ML20151R790

Rev 2 to Hb Robinson Unit 2,Cycle 11 Sar

ML20151R790
Person / Time
Site:	Robinson
Issue date:	01/10/1986
From:	Adams F, Stone I, Stricker M SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML14192A743	List: ML20151R790
References
XN-NF-85-103, XN-NF-85-103-R02, XN-NF-85-103-R2, NUDOCS 8602060235
Download: ML20151R790 (53)

Text

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Xh-N =-85-103 i REVISION 2 I,

.B. 9031\!SO \ U \

2, CYC _E 11 SA=E- Y A\ A_YSIS RE30T i

JANUARY 1986 l 9 C _AN J, WA 99352 ,

i EXXON NUCLEAR COMPANY, INC.

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XN-NF-85-103 Revision 2 Issue Date 1/10/86 f H.B. ROBINSON UNIT 2, CYCLE 11 SAFETY ANALYSIS REPORT t

v Written by: ,-

I. W 5tche, % clear Engineer PWR Neutronics Written by: .1 [ # [7A76 1% T.( Adams, Team Leader PWR Safety Analysis Written by: '41 k Skwkn / - F - }%

M. 5. Stricker 3 PWR Safety Analysis Approved by: , //c7 fl J // 5'. Holm, Manager PM Safety Analysis

• Approved by: Wj ///v/r0 T. W. Patten, Manager Neutronics & Fuel Management Approved-by: Y 4 H. E.

[Williamson, Manager

, , _ . ///: /,1 Licensing & Safety Engineering Approved by- # ,s m /[/'/r c-G. J. BuYselmin, Manager Fuel Design Approved by: / I t f

//C[US G. L. Ritter, Manager Fuel Engineering & Technical Services Concurred by: 1! Fr e ~ 87 J. N. Morgan, fianager /

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Proposals & t6stomer Slirvic/ es Engineering wrg ERON NUCLEAR COMPANY INC. .

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CUSTOMER DISCLAIMER t

IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLE ASE READ CAREFULLY Exxon Nucieer Coneeny's warrantms and representsoons concommg the subsect metter of this document are those not form in the Agreement f between Exxon Nucteer Company, Inc. and the Customer purwant to which the document e neued. Aa:onfmgly, except as otherwise empressly provided in such Agreement, neither Emmon Nuclear Company, Inc, not any person l actmg on :ts behelf maket any warranty or representauon, expreened or irnpiiod, with roepect to the accuracy, completonces, or usefulness of the .

mformecon contemed in that document, or that the use of any informtoon, apparotus, rnseed or process diecioned in the document well not mfnnge privsesty oumed rights; or aneumes any lintuhtus with roepect to the use of any enformecon, apparetus, metod or process decioned an thre document The mformenon centeened herein is for the sole use of Customer.

In order to avoed impearment of rights of Eason Nucteer Company, Inc.

in potents or smennone which may be encluded en the mformata conteened m eis document, the recesent, by its esceptance of thes document eyees not to putdish or make public use (in the pesent use of the term) of such information until so authortaed in wertirig by Emmon Nucteer Company, Inc.

or unal efter en (8) months followsng termmecon or superate of the aforossed Ayeoment and any ensenssort thereof, untees otherwise expreesty provided in the Agreement. No rights or hcenses m or to arty potents are imphed by the fumshes of the document,

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s i XN-NF-85-103 Revision 2 TABLE OF CONTENTS f

Page

1.0 INTRODUCTION

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2.0

SUMMARY

...................................................... 2 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE..................... 3 4.0 GENERAL DESCRIPTION.......................................... 6 5.0 MECHANICAL 0ESIGN............................................ 11 6.0 NUCLEAR CORE DESIGN.......................................... 12 6.1 PHYSICS CHARACTERISTICS...................................... 13 6.2 P0 DER DISTRIBUTION CONTROL PROCEDURES........................ 15 6.3 ANALYTICAL METHODOLOGY....................................... 15 7.0 THERMAL-HYORAULIC DESIGN..................................... 22 8.0 FSAR CHAPTER 15 EVENTeANALYSIS............................... 23

9.0 REFERENCES

................................................... 43 i

ii XM-NF-85-103 Revision 2 LIST OF TABLES Table Pace 4.1 H. B. Robinson Unit 2, Cycle 11, Fuel Assembly Design Parameters..................................................... 8 ,

6.1 H. B. Robinson Unit 2, Neutronics Characteristics of Cycle 11 Compared with Cycle 10 Data........................... 16 6.2 !!. B. Robinson Unit 2, Control Rod Shutdown Margin and Requirements for Cycle 11...................................... 17 8.1 Plant Rated Operating Conditions............................... 36 j 8.2 Core and Fuel Design Parameters Used in Chapter 15 Event Analysis................................................. 37 8.3 Bounding Physics Parameters for Cycles 10 and 11............... 38 8.4 Bounding Power Peaking Factors and Augmentation Factors for Cycles 10 and 11................................... 39 8.5 H.B. Robinson Unit 2 Cycle 11 Results of the Analyses of CVCS Malfunction (Boron Dilution)........... ..... 40 8.6 H.B. Robinson Unit 2 Cycle 11 Ejected Rod Analysis, HFP.................................................. 41 8.7 H.B. Robinson Unit 2 Cycle 11 Ejected Rod Analysis, HZP.................................................. 42 A

iii XN-NF-85-103 Revision 2 LIST OF FIGURES Figure Page 3.1 H. B. Robinson Unit 2, Critical Boron Concentration versus Exposure for Cycle 10 at 2,300 MWt............................. 4 3.2 H.B. Robinson Unit 2, Cycle 10 Power Distribution Map at 98.6% of 2,300 MWt with Bank D at 219 Steps for 6,516 mwd /MTU of Cycle Exposure............................ 5 4.1 H. B. Robinson Unit 2 Cycle 11, Reference Loading Pattern for an E0C10 Exposure of 11,150 mwd /MTU........................ 9 4.2 H. B. Robinson Unit 2, BOC11 and EOC11 Quarter Core Exposure Distribution and Region ID for E0C = 11,150 mwd /MTU... 10 6.1 H. B. Robinson Unit 2, Cycle 11, HFP Critical Boron Concentration versus Cycle Exposure for 2,300 MWt.............. 18 6.2 H. B. Robinson Unit 2 Cycle 11, Assembly Relative Power Distribution, 100 mwd /MTV, 2,300 MWt, 3-D XTGPWR............... 19 6.3 H. B. Robinson Unit 2' Cycle 11, Assembly Relative Power Distribution, 7,000 mwd /MTU, 2,300 MWt, 3-D XTGPWR............. 20

( 6.4 H. B. Robinson Unit 2 Cycle 11, Assembly Relative Power Distribution, 11,530 mwd /MTU, 2,300 MWt, 3-D XTGPWR............ 21 f

iv XN-NF-85-103 Revision 2 H. B. ROBINSON UNIT 2, CYCLE 11

> SAFETY ANALYSIS REPORT PROLOGUE ,

This report is a revision to the third (3) in a series of f,ive (5) reports which address the neutronic characteristics of the H. B. Robinson Unit 2, Cycle 11 core. A preliminary assessment of the uranium requirements for Region 14 (XN-8) was provided in rosponse to the Tentative Scheduled Delivery Date (TSDD) notice (PWR:018:84). Subsequently, a response to the Final Scheduled Delivery Date (FSDD) notice was provided (PWR:029:84). The third report in this series was the Cycle 11 Safety Analysis Report (XN-NF 103). This was followed by the Cycle 11 Fuel Cycle Design Report (XN-NF ~ 109(P)). The final report in this series will be the Cycle 11 Startup and Operations Report.

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a 1 XN-NF-85-103 Revision 2

1.0 INTRODUCTION

Results of the safety evaluation for Cycle 11 of the H. B. hbinson 'Jnit 2 nuclear plant are presented in this report. The Cycle 11 analysis reflects plant operation at 2,300 MWt. The topics addressed herein include operating history of the reference cycle, power distribution considerations, control rod reactivity requirements, temperature coef-ficient considerations, setpoint analysis, and Standard Review Plan Chapter 15 Event analysis, including LOCA/ECCS.

The Cycle 11 design requires the loading of forty-eight (48) fresh Exxon Nuclear Company (ENC) supplied fuel assemblies. The forty-eight (48) fresh XN-8 fuel assemblies utilize nataral uranium axial blankets (NUABs) with forty (40) of the assemblies also containing gadolinia-bearing fuel rods.

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2 XN-NF-85-103 Revision 2 2.0

SUMMARY

Cycle 11 of the H. B. Robinson Unit 2 nuclear plant is designed to operate at 2,300 MWt starting in'the spring of 1986. The characteristics of the fuel and of the reload cora reselt in conformance with required shutdown margins and thermal limits. This document provides the safety analysis for the plant during Cycle 11 operation.

The ENC fuel mechanical design is presented in References 1 and 2. The generic Control Rod Ejection Analysis is provided in Reference 16. These analyses are all applicable to the Cycle 11 operating conditions at 2,300 MWt.

