Cycles 9 and 10. The Cycle 10     loading pattern is applicable for Cycle 9 lengths of +700 MWD/MT and -800     MWD/MT about the nominal length of 9,570 MWD/MT.
The calculated boron letdown curve for Cycle             10 is shown   in Figure 5.1. The curve     indicates     a BOC10, no   xenon, critical boron concentration of 1,254 ppm.       At 150   MWD/MT,     equilibrium xenon, the critical boron concentration is 921 ppm.       The Cycle 10       length is projected to     be 9,500+300 MWD/MT     with 7 ppm   of boron   at EOC.
5.1.1     Power   Distribution Considerations P
Representative     predicted power     maps for   Cycle 10 are shown in Figures 5.2         and 5.3   for   BOC and   EOC conditions, respectively.       The power distributions       were obtained from a three-dimensional           model   with moderator I,
density     and Doppler feedback         effects incorporated.       For the projected Cycle 10   loading pattern the calculated             BOC   nuclear power peaking factors, F~,     N N             N F   , and Fz,     are,l.745, 1.433,       and 1.201,     respectively. At EOC conditions the corresponding values are 1.517, 1.358, and 1.098.                     The Technical     Specifi-N cation limits relative to                       N F~ and F>H,   with the measurement uncertainties backed     out, are 2.15       and   1.60. Additionally the predicted axial         F   distri-butions are well below the axially dependent Technical Specification limits N
on F~.     The BOC     F   value of 1.745 compares with the measured Cycle             9 value in Table 5.1 of 1.758.
 
12                           XN-NF 79-,103 The   control of the core     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", September 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 value of 2.30.     The Technical Specification       limit for   R. E. Ginna has a peak       of 2.32 and an   axial dependence   identical to that supported         by the procedures.       The physics characteristics of the       Ginna Cycle 10 core are       similar to those utilized in the PDC supporting analysis.         The Ginna Technical       Specification limits     on F can therefore   be protected by operation under the           PDC procedures   as stated in XN-76-40.
5.1.2   Control   Rod Reactivit     Re uirements Detailed calculations of shutdown margins for Cycle               10 are compared with Cycle     9 data in Table 5.2.       The ENC   Plant Transient Simulation (PTS) Analysis indicates that the minimum required shutdown margin is 1,800 pcm based   upon the steamline break         accident analyzed for       ENC fuel at the   EOC conditions. A value of 1,900     pcm is used   at EOC in the evaluation of the shutdown margin     to be   consistent with'the Technical Specifications.             The Cycle 10 analysis indicates excess shutdown margins               of 1,414 pcm at the BOC and 344 pcm   at the   EOC. The Cycle 9   analysis       indicates excess shut-down margins   for that cycle of 1,795         pcm at the   BOC and 393 pcm   at the   EOC.
The slightly lower   Cycle 10 excess shutdown margins, when compared to the Cycle 9 values, are due to       slightly   lower calculated rod worths.
 
13                         XN-NF-79 103 The   control- rod groups and insertion limits for Cycle             10 will remain unchanged     from Cycle 9. With these   limits the'nominal worth of the control bank, D-bank, inserted to the insertion             limits'at   HFP is   122 pcm at, BOC and'70   pcm at EOC. The control rod shutdown requirements in Table 1
5.2 allow   for a HFP   D-bank insertion equivalent to     300 pcm   for both   BOC and EOC.
5.1.3   Moderator   Tem erature Coefficient Considerations The   reference Cycle   10 design calculations indicate that the moderator temperature       coefficient is negative at all times during the cycle as shown   in Table 5.1.     This meets the Technical Specification requirement that the moderator temperature coefficient           be negative at   all times during power operation and the design         criteria that   the power coefficient be nega-tive. The least negative moderator temperature coefficient occurs at               BOC HZP and is -2.0+2pcm/     F. This compares with the     BOC9 HZP   value of -2.0 pcm/ F.
5.2   ANALYTICAL METHODOLOGY The methods   used in the Cycle   10 core analyses are described in References   6, 7, and 8.     These methods have been     verified for   both U02 and Pu02-U02   lattices.     In summary, the reference neutronic design analysis of the reload core     was performed   using the   XTG   (Reference 9) reactor simulator system-. The input exposure data were       based on   quarter core depletion calcu-lations performed from Cycle         5 to Cycle 9 using the   XTG   code. The BOC5 exposure   distribution     was obtained from plant data.       The fuel shuffling between c'ycles was accounted       for in the calculations.
 
14                     XN-NF-79-103 Predicted values of F~,   Fx , and F   were studied, with the XTG 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.
 
15                       XN-NF-79-103 Table 5.1 R. E. Ginna   Neutronics Characteristics   of Cycle   10 Compared with Cycle   9 Data C cle 9             C cle 10 BOC           EOC       BOC          EOC
(
Critical Boron
                                                  ')
HFP, ARO, Equilibrium Xenon (ppm)         961          12(            921 HZP, ARO No Xenon (ppm)               1,41 0(2)                  1,414 Moderator Temperature Coefficient HFP, (pcm/oF)                             -7.6(2)    -30.4          -8.1    -30.4 HZP, (pcm/oF)                             -2.0        -21.5          -2.0    -21.6 Doppler Coefficient, (pcm/ F)               -1.25 to      -2.0        -,1.35    -1.84 Boron Worth, (pcm/ppm)
HFP                                    -8.12         -8.72        -7.95    ,-8.62 HZP                                    -8.58 Total Nuclear Peaking Factor Fq, HFP                                1.758(                     1.745      1.517 Delayed Neutron Fraction                    .0061        .0051        .0058      .0052 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm)        5,751    )  5,821.       5';341    5,696 Excess  Shutdown Margin (pcm)              1,795(    )      393(      1,414          344 Moderator Pressure   Coefficient (pcm/psi)                                             0.35                    0.35 (1)   Extrapolated from measured data (2)   Measured Data (3)   70/ Power Map (4)   Reference 5
 
16                         XN-NF-79-103 Table 5.2, R. E. Ginna Control     Rod Shutdown   Margins and Requirements   for Cycle 10 C cle 9 **         C cle 10 BOC           EOC      BOC        EOC Control   Rod Worth   HZP , cm All Rods Inserted (ARI)               6,407        6,634      5,949    6,420 ARI  less most reactive (N-1)           5,751        5,821      5,341    5,696 N-1 less lOX allowance L(N-1)* 9l                             5,176         5,239       4,807      5,125 Reactivit    Insertion    cm Moderator plus Doppler                    1,431        1,996      1,443      1,932 Flux Redistribution                         600          600        600      600 Void                                         50            50        50        50 Sum of the above three                 2,081        2,'646     2,093     2,582 Rod  Insertion Allowance                    300         300         300      300 Total Requirements                              2,381        2,946      2,393     2,882 Shutdown. Margin (N-l)*.9 - Total     Requirements         2,795       2,293       2,414      2,244 Required Shutdown Margin*                      1,000         1,900     10000      1,900 Excess Shutdown Margin                          1,795           393     1,414         344
* Technical Specification 3.10
"* Calculated     values from Reference   5
 
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                                                                                                                    ~ ~
Figure 5tl     R. E. Ginna       Cycle 10   ARO Critical Boron Concentration vs. Exposure r                                                                 4
 
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20                     XN-NF-79-103 6.0 THERMAL HYDRAULIC DESIGN The thermal and hydraulic considerations in the Region 12 design are unchanged from those presented in Reference 4 for Region 10 fuel.
 