The thermal-hydraulic design of the Cycle 11 core is discussed in Section 7.0 of this document. A review of the applicability to Cycle 11 of the current analysis of record for the Standard Review Plan Chapter 15 events and reactor trip setpoint analyses is provided in Section 8.0 of this document.

Plant transient and setpoint analyses reported and reviewed here support operation at 2300 MWt within the licensed limits. The LOCA/ECCS analysis of record is discussed in Section 8.0.

H. 8. Robinson Cycle 10 has been chosen as the reference neutronics cycle due to the close resemblance of the overall neutronic characteristics of Cycle 11 to Cycle 10.

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3 XN-NF-85-103 Revision 2 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE H. B. Robinson Cycle 10 has been chosen as the reference neutronics cycle due to the close resemblance of the overall neutronic characteristics of Cycle 11 to Cycle 10.

The measured power peaking factors have remained within the Technical Specification limits for Cycle 10. The total nuclear peaking f actor, F Tg, and the radial nuclear pin peaking f actor FAH, have remained below 2.32 and 1.65, respectively. Cycle 10 operation has typically been rod free with Control Bank 0 positioned in the range of 220 steps; 228 steps being fully withdrawn.

It is anticipated that similar control bank insertions will be seen in Cycle 11.

The Cycle 10 critical boron concentration as calculated by ENC is in good agreement when compared to the observed values (see Figure 3.1). A power distribution calculated with the PDQ model is compared to measured values shown in Figure 3.2. The comparison is made at a Cycle 10 exposure of 6,516 mwd /MTU and 0-bank inserted to 219 steps for a core power level of 98.6% of 2,300 MWt.

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?8 Cycle Exposure, GWd/MTU s's am Figure 3.1 H. B. Robinson Unit 2, Critical Baron Concentration versus Exposure for Cycle 10 at 2,300 MWt Nb "

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XN-fiF-85-103 Revision 2 H G F E D C 8 A

, l.hkb k bhb b 9hh k$h4h k kkh k bhh k hk7 b$h4b 1.314 1.060 0.925 1.338 1.126 1.032 1.237. 0.348 8 0.3 0.8 1.1 0.4 -1.0 -1.0 -1.6 -2.3 1.074 1.059 1.063 0.979 1.130 1.160 1.155 0.273 1.061 1.050 1.061 0.977 1.130 1.174 1.173 0.280 9 1.2 0.9 0.2 0.2 0.0 -1.2 -1.5 -2.5 0.938 1.067 1.009 1.390 1.028 0.986 0.924 0.925 1.061 1.011 1.382 1.019 0.992 0.942 10 1.4 0.6 -0.2 0.6 0.9 -0.6 -1.9 1.347 0.975 1.380 1.063 1.190 1.188 0.728 1.339 0.982 1.384 1.050 1.177 1.179 0.729 11 0.6 -0.7 -0.3 0.3 1.1 0.8 0.1 1.119 1.121 1.017 1.184 1.215 0.562 1.128 1.129 1.022 1.179 1.191 0.549 12

-0.8 -0.7 -0.5 0.4 2.0 2.4

! 1.037 1.168 0.982 1.181 0.561 1.033 1.175 0.993 1.180 0.550 13 0.4 -0.6 -1.1 0.1 2.0 1.242 1.172 0.927 0.735 Measured 1.238 1.173 0.942 0.729 Calculated 14 0.3 -0.1 -1.6 0.8 (M-C)/C*100 0.347 0.273

( 0.349 0.280 15

% -0.6 -2.5

( Figure 3.2 H. B. Robinson Unit 2, Cycle 10 Power Distribution Map at 98.6% of 2,300 MWt with Bank 0 at 219 Steps for 6,516 mwd /MTU of Cycle Exposure E

6 XN-NF-85-103 Revision 2 r

4.0 GENERAL DESCRIPTION The H. B. Robinson reactor consists of 157 assemblies, each having a 15x15 fuel rod array. Each assembly contains 204 fuel rods, twenty RCC guide tubes,. and one instrumentation tube. The RCC guide tubes and the instrumentation tube are made of zircaloy. Each ENC assembly contaihs seven zirchloy spacers with Inconel springs; six of the spacers are located

' within the active fuel region. The fuel rods consist of slightly enriched U02 pellets inserted into zircaloy tubes.

The Cycle 11 design reflects the loading of forty-eight (48) fresh ENC supplied fuel assemblies. This core design contains the second reload of axially blanketed fuel and the third reload of gadolinia-bearing fuel pins for H. B. Robinson Unit 2. Forty (40) of the forty-eight (48) fresh Region 14 (XN-8) fuel assemblies contain gadolinia-bearing pins.

The design for Batch XN-8, Region 14, includes natural uranium axial blankets in the top and bottom six (6) inches of the active fuel region of the nongadolinia-bearing pins. In the three Fundred and eighty-four (384) gadolinia-bearing fuel pins, extended natural uranium blankets have been incorporated. The additional six (6) inches of natural uranium has been added to the gadolinia-bearing fuel pins to flatten the axial power distribution.

The batch average enrichment for the blanketed assemblies is 3.42 w/o U-235. This average enrichment is achieved by using a central axial zone enrithment of 3.73 w/o in fuel pins which contain no gadolinia and 2.60 w/o in the gadolinia-bearing fuel pins. In sixteen (16) Region 14 assemblies, twelve (12) fuel pins per assembly will contain 4 w/o gadolinia. In eight i

(8) Region 14 assemblies, twelve (12) fuel pins per assembly will contain 6 w/o gadolinia. In another eight (8) assemblies, eight (8) pins of 4 w/o

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gadolinia per assembly will be utilized. In addition, in eight (8) Region

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14 assemblies, four (4) pins per assembly will contain 4 w/o' gadolinia. In ,

the remainin'g eight (8) Region 14 (XN-8) fuel assemblies no gadolinia pins will be used. Thus, the total number of gadolinia pins required for Cycle 11 is 384 l

The projected Cycle 11 loading pattern is shown in Figure 4.1 with the l assemblies identified by assembly fabrication ID and by their core '

location in the previous cycle or by fresh fuel region. BOC11 exposures, based on an EOC10 exposure of 11,150 mwd /MTU, along with Region ID's, are j shown in quarter core representation in Figure 4.2. The initial enrichments in the various fuel regions are listed in Table 4.1. Also included in Table 4.1 are the peak assembly exposures by fuel region and fuel type.

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Table 4.1 H. 8. Robinsan Unit 2 Cycle ll, Fuel Assembly Design Parameters PLSA 12 128 12* 12** 12*** 12C*** 13 13** 13*** 14 14* 148* 14*** 14****

IN Number 7 6 6 6 6 6 6 7 7 7 8 8 8 8 8 mumber of Asseseltes 12 16 ,

8 12 8 8 8 24 1 12 8 8 8 16 8 Pellet Density, 1 TD 94.0 94.0 94.0 M.0 94.0 94.0 94.0 94.0 9.40 94.0 94.0 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 7.5 7.5 7.5 7.5 7.5 7.5 7.5 7.5 Initial Enrichment (w/o U-235)

Upper 6 Inches UO 0.71 2.85 2.85 2.85 2.95 2.85 2.85 0.71 0.71 0.71 0.71 0.71 0.71 0.71 0.71 002-Gdzo3 --- --- --- --- --- --- --- --- --- --- --- --- --- --- ---

Central.132 Inches UO2 i.24t 2.85 2.85 2.85 2.85 2.85 2.85 3.34 3.34 3.34 3.73 3.73 002-Gd 023 ----

2.20 2.20 3.73 3.73 3.73 2.20 2.20 --- 2.37 2.37 --- 2.60I 2.60I 2.60I 2.603 Lower 6 Inches CD 002 30455TL 2.85 2.85 2.85 2.85 2.85 2.85 0.71 0.71 0.71 0.71 0.71 .0.71 0.71 0.71 u0 2-G4 023 --- --- --- 2.20 2.20 2.20 2.20 --- --- --- --- --- --- --- ---

Average 0.86 2.85 2.85 2.84 2.82 2.81 2.8) 3.12 3.09 3.02 3.48 3.45 3.43 3.41 3.41 Initial Gd23 0 , w/o --- --- --- 4 4 4 4 --- 4 4 4

--- 4 4 6 Batch Average Burnup at 80Cll MWD /MT 4,779 21,144 10.775 23,528 23,313 25,216 14.437 8.135 12,220 14.066 0 0 0 0 0 Feak Assembly Burnup at EOCll, MWD /MT 9,895 33,008 23,743 34,244 34,311 34,980 26.2 % 21,965 25,618 26,821 9,218 12,828 11,873 15,067 14,775 Fuel with 4 pins of 4 w/o gadolinta per assembly Fuel with 8 pins of 4 w/o gadolinta per assembly Fuel with 12 pins of 4 w/o gadolinta per assemely Fuel with 12 pins of 6 w/o gadolinta per essesely i

Lower 36 inches of central 132 inches contains 30455fL x PLSA Part Length Shield Assemely z 1  :*)

6 additional inches of natural urantum antal blanket fuel is uttllred in the top and bottom of gadolinta-bearing fuel pins.

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Xh-NF,85-103 9 Revision 2 R P 4 M L K J M G F E O C 8 A 1 PLSA PLSA PLSA l 455 m51 m47 I" ** *

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E12 E14 L12 04 P1:1 14 5 Pos P13 M24 437 27 N03 M34 N04 MOS N38 M25 P11 Poe L6 C11 011 G12 J12 411 mil E6 "

6 P37 413 435 M11 P26 M43 P17 M64 P25 M12 06 414 P38 PLSA

• 47 89

• PLSA 811 09 J7 M14 G7 M9~ 711 79 C7 7 246 P45 M13 m30 NO7 M52 M32 427 M31 M45tt 408 424 M14 P46 m56 FLSA E8 "" K9 F7 = 88 MB P8

• E7 F9 "" L8 Pt.5A 8 NGO 418 P28 M39 M38 P16 426 M53 E8 P18 M40 MJ7 P29 C0 452

• 49 87 85 07 J9

• PL$A M2 69 M7 P5 P7 C9 PLLA 9 m54 P48 M16 422 N06 M51 M30 m25 M29 M6 405 432 M15 P47 M8

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P40 m16 434 M10 P31 M42 P19 M41 P27 M09 m33 m15 .P39 14 = 86 *"

M12 E4 E2 J6 L2 L4 012 P6 14 11 P08 P9 M23 440 M06 402 M36 401 MOS 439 M22 P24 P07 M12 ~ F2 E3 J2

• 12 G2 G10 L3 E2 G8 M47 P13 M19 142 m29 M33 m21 241 M18 P12 M27 M9

• K11 J13 "" G13 F11

• 08 13 MZ8 P23 412 Moe P32 M03 411 P22 H50 14 "

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PL5A PLSA PL5A i 15

• 4 Pins of 4 w/o Gadolinta Per Assembly

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( 8 Ptns of 4 w/o Gadoltata per Assemely in Region 14 Fuel 12 Pins of 4 w/o Gadolinta Per Assemelf JFresh Gadoltata Seering Asswy in Region 14 Fuel ICycle 10 core location or region number PLSA part Lee 9*m Shield Assemely qaggeselyfabetCatton10

• • 12 Pins of 6 w/o Gadolinta Per Asseenly in Region 14 Fuel l it This assemely contains one (1) trwet J Itrcontue rod

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Ftpre 4.1 H. 8. Mottnson Unit 2 Cycle 11. Peference Loading Pattern for an EOC10 Esoasure of 11,150 med/MT e

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XN-NF-85 -103 Revision 2 H G F E D C B A 14,437 12,483 0 23,195 23,482 0 14,160 5,430 8 26,2 % 25,084 15,064 34,311 34,305 14,409 24,7.79 9,891 12 13 14*** 12 12 14**** 13 PLSA 12,493 24,683 23,190 8,133 12,172 20,935 0 4,451 9

13 12 12 13 13 12 14* PLSA 0 23,185 0 21,342 12,193 14,013 0 10 15,067 34,241 14,775 32,994 25,602 26,802 11,860 14*** 12 14**** 12 13 13 14**

23,156 8,138 21,355 12,086 10,771 0 0 34,282 21,965 33,008 25,478 23,720 13,304 9,106 gg 12 13 12 13 128 14*** 14' 23,418 12,181 12,202 10,779 0 25,749 34,254 25,279 25,618 23,743 12,586 31,151 12 12 13 13 128 14*** 12 .

0 20,943 14,020 0 24,510 ,

13 14,415 32,234 26,821 13,353 30,040 14**** 12 13 14*** 12 14,171 0 0 0 80C11 Exposure, mwd /MTV 14 24,792 12,828 11,873 9,128 EOC11 Exposure, mwd /MTU 13 14* 14** 14 Region ID 5,434 4,454 15 9,895 8,408 PLSA PLSA

• Fuel with 4 pins of 4 w/o gadolinia per assembly

• • Fuel with 8 pins of 4 w/o gadolinia per assembly

• • • Fuel with 12 pins of 4 w/o gadolinja per assembly

• • • • Fuel with 12 pins of 6 w/o gadolinia per assembly PLSA Part Length Shield Assembly Figure 4.2 H.8. Robinson Unit 2, 80C11 and E0C11 Quarter Core Exposure Distribution and Region ID for EOC10 =

11,150 mwd /MTU l

11 XN-NF-85-]03 Revision 2 5.0 MECHANICAL DESIGN The forty-eight (48) Batch XN-8 reload fuel assemblies are mechanically identical to the previous reload assemblies with the exception of the gadolinia-bearing fuel rods. Each fresh gadolinia-bearing fuel rod contains an enriched gadolinta-bearing region of 120 inches with 12 inches of natural uranium at each end of the rod. For non-gadolinia-bearing fuel rods, the 144 inch fuel column includes a six (6) inch column of natural 002 pellets at each end. The natural U02 pellets have a total dish volume of 0.75% as compared to 1.0% for the enriched pellets.

A description of the basic Exxon Nuclear supplied fuel design and design methods is contained in Reference 1. In addition, mechanical design analysis

• of the Batch XN-8 fuel and of the resident XN-6 and XN-7 fuel with current methodology is contained in Reference 2. The reference analysis .

encompasses the expected conditions to assure the relevant mechanical criteria will be met.

Utilization of the Partial Length Shield Assemblies (PLSAs) is described in Reference 3. The PLSAs are mechanically identical to the Reload XN-7 assemblies, with the exception of the bottom 42 inches of the fuel column, which contains stainless steel inserts.

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i 12 XN-NF-85-103 Revision 2 6.0 NUCLEAR CORE DESIGN The H. B. Robinson Unit 2, Cycle 11, Region 14 reload design has been developed in accordance with the following requirements:

1. The Cycle 11 reload shall contain forty-eight (48) new fuel assemblies.

l 2. The length of Cycle 11 shall be maximized.

l 3. The rated power for Cycle 11 shall be 2,300 MWt.

4. The length of Cycle 11 shall be determined based on a projected E0C10 exposure of 11,150 mwd /MTU.

5. Cycle 11 Operation is anticipated to be base loaded; however, the reload .

fuel shall be designed to accommodate load following operation between 50% and 100% of rated power while not precluding the current ramp and i step change bases as set forth in the FSAR.

6. In accordance with plant Technical Specifications, the control rod worth requirements shall be met.

7. The loading pattern shall be designed to produce acceptable power distributions. The design TF g, including uncertainties, shall be less than or equal to 2.32 at 2,300 MWt. The integrated peak to average pin power, Ft2H, including measurement uncertainties, shall be less than or l equal to 1.65 at 2,300 MWt.

8. The itiding pattern shall be designed to accomodate the PLSAs used to reduce the neutron fluence at the pressure vessel, l

The neutronic design methods utilized in the analyses are consistent with those described in References 4 through 8.

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13 XN-NF-85-103 Revision 2 6.1 PHYSICS CHARACTERISTICS The neutror.ic characteristics of the Cycle 11 core are compared to those of Cycle 10 in Table 6.1. The data presented in the table indicates the neutronic similarity between Cycles 10 and 11. The reactivity coefficients of the Cycle -

11 core are bounded by the coefficients used in the safety analysis. The safety analysis for Cycle 11 is based on physics characteristics repre-sentative of those expected for Cycle 10 lengths of -500 mwd /MTU and +500 mwd /MTU about the expected length of 11,150 mwd /MTU.

The boron letdown curve for Cycle 11 operation at 2,300 MWt is shown in Figure 6.1. As shown, the 80C11, no xenon, hot full power (HFP) critical boron concentration is predicted to be 1,132 ppm. At 100 mwd /MTV, equilibrium xenon, the critical baron concentration at HFP is 826 ppm. The Cycle 11 length is projected to be 11,530 mwd /MTU (332 EFPDs) with no boron at E0C at a power level of 2,300 MWt.

6.1.1 Power Distribution Considerations At a power level of 2,300 MWt at equilibrium xenon conditions, 100 mwd /MTV, the calculated XTG peak FAH is 1.52 including a 4% measurement uncertainty. At T

the same exposure, the peak F g is 2.08 including a 3% engineering factor, a 5% measurement uncertainty, K(Z) considerations (32), and an 8% allowance for

, operation with PDC-II for +3% target bands.

T The peak F Q in Cycle 11 is calculated to occur at a cycle exposure of 6,000 T

mwd /flTU. The calculated peak F g at this exposure, again including V(Z), K(Z)

. considerations and appropriate uncertainties, is 2.15 with + 3% target bands.