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21                           XN-NF-79-103 7.0   ACCIDENT AND TRANSIENT ANALYSIS 7.1   PLANT TRANSIENT AND ECCS ANALYSES fOR R. E. GINNA The ECCS   analysis provided in Reference     3 is applicable to all ENC fuel residing in the core during Cycle       10 operation.
(2) for the The Plant Transient Analysis reported in XN-NF-77-40                         R.
E. Ginna   plant was   intended to cover   all anticipated   ranges   of values for all significant fuel     dependent   plant parameters for Cycle       8 and for all future     7 reloads. Table 7.1 presents     a comparison of the kinetic parameters       used in the Plant Transient Analysis and the parameters calculated specifically                 for Cycle 10. Due to the introduction of the     4 MOX   assemblies   the reactivity worth of the -boric acid used by the HPSIS (High Pressure           Safety Injection System) and the     BOC delayed neutron   fraction   have been   calculated to   be outside the range reported in the XN-NF-77-40 analysis.             The analysis   was reviewed and   it was   found that the change in boric acid worth affects the smal,l and large steamline break transients       and that the delayed neutron fraction   most affects the fast uncontrolled rod withdrawal transient.
The enveloping data     for both steamline breaks are the       EOC data and for the fast uncontrolled       rod withdrawal are   BOC data. The impact   of the Cycle 10 parameters       (see Table 7.1) have been evaluated       for each of the transients. The results of the evaluation for the transients           were found   to be nearly equivalent to the previous results         and that the figure of merit for the transients were not violated,       i.e. for the small steamline break the system does not go     critical, for the large steamline       break the   MDNBR is greater than the     1.30 limit and for the uncontrolled         rod withdrawal the     MDNBR margin   is not altered.
 
22                           XN-NF-79-103 7.2   ROD EJECTION ANALYSIS FOR R. E. GINNA CYCLE 10 A Control   Rod Ejection Accident is defined         as the mechanical     fail-ure 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 mechanical     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-78-53       is still   applicable to Cycle     10 operation. The location of the   4 MOX assemblies   introduces minimal effects       on ejected rod worths     and hot pellet peaking factors.       The e'jected rod worths and       hot pellet peaking factors are cal'culated using the         XTG code. No credit was taken for the   powei flattening effects, of Doppler or moderator           feedback in the calculation       of ejected rod   worths'r     peaking 'factors.     The calculations   made for Cycle 10 using XTG   were two-dimensional       (x-y) with appropriate axial buckling correc-tion term's. The total'eaking factors       (F~) were determined as the product         of the radial peaking facto'r (as calculated using             XTG) and a   conservative axial peaking   factor;   The pellet   energy deposition resulting from an ejected rod was evaluated to be less than the r'esults reported in References 4 and 5.
The rod   ejection accident   was found     to result in energy deposition of less than 280 cal/gm st'ated in Regulatory Guide 1.77 'and provides a greater energy deposition marg'in than that determined by Reference 4.               The results of the control rod ejection transient for this case are presented in Table 7.2 along with results 'from References         4 and 5.
 
23                     XN-NF-79-103 7.1 R. E. Ginna Kinetic Parameters Reference Cycle (1)        Cycle 10 Parameters                     BOC      E C        B C          E C
, Moderator Temperature Coefficient (pcm/oF)                               0.0    -35.0       -8.1         -30.4 Moderator Pressure  Coefficient (pcm/psia)                           +.25    +.35        +.09          +.35 Moderator Density Coefficient (pcm/gm/cm3)                          0.0  +29635.0    +6858.0      +25740.0 Doppler Coeffi:cient (pcm/ F)         -1.25     -2.00     -1.35         -1.84 Boron Worth  Coefficient (pcm/ppm)                           -8.75     -8.72     -7.95         -8.62 Delayed Neutron Fraction              .0061    .0051      .0058          .0052 Reference 2
 
XN-NF-7.9.-103 Table 7.2 Ejected   Rod Worth and   Peaking Factors'~
                                      ~C1   8( )
                                                        ~C1   9(         )
                                                                                  ~C1   10(
HFP       HZP   HFP                 HZP   HFP         HZP Before Ejection                  2,25     2.82   2.24     2.62             2.15     2.59 N
F~ After Ejection                   4.36   '.30     2.96     5.59             g g4( )   6 01(
Maximum Rod Worth   from a Full Inserted   Bank (X hp)         0.470   0.640   0.362     0.553           0.280     0.435 Energy Deposition (cal/gm)           171     37 (1)   Includes a conservative estimate of F     at HFP of 1.4         and   at HZP   of 1.8.
(2) Reference 4, calculated with XTRAN.
(3) Reference 5, calculated with XTGPWR.
(4) Calculated with XTGPWR.
 