The predicted value of FaH at this exposure is 1.56 including the 4%

measurement uncertainty, r

14 XN-NF-85-103 Revision 2 The peak Fgg in Cycle 11 is calculated to occur at a cycle' exposure of 7,000 mwd /MTU. The predicted value of Fgi at this exposure is l'.58 including the 4%

measurement uncertainty. T The corresponding F g, including V(Z) and K(Z) considerations, and appropriate uncertainties, is 2.15.

The quarter-core radial power distributions are presented in Figures 6.2 through 6.4 for Cycle 11 exposures of 100 mwd /MTV, 7,000 mwd /MTU, and (EOC) 11,150 mwd /MTU, respectively.

6.1.2 Control Rod Reactivity Requirements Detailed calculations of shutdown margins for Cycle 11 are compared with Cycle 10 data in Table 6.2. A value of 1,770 pcm is used at E0C in the evaluation of the shutdown margin to be consistent with the Technical Specifications.

The Cycle 11 analysis indicates excess shutdown margins of 2,420 pcm at the BOC and 508 pcm at the EOC. The Cycle 10 analysis indicated excess shutdown margins for that cycle of 1,911 pcm at the B0C and 461 at the E0C. The Cycle 11 excess shutdown margins are seen to be similar to the Cycle 10 values.

The control rod groups and insertion limits for Cycle 11 will remain unchanged from Cycle 10. The control rod shutdown requirements in Table 6.2 allow for a HFP D Bank insertion equivalent to 600 pcm for BOC and 400 pcm for E0C, to bound the Cycle 11 control rod worths.

6.1.3 Isothermal Temperature Coefficient Considerations The Cycle 11 isothermal temperature coefficients are shown in Table 6.1 for HFP and HZP conditions at both B0C and E0C. At 80C11, following a nominal EOC10 shutdown exposure of 11,150 mwd /MTU, the HFP isothermal temperature coefficient is projected tc be -5.5 pcm/0F at a critical boron concentration of 1,132 ppm. The corresponding HZP critical boron concentration is 1,268 ppm with the isothermal temperature coefficient being -1.0 pcm/0F.

15 XN-NF-85-103 Revision 2 The Technical Specification for the moderator temperatur~e coefficient for Cycle 11 allows a +5.0 pcm/oF (Reference 15) at HZP conditions. With a calculated value of + 0.7 pcm/0F at HZP, the Technical Specification is expected to be met, with no control rod insertion anticipated for ascension to HFP conditions. Similarly, at HFP conditions, the calculated moderator temperature coefficient, shown in Table 6.1 for no xenon, is well below the requirement of 0 pcm/0F.

6.2 POWER DISTRIBUTION CONTROL PROCEDURES The control of the core power distribution is accomplished by following the procedures for " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase 11"(9,10,11). These procedures, denoted PDC-II, have been generically approved by the NRC for application to Westinghouse type PWRs.

~

6.3 ANALYTICAL METHODOLOGY The methods used in the Cycle 11 core analyses arc described in References 4 through 8. In summary, the reference neutronic design analysis of the reload core was performed using the XTGPWR(12) reactor simulator system. The fuel shuffling between cycles was accounted for in the calculations. The PDQ/ HARMONY (13,14) code package will be utilized to monitor the power distribution.

Calculated values of Fg, Fxy, and Fg g were studied with the twenty-four (24) axial nodal XTGPWR reactor model. The thermal-hydraulic feedback and axial exposure distribution effects of power shapes, rod worths, and cycle lifetime are explicitly included in the analysis.

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16 XN-NF-85-103 Revision 2 Table 6.1 H. B. Robinson Unit 2, Neutronics Characteristics of Cycle 11 Compared with Cycle 10 Data Cycle 10 Cycle 11 BOC EOC BOC EOC Critical Boron HFP,AR0,(ppm). 1,002 0 1,132 0 HZP, AR0, No Xenon (ppm) 1,134 -- 1,268 --

Moderator Temp. Coefficient HFP,(pcm/0F) -3.8 -27.6 -4 . 2 -28.9 HZP,(pcm/0F) +1.0 -21.0 +0.7 -22.6 Isothermal Temp. Coefficient HFP,(pcm/0F) -5.1 -29.1 -5.5 -30.4 HZP, (pcm/0F) -0.7 -22.8 -1.0 -24.4 U-238 Atoms Consumed Per Total Atoms Fissioned 0.50 0.76 0.50 0.76 Pressure Coefficient HFP,10-6 ap/ psia ---- +3.1 ----

+3.2 HZP,10-6 ao/ psia -0.10 ----

-0.08 ----

Doppler Coefficient (pcm/0F) -1.3 -1.5 -1.3 -1.5 Power Defect (Moderator +

Doppler),'cmp 1,532 2,030 1,574 2,032 Boron Worth, (ppm /103 pcm)

HFP -107 -96 -117 -102 HZP -105 -92 -114 -98 Prompt Neutron Lifetime (usec) 25.7 24.2 23.8 23.4 Delayed Neutron Fraction 0.0061 0.0053 0.0059 0.0052 Control Rod Worth of All Rods In Minus Most Reactive Rod,HZP,(pcm) 6,326 5,901 6,938 5,956 Excess Shutdown Margin, (pcm) 1,911 461 2,420 508

17 XN-NF-85-103 Revision 2 e

Table 6.2 H.B. Robinson Unit 2, Control Rod Shutdown Margin and Requirements for Cycle 11 (pcm)

B0C 10 EOC 10 BOC 11 EOC 11 Control Rod Worth ARI 7,626 7,201 S,338 7,356 N-1 6,326 5,901 6,938 5,956 (N-1)

• 0.9 5,693 5,311 6,244 5,360 Reactivity Insertion Power Defect (Moderator

+ Doppler) 1.532 2,030 1,574 2,032 Full Power D Bank Insertion 600 400 600 400

, Flux Redistribution 600 600 600 600 Void 50 50 50 50 Total Requirements 2,782 3,080 2,824 3,082 Shutdown Margin I (N-1)

• 0.9 - (Total Requirements) 2,911 2,231 3.420 2,278 Required Shutdown Margin 1,000 1,770 1,000 1,770 Excess Shutdown Margin 1,911 461 2,420 508 f

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XN-NF-85-103 Revision 2 H G F E D C B A 8 1.099 1.151 . 1.261 . 1.031 1.001 1.216' O.920 0.243 .

. . . PLSA .

9 . 1.151 . 0.919 0.991 1.339 1.260 1.015 1.109 0.215 .

. . . . PLSA 10 . 1.261 . 0.991 . 1.281 1.096 1.282 . 1.154 0.976 11 1.032 . 1.339 1.096 1.270 1.170 1.049 . 0.767 .

12 . 1.003 . 1.261 . 1.284 1.171 . 0.978 . 0.412 13 . 1.218 . 1.016 . 1.156 . 1.054 . 0.423 Assembly Relative Power l

Peak Assembly = 1.34 (E9)

Pe ak F N ",H = 1. 47 ( F10 )

14 . 0.920 . 1.110 0.978 . 0.770

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. PLSA . PLSA .

4 pins of 4 w/o gadolinia per assembly

• • 8 pins of 4 w/o gadolinia per assembly

• • • 12 pins of 4 w/o gadolinia per assembly

• • • • 12 pins of 6 w/o gadolinia per assembly PLSA Part Length Shielding Assembly w

Figure 6.2 H. 8. Robinson Unit 2 Cycle 11, Assembly Relative Power Distribution, 100 mwd /MTU, 2,300 MWt, 3-0 XTGPWR

XN-NF-85-103 Revision 2 H G F E D C B A 8 . 1.018 1.093 1.364 0.961 0.937 1.300- 0.951 0.284

. . . . . PLSA .

9 . 1.093 . 0.901 . 0.969 . 1.174 1.116 0.994 1.154 0.252 .

. . . . . . PLSA-10 . 1.364 0.970 1.315 . 1.002 . 1.149 1.131 1.093 11 . 0.961 . 1.174 . 1.002 . 1.152 . 1.143 . 1.246 . 0.835 .

12 0.938 . 1.116 1.150 . 1.145 1.180 0.501 .

13 . 1.301 . 0.994 . 1.132 . 1.250 0.520 Assembly Relative Power Pea < Assemoly = 1.26 (H10)

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• 1.52 (F13) 14 . 0.951 . 1.154 . 1.094 . 0.837 .

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. PLSA . PLSA .

4 pins of 4 w/o gadolinia per assembly

• • 8 pins of 4 w/o gadolinia per assembly

• • • 12 pins of 4 w/o gadolinia per assembly

• • • • 12 pins of 6 w/o gadolinia per assembly PLSA Part length Shielding Assembly 1

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Figure 6.3 H. 8. Robinson Unit 2 Cycle 11, Assembly Relative Power Distribution, 7,000 mwd /MTV, 2,300 MWt, 3-D XTGPWR i

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X N-NF 103 Revision 2 H G F E D s C B A 8 . 1.023 . 1.088 . 1.329 . 0.970 . 0.959 . 1.337' . 0.978 . 0.319 .

. . . . . . PLSA .

9 . 1.088 . 0.922 . 0.986 . 1.158 . 1.105 . 1.008 . 1.155 . 0.280 .

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10 . 1.329 . 0.906 . 1.339 . 1.009 1.124 1.106 . 1.084 .

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11 . 0.970 . 1.158 . 1.009 . 1.130 . 1.121 . 1.222 . 0.821 .

12 . 0.959 . 1.105 . l .J 23 . 1.122 . 1.185 . 0.533 .

13 . 1.337 . 1.007 . 1.106 . 1.225 . 0.545 . Assembly Relative Power

.........:.........:.........:.........:.........: Peak Assemly = 1.34 (F10)

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= 1.69 (E13)

PeakF}

15 . 0.319 . 0.279 .

. PLSA . PLSA .

• 4 pins of 4 w/o gadolinia per assembly

• • 8 pins of 4 w/o gadolinia per assembly

• • • 12 pins of 4 w/o gadolinia per assembly

• • • • 12 pins of 6 w/o gadolinia per assembly PLSA Part length Shielding Asaembly Figure 6.4 H. 8. Robinson Unit 2 Cycle 11, Assembly Relative Power Distribution, 11,530 mwd /MTV, 2,300 MWt, 3-D XTGPWR

22 XN-NF-85-103 Revision 2 7.0 THERMAL-HYDRAULIC DESIGN The Cycle 11 core consists of 157 ENC fuel assemblies. Fuel assemblies having natural uranium axial blankets and gadolinia bearing fuel co-reside in the core, as noted in Section 4.0. The hydraulic designs of the ENC fuel assemblies are essentially the same. Therefore, ne mixed core DNB penalty need be applied.(15) The results of thermal-hydraulic analyses at design hydraulic conditions are reported in Reference 16.

Local power distributions for gadolinia-bearing assemblies are less limiting with respect to MDNBR than the design local power distribution used in Reference 16 analyses. The impact of gadolinia fuel on calculated limiting assembly thermal margins is therefore negligible. Similarly, the use of axial blankets has no impact on thermal margins because these are evaluated at

. Technical Specification peaking limits. The use of axial blankets or ga'dolinia does not affect the hydraulic character of the fuel. Evaluation of the effect of axial blankets on the overpower AT and overtemperature aT reactor trip setpoint reset function, f(AI), is discussed in Section 8.0.

There are twelve part length shielding assemblies (PLSA) loaded in the core periphery. The PLSA assemblies will not approach limiting assembly condi-tions in their lifetime due to their low operating power. The hydraulic impact of these twelve assemblies on the limiting assembly is negligible.

The thermal-hydraulic performance of the Cycle 11 core under postulated transient and accident conditions is evaluated in Section 8.0 of this document.

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23 XN-NF-85-103 Revision 2 l

8.0 FSAR CHAPTER 15 EVENT ANALYSIS (Transient and Accident Analysis) l Extensive reanalysis of Chapter 15 events was performed to support increased relative power peaking limits in Cycle 10. That analysis is reported in References 17 and 18 and was performed in accordance with methodology described in Reference 19. The LOCA/ECCS reference analysis is described in Subsection 15.6.5. The applicability of that reference analysis to Cycle 11 is reviewed in this section. Rated operating conditions are summarized in Table 8.1.

A detailed review of the reload fuel design and cycle physics parameters was performed for Cycle 11 to identify any significant changes from the reference analysis. The reference Chapter 15 event analyses and the reference analysis Disposition of Events (17) were reviewed to determine the effectL I]f these changes. Changes in fuel design and cycle physics parlmeters are di! cussed below. An event by event review of the reference analysis is given be'ow. The reference analysis is found to bound Cycle 11 operation.

Significant fuel and core thermal-hydraulic design parameters employed in the reference Chapter 15 event analysis are listed in Table 8.2. The values are unchanged for Cycle 11. With respect to fuel design, the reference Chapter 15 event analysis is therefore applicable to Cycle 11.

The Cycle 11 core will contain two reloads of axially blanketed fuel.

Additionally, gadolinia rods in the fresh fuel loaded in Cycle 11 will have 12 inches rather than 6 inches of natural uranium axial blankets. These changes have no impact on the reference Chapter 15 event analyses because these events have been evaluated at Technical Specification peaking limits. Cycle 10 setpo~ int calculations verifying the adequacy of the overtemperature AT and overpower AT reactor trip reset function, f( AI), have been repeated using axial power profiles prototypic of Cycle 11 and future cycles. These calculations confirm the adequacy of the existing f(al) function.

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F 24 XN-NF-85-103 Revision 2 Cycle 11 bounding physics parameters are compared in Tables 8.3 and 8.4 with the bounding values of those parrieters employed in the reference analyses.

'Jith the exception of the minimum BOL values of the ef fective delayed neutron

, fraction (Beff) and the BOC value of the prompt neutron lifetime (A), the range l

of reference analysis values bounds Cycle 11 values. The impact of slightly smaller BOC Seff and A values on the reference analysis were therefore considered in this Cycle 11 review.

The DNB local peaking and rod bow requirerrents(20) were reviewed for Cycle 11.

No bow penalties need be applied for assembly burnups equal to or less than the design burnup of 44,000 mwd /MTV.

The Chapter 15 events are reviewed on an i nd ividual basis below. Event numbering follows that employed in Reference 17 Reanalysis performed for Cycle 11 employed methods detailed in Reference 19.

25 XN-NF-85-103 Revision 2 15.1.1 Decrease in Feedwater Temperature The event was evaluated in Reference 17 to be bounded by the Increase in Steam Flow event (15.1.3). The basis for evaluation was a comparison of rate and magnitude of thermal load increase -esulting from the events. The evaluation is independent of cycle physics parameters and fuel design. The disposition of this event for Cycle 11 is thus unchanged from the reference cycle analysis.

15.?.2 Increase in Feedwater Flow Two events are considered in Reference 17: 1) at full power, one steam generator feedwater regulating valve opens to full capacity; 2) at startup, while operating on the feedwater bypass system, a feedwater flow regulating' valve opens to full capacity.

Subevent I was determined to be bounded by the Increase in Steam Flow event (15.1.3), based on a comparison of the magnitude of the thermal load increase resulting from the two events. The evaluation is independent of cycle pnysics parameters and fuel design. The disposition for subevent 1 is thus unchanged from the reference cycle analysis.

Subevent 2 was determined to be bounded by the Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition event (15.4.1). The basis for the determination was a comparison of reactivity insertion rates for the two events, computed for event 15.1.2 as the product of a maximum primary system cooldown rate and a most negative E0C moderator temperature coefficient of -35 pcm/0F. The rate of cooldown is independent of cycle physics parameters and fuel design. The E0C MTC for Cycle 11 will j remain more positive than -35 pcm/0F. The reactivity insertion rate employed in the Reference 17 evaluation of subevent 2 is therefore bour. ding of Cycle 11, and the disposition of the subevent is unchanged from the reference l

analysis.

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26 XN-NF-85-103 Revision 2 15.1.3 Increase in Steam Flow A 10% Increase in Steam Flow event was analyzed for Cycle 10.(18) The event was reanalyzed for Cycle 11 to determine the impact of the minor changes in cycle physics parameters (Table 8.3). The MDNBR of 1.30 reported in Table 2.2 of Reference 18 is bounding of Cycle 11.

15.1.4 Inadvertent Opening of Steam Generator Relief or Safety Valve Two events are considered in Reference 17: 1) opening of a PORV or safety valve at power; and 2) opening of a PORV or safety valve after trip.

Subevent I was determined to be bounded by the Increase in Steam Flow event (15.1.3) based on a cunparison of the magnitude of the thermal load increase resulting for the two events. The evaluation is independent of cycle physics parameters and fuel design. The disposition for subevent 1 is thus unchanged from the reference cycle analysis.

Subevent 2 was determined to be bounded by the Main Steam Line Break event (15.1.5) based on a comparison of the magnitude of the steam flow for the two events. The event 15.1.5 analyses (21,22) demonstrate results in accordance with Condition II acceptance criteria.

15.1.5 Steam System Piping Failures Inside and Outside Containment The event was analyzed in Reference 21 and in Reference 22 (with removal of or reduction of boron concentration in the boron injection t en k ) . Event results are chiefly dependent on break steam flow, moderator and Doppler feedbacks, shutdown margin, and core power distribution. Moderator feedback and shutdown margin were set to Technical Specification limits, and are therefore treated in a bounding manner in the reference analysis. The reference analyses enployed a steam finw rate approximately 20% larger than that predicted by - the conservative Moody critical flow model (saturated s te am ), leading to a significantly conservative prediction of moderator

27 XN-NF-85-103 Revision 2 cooldown and thus an exaggerated peak power level. Power peaking is cycle specific; the essential similarity of the Cycle 10 and Cycle 11 core would indicate only small changes in power peaking between the cycles. Substantial conservatism in the predicted steam flow, and thus peak core power, in the reference analysis ensures that cycle specific variations in power peaking are bounded. The reference analysis is thus applicable to Cycle 11.