25                     XN-NF-79-103


==8.0 REFERENCES==
==8.0 REFERENCES==
: 1. XN-NF-77-52, "R. E. Ginna Reload Fuel Design", November, 1977.
: 2.  'XN-NF-77-40, "Plant Transient Analysis for R. E. Ginna, Unit    1 Nuclear Power Plant", Revision 1, July, 1979.
: 3. XN-NF-77-58, "ECCS Analysis  for the R. E. Ginna Reactor with  ENC WREM-II  PWR Evaluation  Model",  December, 1977.
: 4. XN-NF-77-53, "R. E. Ginna Nuclear Plant Cycle 8 Safety Analysis Report", December, 1977.
: 5. XN-NF-78-50, ="R. E. Ginna Cycle  9 Safety Analysis Report,"
December, 1978.
: 6. F. B. Skogen, "Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April, 1977.
: 7. XN-75-27(A), Supplement  1 to Reference 6, April, 1977.
: 8. XN-75-27, Supplement  2 to Reference 6, December, 1977.
: 9. XN-CC-28, Rev. 3, "XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January, 1975.
I l
x I
26                      XN-NF-79-103 R. E. GINNA CYCLE 10 RELOAD SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES DISTRIBUTION K. H. Blank                                L. A. Nielsen G. J. Busselman                            G. F. Owsley L. J. Federico                            J. F. Patterson R. L. Feuerbacher                          A. W. Prichard R. G. Grummer                              F. B. Skogen B. L. Johnson  (2)                        G. A. Sofer M. R. Killgore                            A. V. Wojchouski T. L. Krysinski                            C. H. Wu C. E. Leach                                RG&E/L. J. Federico (80)
J. N. Morgan                              Document Control (10)
W. S. Nechodom


1.XN-NF-77-52,"R.E.Ginna Reload Fuel Design", November, 1977.2.3.4.5.6.'XN-NF-77-40,"Plant Transient Analysis for R.E.Ginna, Unit 1 Nuclear Power Plant", Revision 1, July, 1979.XN-NF-77-58,"ECCS Analysis for the R.E.Ginna Reactor with ENC WREM-II PWR Evaluation Model", December, 1977.XN-NF-77-53,"R.E.Ginna Nuclear Plant Cycle 8 Safety Analysis Report", December, 1977.XN-NF-78-50,="R.E.Ginna Cycle 9 Safety Analysis Report," December, 1978.F.B.Skogen,"Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April, 1977.7.XN-75-27(A), Supplement 1 to Reference 6, April, 1977.8.XN-75-27, Supplement 2 to Reference 6, December, 1977.9.XN-CC-28, Rev.3,"XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January, 1975.
l I
I l x I 26 XN-NF-79-103 R.E.GINNA CYCLE 10 RELOAD SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES DISTRIBUTION K.H.Blank G.J.Busselman L.J.Federico R.L.Feuerbacher R.G.Grummer B.L.Johnson (2)M.R.Killgore T.L.Krysinski C.E.Leach J.N.Morgan W.S.Nechodom L.A.Nielsen G.F.Owsley J.F.Patterson A.W.Prichard F.B.Skogen G.A.Sofer A.V.Wojchouski C.H.Wu RG&E/L.J.Federico (80)Document Control (10) l I 1}}
1}}

Revision as of 20:01, 29 October 2019

Cycle 10 SAR W/Mixed Oxide Assemblies.
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0 KN-N 103 P E (lllklk PlljlllLEkk PILklll'NllllLEJIB MFET7 kkkL7SIIS PEPBP'll'IIYM Iw3IIXEB QXIIBE bhSSEliNBILIIES DECEMBER 1979 RICHLAND, NA 99352

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0: XN-NF-79-103 IR/R14 79 R. E. GINNA NUCLEAR PLANT CYCLE 10.

SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES Prepared:

G. J. Buss man, Manager Neutronics and Fuel Management Approved:

G. A . Sofe nager Nuclear Fue s Engineering Concurred:

J. N. Morgan, Manage Licensing and Safety Engineering Concurred: l'~/ 7 L. J. Federico, Manager Nuclear Fuels Project E)j(ON NUCLEAR COMPANY, Inc.

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT Ir PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being Sub.

mitted by Exxon Nuclear to the USNRC as part of a technical contri.

bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear. fabricated reioarl fuel or other teclmical services provided by Exxon Nuclear for lieht water power reactors anH, with the measurement uncertainties backed out, are 2.15 and 1.60. Additionally the predicted axial F distri-butions are well below the axially dependent Technical Specification limits N

on F~. The BOC F value of 1.745 compares with the measured Cycle 9 value in Table 5.1 of 1.758.

12 XN-NF 79-,103 The control of the core 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", September 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 value of 2.30. The Technical Specification limit for R. E. Ginna has a peak of 2.32 and an axial dependence identical to that supported by the procedures. The physics characteristics of the Ginna Cycle 10 core are similar to those utilized in the PDC supporting analysis. The Ginna Technical Specification limits on F can therefore be protected by operation under the PDC procedures as stated in XN-76-40.

5.1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 10 are compared with Cycle 9 data in Table 5.2. The ENC Plant Transient Simulation (PTS) Analysis indicates that the minimum required shutdown margin is 1,800 pcm based upon the steamline break accident analyzed for ENC fuel at the EOC conditions. A value of 1,900 pcm is used at EOC in the evaluation of the shutdown margin to be consistent with'the Technical Specifications. The Cycle 10 analysis indicates excess shutdown margins of 1,414 pcm at the BOC and 344 pcm at the EOC. The Cycle 9 analysis indicates excess shut-down margins for that cycle of 1,795 pcm at the BOC and 393 pcm at the EOC.

The slightly lower Cycle 10 excess shutdown margins, when compared to the Cycle 9 values, are due to slightly lower calculated rod worths.

13 XN-NF-79 103 The control- rod groups and insertion limits for Cycle 10 will remain unchanged from Cycle 9. With these limits the'nominal worth of the control bank, D-bank, inserted to the insertion limits'at HFP is 122 pcm at, BOC and'70 pcm at EOC. The control rod shutdown requirements in Table 1

5.2 allow for a HFP D-bank insertion equivalent to 300 pcm for both BOC and EOC.

5.1.3 Moderator Tem erature Coefficient Considerations The reference Cycle 10 design calculations indicate that the moderator temperature coefficient is negative at all times during the cycle as shown in Table 5.1. This meets the Technical Specification requirement that the moderator temperature coefficient be negative at all times during power operation and the design criteria that the power coefficient be nega-tive. The least negative moderator temperature coefficient occurs at BOC HZP and is -2.0+2pcm/ F. This compares with the BOC9 HZP value of -2.0 pcm/ F.

5.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 10 core analyses are described in References 6, 7, and 8. These methods have been verified for both U02 and Pu02-U02 lattices. In summary, the reference neutronic design analysis of the reload core was performed using the XTG (Reference 9) reactor simulator system-. The input exposure data were based on quarter core depletion calcu-lations performed from Cycle 5 to Cycle 9 using the XTG code. The BOC5 exposure distribution was obtained from plant data. The fuel shuffling between c'ycles was accounted for in the calculations.

14 XN-NF-79-103 Predicted values of F~, Fx , and F were studied, with the XTG 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.

15 XN-NF-79-103 Table 5.1 R. E. Ginna Neutronics Characteristics of Cycle 10 Compared with Cycle 9 Data C cle 9 C cle 10 BOC EOC BOC EOC

(

Critical Boron

')

HFP, ARO, Equilibrium Xenon (ppm) 961 12( 921 HZP, ARO No Xenon (ppm) 1,41 0(2) 1,414 Moderator Temperature Coefficient HFP, (pcm/oF) -7.6(2) -30.4 -8.1 -30.4 HZP, (pcm/oF) -2.0 -21.5 -2.0 -21.6 Doppler Coefficient, (pcm/ F) -1.25 to -2.0 -,1.35 -1.84 Boron Worth, (pcm/ppm)