15.2.1 Steam Pressure Regulator Failure The plant has no main steam line pressure regulators. The event is thus not applicable to H.B. Robinson Unit 2.

15.2.2 Loss of External Load Two cases were analyzed in the reference analysis (18); one to evaluate the challenge to the MDNSR SAFDL, the other to assess the challenge to vessel pressurization limits. Botn cases were reanalyzed to determine the effect of minor changes in cycle physics carameters (Table 8.3). The MDNBR of 1.19 and the peak pressurizer pressure of 2661 psia reported in Table 2.2 of Reference 18 bound Cycle 11.

15.2.3 Turbine Trip; 15.2.4 Loss of Condenser Vacuum; 15.2.5 Closure of Main Steam Isolation Valve These events are similar to the Loss of External Load event (15.2.2), which was analyzed (18) in a mar.ner to bound the outcome of events 15.2. 3,15.2.4 and 15.2.5 as well. The disposition is independent of cycle-to-cycle variations and therefore applicable to Cycle 11.

15.2.6 Loss of Nonemergency A.C. Power to the Station Auxiliaries The significant event considered in Reference 17 results in turbine trip with consequent coastdown of primary coolant pumps and trip of main feedwater pumps.

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28 XN-NF-85-103 Revision 2 The short term phase of the event, during which a challenge to the ONB SAFDL may develop, is determined in the reference to be no worse than that occurring as a result of the Loss of Forced Reactor Coolant Flow (15.3.1). The basis for the disposition is that direct reactor trip will occur as a result of the event initiator in event 15.2.6. This basis is independent of cycle, and the disposition of the short term event for Cycle 11 is unchanged from the reference.

In the longer term, during which a challenge to post-trip decay heat removal capability may develop, the event is determined to be bounded by the Loss of Normal Feedwater event (15.2.7). The basis for the event disposition is again the occurrence of an anticipatory trip for event 15.7.6 and therefore also independent of cycle. The disposition of the longer term phase of event 15.2.6 is thus also unchanged from the reference.

1512.7 Loss of Normal Feedwater The event was reanalyzed in Reference 18. Two cases were treated: one with and one without forced primary coolant flow. Controlling variables are reactor decay heat and auxiliary feedwater flow capacity. The analysis is independent of cycle physics parameters and fuel design, and thus remains applicable to Cycle 11.

15.2.8 Feedwater System Pipe Breaks i

The event was evaluated in Reference 17 to be bounded by the main steamline break (15.1.5). The basis for this disposition is a comparison of possible break flow areas. The evaluation is independent of cycle physics parameters and fuel design. The disposition for Cycle 11 is thus unchanged from the reference cycle.

l

29 XN-NF-85-103 Revision 2 15.3.1 Loss of Forced Reactor Coolant Flow Coastdown of three primary coolant pumps is analyzed in Reference 18. Two cases are considered: one to evaluate MDNBR, and a second to assess the challenge to sessel pressurization limits. The reference analysis shows the pressurization case to be clearly bounded by the Loss of External Load (15.2.2), a result not dependent on cycle specific parameters. The MDNBR case was reanalyzed to assess the impact of minor changes in cycle physics parameters. The MDNBR of 1.25 reported in Reference 18 for this event is bounding of Cycle 11.

15.3.2 Flow Controller Malfunction The plant has no primary coolant flow controllers. The event is thus not applicable to H.B. Robinson Unit 2. ~

15.3.3 Reactor Coolant Pump Rotor Seizure The event was reanalyzed for Cycle 10 in Reference 13. Two cases were considered: one to evaluate MDNBR, and a second to assess the challenge to vessel pressurization limits. The peak pressure reached in the pressur-ization case is well below that achieved in the Loss of External Load (15.2.2) a result not dependent on cycle specific parameters. The MDNBR case was reevaluated for Cycle 11 to assess the effects of minor changes in cycle physics parameters (Table 8.3). The MDNBR of 0.90 reported in Reference 18 for this event is bounding of Cycle 11.

15.3.4 Reactor Coolant Pump Shaft Break l

In Reference 17, this event is determined to be bounded by the Reactor Coolant Pump Rotor Seizure event (15.3.3). The basis of the evaluation is a comparison of flow decay rates. The evaluation is independent of cycle physics parameters and fuel design. The disposition of this event for Cycle 11 is thus unchanged from the reference analysis.

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30 XN-NF-85-103 Revision 2 15.4.1 Ungontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition The event was reanalyzed for Cycle 10(18) for two pump operation and a bounding Control Rod Assembly bank worth. The event was reanalyzed for Cycle 11 to determine the effect of slightly changed cycle physics parameters (Table 8.3). The MDNBR of 1.26 reported in Reference 18 for this event remains bounding for Cycle 11.

15.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power The event was analyzed for Cycle 10 at rated, mid and low power initial conditions, for BOC and E0C conditions, at a spectrum of possible reactivity insertion rates. The limiting case for Cycle 10 was reanalyzed for Cycle 11 to assess the effects of slightly changed cycle physics parameters (Table 8.3). The MDNBR of 1.19 reported for this event in Reference 18 is bounding of Cycle 11.

Limiting low power cases may initiate with significantly higher radial power peaking than can actually exist at the core conditions reached at reactor trip. The transient reduction in radial power peaking which occurs in these events due to Doppler and moderator feedback is sufficient to offset the initially elevated radial peaking. Consideration of radial power peaking f actors in excess of the rated power design limit is thus unnecessary for Cycle 11.

1 l

l 15.4.3 Control Rod Misoperation ,

This event includes the single RCCA withdrawal, the static RCCA misalignment, and the Dropped RCCA or RCCA Bank events. The events were reanalyzedt for Cycle 10 in Reference 18 and are addressed individually for Cycle 11 below.

tSafety issues raised in References 29 and 30 are addressed in the reference analysis.

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31 XN-NF-85-103 Revision 2 l

l Bounding radial peaking augmentation f actors for these events are the same for Cycle 11 as for Cycle 10 (Table 8.4).

The single RCCA withdrawal event was reanalyzed in Reference 18 by combining a radial peaking augmentation factor calculated for the ~ event with core thermal-hydraulic boundary conditions calculated for the most limiting case of event 15.4.2. Since the radial peaking augr:ntation factor and the limiting case of event 15.4.2 are unchanged for Cycle 11 relative to the reference analysis, no Cycle 11 specific analysis is required.

The static rod misalignment event results are cycle specific only through the radial peaking augmentation f actor. The bounding value of 1.27 calculated for the reference cycle bounds Cycle 11 also. Thus, the Cycle 10 analysis is bounding of Cycle 11.

The radial peaking augmentation f actors for the dropped RCCA/ dropped RCCA' bank event employed in the Cycle 10 analyses bound Cycle 11 (Table 8.4). The effect of slightly changed cycle physics parameters (Table 8.3) were evaluated by reanalysis of limiting cases for Cycle 11. The results reported for this event in the reference analysis are bounding of Cycle 11.

15.4.4 Startup of an Inactive Loop at an Incorrect Temperature Power operation with less than three loops in service is prohibited by Technical Specifications. Analysis of this event for H.B. Robinson is thus unnecessary.

15.4.5 Flow Controller Malfunction The H.B. Robinson Unit 2 plant has no primary loop isolation valves nor means to control primary flow. Therefore, this event is not applicable to H.B.

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! Robinson Unit 2.

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32 XN-NF-85-103 Revisian 2 15.4.6 Chemical and Volume Control System Malfunction that Results in a Uecrease in Baron Concentration in the Reactor Coolant The event result as analyzed for Cycle 10 is dependent on dilution flow rate, mixed reactor coolant volume, the BOC boron worth coefficient, and the B0C critical baron concentration. Dilution flow rate and reactor coolant volume are fixed system parameters independent of cycle. The event was reanalyzed in all modes of operation for Cycle 11,to assess the impact of increased BOC critical bornn concentration and a slightly decreased boron worth coef-ficient. The results of the Cycle 11 analysis are reported in Table 8.5. The maximum reactivity addition due to this event is suf ficiently slow to allow the operator to determine the cause and to take corrective action before shutdown margin is lost.

15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an '

Improper Position The event was reanalyzed for Cycle 10'and reported in Reference 18.

Reanalysis of this event for Cycle 11 indicates that a maximum undetected FAH l of 1.77 may occur. Since this value is less than the value of 1.94 treated in the reference analysis, the reference analysis result is bounding of Cycle 11.

15.4.8 Spectrum of Rod Ejection Accidents (PWR)

A Control Rod Ejection Accident is defined as the mechanical f ailure of a control rod mechanism pressure housing, resulting in the ejection of a Rod Cluster Control Assembly (RCCA) and drive shaf t. The consequence of this mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

The rod ejection accident has been evaluated with tne procedures developed in the ENC Generic Rod Ejection Analysis, Reference 23. The ejected rod worths and hot pellet peaking f actors were calculated using the XTGPWR code. lio credit was taken for the power flattening effects of Doppler or moderator

33 XN-NF-85-103 Revision 2 feedback in the calculation of ejected rod worths or result Ant post-transient peaking factors. The pellet energy deposition resulting'from an ejected rod was conservatively evaluated explicitly for BOC, M0C and E0C conditions at HFP and B0C and E0C at HZP conditions. The HFP pellet energy deposition was calculated to t" 152 cal /gm at BOC,147 cal /gm at MOC, and 154 cal /gm at EOC.

The HZP pellet energy deposition was calculated to be less than 40 cal /gm for both B0C and E0C conditions. The rod ejection accident was found to result in an energy deposition of less than the 280 cal /gm limit as stated in Regulatory Guide 1.77. The significant parameters for the analyses, along with the results, are summarized in Tables 8.6 and 8.7.

15.4.9 Spectrum of Rod Ejection Accidents (BWR)

This event is not applicable to pressurized water reactors.

15.5.1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory This event is discussed in greater detail in Reference 17. The disposition of this event is unchanged from that for Cycle 10.

15.5.2 Inadvertent Operation of Chemical and Volume Control System that Increases Reactor Coolant Inventory The consequences of unplanned additions to inventory and effect of reactivity additions due to dilution during refueling and startup are treated in the analysis of event 15.4.6. The consequences of dilutions at power are bounded by the evaluation of event 15.4.2, Uncontrolled RCCA Bank Withdrawal at Power.

l l

l The consequences of volumetric addition and effect on the pressure boundary during all operational modes have been earlier addressed in Section 15.5 of the H.B. Robinson Unit 2 FSAR.(24) The evaluation is determined by l

pressurizer PORV capacity, and is thus independent of cycle.

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34 XN-NF-85-103 Revision 2 15.6.1 Inadvertent Opening of a Pressurizer Pressure Re' lief Valve A bounding. analysis of the event for Cycle 10 is given in Reference 17. The input parameters to the analysis were reviewed and found to be non-cycle specific. The analysis is thus applicable to Cycle 11.

15.6.2 Loss of Reactor Coolant from Rupture of Small Pipes or from Cracks in large Pipes which Actuate the Emergency Core Cooling System An analysis was performed in 1975 with ENC's approved small break model.

Substantial margin to acceptance criteria was demonstrated in that anal-ysis.(24I The analysis was performed for breaks ranging in size from a 6 inch diameter break to a complete double ended guillotine cold leg break. That analysis demonstrated that the large break bounded the small breaks with substantial margin. There have been no changes which would change the relative aspects of small and large breaks. Therefore, the event as analyzed in 15.6.5 (large break LOCA) bound the results of this event.

15.6.3 Radiological Consequences of Steam Generator Tube Failure l The event was reanalyzed for Cycle 10 with results reported in Reference 18.

Radioactive releases depend on the Technical Specification limits on primary coolant activity and the amount of coolant released. Cycle 11 primary coolant activity limits are unchanged from Cycle 10. Primary coolant release is dependent on primary coolant thermodynamic state and tube rupture area. No changes have occurred between Cycles 10 and 11 which would change these controlling parameters. The reference analysis is thus applicable for Cycle 11.

15.6.4 Radiological Consequences of Main Steam Line Failure Outside Containment This event is not applicable to pressurized water reactors.

35 XN-NF-85-103 Revision 2 15.6.5 Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary The event was analyzed to support Cycle 10 and reported in References 25, 26 and 27. The result of that analysis and the reactor operating limits supported by it were revised in Reference 28. As modified by Reference 28, these analyses are applicable to Cycle 11. Analyses to support Cycle 11 operation with a 2.32 FQ are reported in Reference 32.

15.7.1, 15.7.2 (Events deleted from the Standard Review Plan.)

15.'.3 Postulated Radioactive Releases due to Liquid-Containing Tank Failure This event has been reviewed and results documen,ted in the H.B. Robinson Unit 2 FSAR. The results of this evaluation are unchanged by the planned licensing actions and therefore bound Cycle 11.

15.7.4 Radiological Consequences of Fuel Handling Accidents The event is characterized by the release of fission products from a single limiting fuel assembly during handling in the spent fuel pool. The event was analyzed to support Cycle 10 and reported in Reference 31. The reference analysis was reviewed for applicability to Cycle 11. Fission product inventories are determined by exposure and power level. The reference analysis employed a maximum design burnup of 44,000 mwd /MTV, a power level of 2300 MWt, and a radial peaking factor of 1.65. These parameters are unchanged; thus, the reference analysis is bounding of Cycle 11.

15.7.5 Spent Fuel Cask Drop Accidents l The results of this accident are unchanged frcm tho ;c documented in the H.B.

Robinson Unit 2 FSAR. The results of the analysis therefore bound Cycle 11.

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36 XN-NF-85-103 Revision 2 Table 8.1 Plant Rated Operating Conditions Core Thermal Power 2300 MWt Core Coolant Inlet Temperature 546.20F Core Coolant Average Temperature 575.40F Vessel Coolant Flow

• 100.3

• 106 lb/hr l

Active Core Flow

• 95.8

• 106 lb/hr Steam Generator Pressure (Dome) 800 psia Steam Flow 10.1

• 106 lb/hr Feedwater Temperature 441.50F Coolant flow reflects 6% steam generator tube plugging for rebuilt steam generators and 3% uncertainty.

37 XN-NF-85-103 Revision 2 Table 8.2 Core and Fuel Design Parameters Used in Chapter 15 Event Analysis Number of fuel assemblies of all types 157 in core Number of part length shielding 12 fuel assemblies Fuel assembly pitch 8.456 in.

Fuel assembly design type 15x15 Fuel rods per assembly 204 Guide tubes per assembly '20 Instrument tubes per. assembly 1 -

Fuel rod pitch .563 in.

Fuel rod 0.D. 424 in.

Guide and instrument tube 0.D. .544 in.

(above dast. pot)

Active fuel length 144 in.

Fuel rod length 152 in.

Number of spacers 7 Maximum spacer span length 26.2 in.

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38 XN-NF-85-103 Revision 2 Table 8.3 Bounding Physics Parameters for Cycles 10 and 11 Item Cycle 10 Cycle 11 BGC EOC BOC EOC

1. Moderator Temperature +5.0 -35.0 + 5 . 0' -35.0 Coefficient,10-5 ao/oF

2. Moderator Pressure .56 +3.9 .56 +3.9 Coefficient ,10-0 6 0/ psia

3. Doppler Coefficient, -1.0 -1.7 -1.0 -1.7 10-b AP/0F ,

4. Delayed Neutron Fraction Maximum .0070 .0055 .0070 .0055 Minimum .0060 .0045 .0059 .0045

5. Rod Worth, (N-1)*.9, .036 .036 .036 .036

6. Effective Neutron Lifetime, 24 22 23 22 usec

7. U-238 Atoms Consumed per .50 .76 .50 .76 Total Atoms Fissioned

39 XN-NF-85-103 Revision 2 Table 8.4 Bounding Power Peaking Factors and Agtsnentation Factors for Cycles 10 and 11 Cycle 10 Cycle 11 Nuclear Enthalpy Rise Factor, FaH, 1.65 1.65 (2100% rated power)

Axial Peaking Factor 1.65 1.65 Total Heat Flux Peaking Factor

• 2.32 2.32 Fraction of Power Deposited in Fuel 0.974 0.974 Fo H Augmentation Used in Single RCCA 1.27 1.27 Withdrawal Event FaH Augmentation Used in Static 1.12 1.12 Misalignment Event FaH Augmentation Used 'in Rod / Bank Drop Analysis Large Bank Worth <1.30 <1.30 Small Rod Worth <1.10 <1.10

• Value supported by non-LOCA/ECCS event analysis.

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i Table 8.5 11. B. Robinson Unit 2 Cycle 11 Results of the Analyses of CVCS Malfunction Time to Critical Boron loss of Shutdown Reactor Conditions Concentration (ppm) Margin (min.)