HFP -8.12 -8.72 -7.95 ,-8.62 HZP -8.58 Total Nuclear Peaking Factor Fq, HFP 1.758( 1.745 1.517 Delayed Neutron Fraction .0061 .0051 .0058 .0052 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm) 5,751 ) 5,821. 5';341 5,696 Excess Shutdown Margin (pcm) 1,795( ) 393( 1,414 344 Moderator Pressure Coefficient (pcm/psi) 0.35 0.35 (1) Extrapolated from measured data (2) Measured Data (3) 70/ Power Map (4) Reference 5

16 XN-NF-79-103 Table 5.2, R. E. Ginna Control Rod Shutdown Margins and Requirements for Cycle 10 C cle 9 ** C cle 10 BOC EOC BOC EOC Control Rod Worth HZP , cm All Rods Inserted (ARI) 6,407 6,634 5,949 6,420 ARI less most reactive (N-1) 5,751 5,821 5,341 5,696 N-1 less lOX allowance L(N-1)* 9l 5,176 5,239 4,807 5,125 Reactivit Insertion cm Moderator plus Doppler 1,431 1,996 1,443 1,932 Flux Redistribution 600 600 600 600 Void 50 50 50 50 Sum of the above three 2,081 2,'646 2,093 2,582 Rod Insertion Allowance 300 300 300 300 Total Requirements 2,381 2,946 2,393 2,882 Shutdown. Margin (N-l)*.9 - Total Requirements 2,795 2,293 2,414 2,244 Required Shutdown Margin* 1,000 1,900 10000 1,900 Excess Shutdown Margin 1,795 393 1,414 344

"* Calculated values from Reference 5

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20 XN-NF-79-103 6.0 THERMAL HYDRAULIC DESIGN The thermal and hydraulic considerations in the Region 12 design are unchanged from those presented in Reference 4 for Region 10 fuel.

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21 XN-NF-79-103 7.0 ACCIDENT AND TRANSIENT ANALYSIS 7.1 PLANT TRANSIENT AND ECCS ANALYSES fOR R. E. GINNA The ECCS analysis provided in Reference 3 is applicable to all ENC fuel residing in the core during Cycle 10 operation.

(2) for the The Plant Transient Analysis reported in XN-NF-77-40 R.

E. Ginna plant was intended to cover all anticipated ranges of values for all significant fuel dependent plant parameters for Cycle 8 and for all future 7 reloads. Table 7.1 presents a comparison of the kinetic parameters used in the Plant Transient Analysis and the parameters calculated specifically for Cycle 10. Due to the introduction of the 4 MOX assemblies the reactivity worth of the -boric acid used by the HPSIS (High Pressure Safety Injection System) and the BOC delayed neutron fraction have been calculated to be outside the range reported in the XN-NF-77-40 analysis. The analysis was reviewed and it was found that the change in boric acid worth affects the smal,l and large steamline break transients and that the delayed neutron fraction most affects the fast uncontrolled rod withdrawal transient.

The enveloping data for both steamline breaks are the EOC data and for the fast uncontrolled rod withdrawal are BOC data. The impact of the Cycle 10 parameters (see Table 7.1) have been evaluated for each of the transients. The results of the evaluation for the transients were found to be nearly equivalent to the previous results and that the figure of merit for the transients were not violated, i.e. for the small steamline break the system does not go critical, for the large steamline break the MDNBR is greater than the 1.30 limit and for the uncontrolled rod withdrawal the MDNBR margin is not altered.