Refueling 1306 41.5 Cold Shutdown Case 1 (3% ap Shutdown) 1331 17.7 (1 pump) 1331 18.6 Case 2 1331 15.6 i l

liot Shutdown 1338 15.3 1

.. 1 Startup 1268 15.2 Power Operation Bounded by analysis in Sections 15.4.1 and 15.4.2 a

=L "S

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l Table 8.6 H.B. Robinson Unit 2. Cycie 11 Ejected Rod Analysis, HFP 1

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BOC MOC E OC l Contribution (a) to ContributionI8} to Contribution (8) to l Energy Deposition. Energy Deposition, Energy Distribution l Value (cal /gm) Value [calfgmj Value (c al/gm)

A. Initial Fuel Enthalpy (cal /gm) 72.2 ---- 75.5 ----

81.8 ----

B. Generic Initial Fuel Enthalpy (cal /gm) 40.8 ---- 40.8 ---- 40.8 ----

C. Delta Initial fuel Enthalpj (c al/gm) 31.4 31.4 34./ 34.7 41.0 41.0 D. Vanimum Coatrol Rod

!!2 Worth (pcm) 100 121 106 95 125 E. Dcppler Coefficient (pcm/of) -1,3(e) 0.99(b) 1,4(e) 0.92(b) _],5(e) 0,gg(b)

F. Delayed Neutron Fraction, 0.0059 1.01(b) 0.0056 1.02(b) 0,0057 1,04(b)

G. Powe r Peak ing Factor 2.4 ---- 2.7 ---- 3.4 ----

H. Power Peaking f actor Used(c) 4.0 ----

4.0 ---- 5.0 -----

Total 152.0(d} 147.0(d) 154(d)

(4) The contritautiar. to the total pellet envr gy dsposit ion is a f unc t ion of init ial f uel enthalpy, man i%m control rod wort h, Doppler coef f ic ient , ar.d delayed neutron f r d( t ion. The energy depasit icn contribution salues and f actors are derived f rom data calculated in the " Generic Analysis of the Control Rod Eject ion Transient. " doc umen t .

( t, ) These salues are multiplication f actors applied to (C

• D). g x~

(c) The er.ergy deposit icn due to L.as imum c ontrol rm1 worth is a f unct ion of the power peak inq f ac tor. Qh (d) Total pellet energy depasit ion (cal /gm) c alculated by the equat ion Total (csl/7) = (C+D)(E )(F)

[

g ,CI3 (e) For this Deppler coef f ic ient, conservat ive v alues of -1.1, -1,3, and -1.4 were assumed at BOC, MUC, "b and f0f, respectively. NO

a Table 8.7 H.B. Robinson Unit 2, Cycle 11 Ejected Rod Analysis, HIP BOC IOC ContributionID to Contribut ion 91 to Energy Deposition, Energy Deposition, Value (c al/gml. Value (cal /gm)

A. Initial f uel Enthalpy (cal /gm) 16.7 - - . - 16.7 ----

B. Generic initial fuel Enthalpy (cal /gm) 16.7 ---- 16.7 ----

C. Delta Initial Fuel Enthalpy (cal /gm) 0.0 0.0 0.0 0.0 D. Maximum Control Rod Worth (pcm) (no 33 600 33 g E. Doppler Coet f ic ient (pcm/Of) -1.7(e) 1.04(b) -1.8(e) o,73(b)

F. Delayed Neutron Fraction. 0.0059 1.02(b) 0.0052 1.16(b)

G. Power Peak ing Factor 6.5 ----

9.1 ----

H. Power Pesking f actor used(c) 13.0 ---- 13.0 ----

TOTAL 35.0(d) ?8.0(d)

( a) The contribution to the total pellet energy deposition is a f unc t ion of initial fuel enthalpy, i manimum cortrol rod worth, Dopolrr ccefficient, and delayed neutron fraction. The energy l dcposition contribut fon values and f actors are derived f rom data calculated in the Wneric l Analysis of the Control Rod Ejection Transient.

• doc unem t .

(b) These values are multiplic aticn f acters applied to (C

• D). x (c) The energy depos it ion due to man imum c ont r ol rod wor t h is a f unc t ion of the power peck ing f ac tor , y7 (d) Total pellet energy depositten (cal /gm) calculated by the equation

• 5. 5 Total (cal /gm) = (C+D)(E)(F) g, &

(e) For this Doppler coef ficient, conser ative values of -1.0 and -1.4 were assumed at BOC $7 and EOC, respectively. y 5 w

43 XN-NF-85-103 Revision 2

9.0 REFERENCES

1. XN-75-39, " Generic Fuel Design for 15x15 Reload Assemblies for Westinghouse Plants," Exxon Nuclear Company, September 1975.

2. XN-NF-83-55, " Mechanical Design Report Supplement for H.B. Robinson Extended Burnup Fuel Assemblies;" Exxon Nuclear Company, August 1983.

3. XN-NF-83-71, " Mechanical Design Report Supplement for H.B. Robinson Part length Shielding Assemblies," Exxon Nuclear Company, September 1983.

4. XN-75-27(A), Exxon Nuclear Neutronics Design Methods for Pres-surized Water Reactors," Exxon Nuclear Company, June 1975.

5. XN-75-27, Supplement 1, September 1976.

6. XN-75-27, Supplement 2, December 1977.

7. XN-75-27, Supplen.ent 3, November 1980.

8. -XN-75-27, Supplement 4, To Be Issued.

9. XN-M 57( A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors - Phase II," Exxon Nuclear Company, January 1978. ,

10. XN-NF-77-57(A), Supplement 1, June 1979.

11. XN-NF-77-57(A), Supplement 2 Septe,mber 1981.

12. XN-CC-28, Revision 5, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)," Exxon Nuclear Company, July 1979.

13. WAPD-TM-678, "PDQ7 Reference Manual", Westinghouse Electric Cor-poration, January 1967.

14. WAPD-TM-478, " HARMONY: System for Nuclear Reactor Deplet ierr Computation", Westinghouse Electric Corporation, January 1965.

15. XN-NF-82-21(A), " Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, September 1983.

16. XN-75-38, "H.B. Robinson Unit 2, Cycle 4 Reload Fuel Licensing Data Submittal," Exxon Nuclear Company, August 1975.

44 XN-NF-85-103 Revision 2

17. XN-NF-83-72, Rev. 2, Supp.1, "H.B. Robinson Unit 2 Cycle 10 Safety Analysis Report; Disposition of Ch;pter 15 Events," Exxon Nuclear Company, July 1984.

18. XN-NF-84-74, Rev. 1, " Plant Transig t Analysis for H.B. Robinson Unit 2 at 2300 MWt with Increased F y" (To be issued).

19. XN-NF-84-73(P), " Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Exxon Nuclear Company, August 1984.

20. XN-NF-75-32(A), Supplements 1,2,3 & 4, " Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, October 1983.

21. XN-NF-84-74(P), Supp. 3, " Plant Transient Analysis for H.S. Rob-inson Unit 2 at 2300 MWt with Increased Fh H, Supplement 3:

Confirmatory Analysis of the Steamline Break Event," Exxon Nuclear Company, January 1985.

. 22. XN-NF-85-17(P), " Analysis of the Steamline Break Event with Baron Injection Tank Removal or Oilutiorf to Zero Concentration Boric Acid for H.B. Robinson Unit 2," Exxon Nuclear Company, May 1985.

23. XN-NF-78-44, "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, February 1979.

24 Final Safety Analysis Report (Updated), H.B. Robinson Steam Elec-tric Plant Unit No. 2.

25. XN-NF-84-72, "H.B. Robinson Unit 2 Large Break LOCA/ECC5 Analysis with Increased Enthalpy Rise Factor," Exxon Nuclear Company, July 1984.

26. XN-NF-84-72, Supp. 1, "H.B. Robinson Unit 2 Large Break LOCA/ECCS Analysis with Increased Enthalpy Rise Factor: Break Spectrum Analysis," Exxon Nuclear Company, August 1984

27. XN-NF-84-72, Supp. 2, "H.B. Robinson Unit 2 Large Break LOCA/ECCS Analysis with Increased Enthalpy Rise Factor: K(Z) Curve," Exxon

. Nuclear Company, August 1984.

28. Letter, A.B. Cutter (CP&L) to S.A. Varga (USNRC), "H.B. Robinson Steam Electric Plant, Unit No. 2 - Operating Plans Under Revised LOCA Analysis," dated October 15, 1985; submitted under. 0ccket No.

50-261/ License No. DPR-23.

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45 XN-NF-85-103 Revision 2

29. Letter, T.G. Satryan (Westinghouse) to W.L. Stewart (Virginia Electric & Power Company),

Subject:

Unreviewed Safety Issue on Dropped Rod on Turbine Runback Plants (VPA-E3-613), dated Aug-ust 24, 1983.

30. " Flux Rate Trip Setpoint," Number NSID-TB-85-13, Westinghouse Technical Bulletin, May 28, 1985.

31. XN-NF-84-6S(P), "H.B. Robinson Unit 2 Radiological Assessment of Postulated Accidents," Exxon Nuclear Company, June 1984. .

32. Letter, S.R. Zimmerman (CP&L) to S. A. Varga (USNRC),

Subject:

"H.B.

Robinson Steam Electric Plant Unit No. 2 - Westinghouse K(Z)

Analysis," dated November 8, 1985; submitted under Docket No. 50-261/ License No. OPR-23.

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XN-NF-85-103 Revision 2 Issue Date: 1/10/86 H. B. Robinson Unit 2, Cycle 11 Safety Analysis Report DISTRIBUTION FT Adams GJ Busselman LJ Federico RC Gottula TJ Helbling JS Holm JW Hulsman WV Kayser CE Leach TR Lindquist JN Morgan TW Patten FB Skogen IZ Stone MS Stricker GN Ward HE Williamson Document Control (5)

CP&L (20)/TJ Helbling l

_