22 XN-NF-79-103 7.2 ROD EJECTION ANALYSIS FOR R. E. GINNA CYCLE 10 A Control Rod Ejection Accident is defined as the mechanical fail-ure 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 mechanical 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-78-53 is still applicable to Cycle 10 operation. The location of the 4 MOX assemblies introduces minimal effects on ejected rod worths and hot pellet peaking factors. The e'jected rod worths and hot pellet peaking factors are cal'culated using the XTG code. No credit was taken for the powei flattening effects, of Doppler or moderator feedback in the calculation of ejected rod worths'r peaking 'factors. The calculations made for Cycle 10 using XTG were two-dimensional (x-y) with appropriate axial buckling correc-tion term's. The total'eaking factors (F~) were determined as the product of the radial peaking facto'r (as calculated using XTG) and a conservative axial peaking factor; The pellet energy deposition resulting from an ejected rod was evaluated to be less than the r'esults reported in References 4 and 5.

The rod ejection accident was found to result in energy deposition of less than 280 cal/gm st'ated in Regulatory Guide 1.77 'and provides a greater energy deposition marg'in than that determined by Reference 4. The results of the control rod ejection transient for this case are presented in Table 7.2 along with results 'from References 4 and 5.

23 XN-NF-79-103 7.1 R. E. Ginna Kinetic Parameters Reference Cycle (1) Cycle 10 Parameters BOC E C B C E C

, Moderator Temperature Coefficient (pcm/oF) 0.0 -35.0 -8.1 -30.4 Moderator Pressure Coefficient (pcm/psia) +.25 +.35 +.09 +.35 Moderator Density Coefficient (pcm/gm/cm3) 0.0 +29635.0 +6858.0 +25740.0 Doppler Coeffi:cient (pcm/ F) -1.25 -2.00 -1.35 -1.84 Boron Worth Coefficient (pcm/ppm) -8.75 -8.72 -7.95 -8.62 Delayed Neutron Fraction .0061 .0051 .0058 .0052 Reference 2

XN-NF-7.9.-103 Table 7.2 Ejected Rod Worth and Peaking Factors'~

~C1 8( )

~C1 9( )

~C1 10(

HFP HZP HFP HZP HFP HZP Before Ejection 2,25 2.82 2.24 2.62 2.15 2.59 N

F~ After Ejection 4.36 '.30 2.96 5.59 g g4( ) 6 01(

Maximum Rod Worth from a Full Inserted Bank (X hp) 0.470 0.640 0.362 0.553 0.280 0.435 Energy Deposition (cal/gm) 171 37 (1) Includes a conservative estimate of F at HFP of 1.4 and at HZP of 1.8.

(2) Reference 4, calculated with XTRAN.

(3) Reference 5, calculated with XTGPWR.

(4) Calculated with XTGPWR.

25 XN-NF-79-103

8.0 REFERENCES

1. XN-NF-77-52, "R. E. Ginna Reload Fuel Design", November, 1977.
2. 'XN-NF-77-40, "Plant Transient Analysis for R. E. Ginna, Unit 1 Nuclear Power Plant", Revision 1, July, 1979.
3. XN-NF-77-58, "ECCS Analysis for the R. E. Ginna Reactor with ENC WREM-II PWR Evaluation Model", December, 1977.
4. XN-NF-77-53, "R. E. Ginna Nuclear Plant Cycle 8 Safety Analysis Report", December, 1977.
5. XN-NF-78-50, ="R. E. Ginna Cycle 9 Safety Analysis Report,"

December, 1978.

6. F. B. Skogen, "Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors", XN-75-27(A), Exxon Nuclear Company, April, 1977.
7. XN-75-27(A), Supplement 1 to Reference 6, April, 1977.
8. XN-75-27, Supplement 2 to Reference 6, December, 1977.
9. XN-CC-28, Rev. 3, "XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", January, 1975.

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26 XN-NF-79-103 R. E. GINNA CYCLE 10 RELOAD SAFETY ANALYSIS REPORT WITH MIXED OXIDE ASSEMBLIES DISTRIBUTION K. H. Blank L. A. Nielsen G. J. Busselman G. F. Owsley L. J. Federico J. F. Patterson R. L. Feuerbacher A. W. Prichard R. G. Grummer F. B. Skogen B. L. Johnson (2) G. A. Sofer M. R. Killgore A. V. Wojchouski T. L. Krysinski C. H. Wu C. E. Leach RG&E/L. J. Federico (80)

J. N. Morgan Document Control (10)

W. S. Nechodom

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