ML20214Q789

Cycle 12,SAR

ML20214Q789
Person / Time
Site:	Robinson
Issue date:	01/31/1987
From:	Adams F, Skogen F, Stone I CAROLINA POWER & LIGHT CO., SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML14188B328	List: ML20214Q789
References
XN-NF-86-149, NUDOCS 8706050189
Download: ML20214Q789 (46)

Text

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XN-NF-86-149 1

' i l .3. ROBINSON UNF 2, CYC_E 12 SAFETY A\A_YSIS RE3 ORT JA\UARY 1987 9 C-l_A\D, WA 99352 EXXON NUCLEAR COMPANY, INC.

l l$A'A86SES$88llgi

XN-NF-86-149 Issue Date 1/7/87

'H.B. ROBINSON UNIT 2, CYCLE 12 SAFETY ANALYSIS REPORT Written by: ,::::- f'- . [/sth

1. Z. Stong, Nuclear Engineer 4 'TANE7 PWR Neutronics Written by: .

!' d - /d F. T. Adar(s, Team Leader PWR Safety Analysis Approved by: M F. B. Skogeny Manager

'[6 / S 7 PWR Neutronics Approved by: f/4 /f 7 Jfj$. Hofm, Manager / '

PWR Safety An sis Approved oy: ff/I29{ / //6/<P~/

T. W. Patten, Manager Neutroni s & Fuel anagement A .'

Approved by: c'.?.4 ~ m .c h % //4// 7 H. E. Williamson, Manager l

Licensing & Safety Engineering Approved by: s //7[r>

ti. J./Busselman, Manager

' Fuel Design l

Approved by: ~7N./// c_ (f44 / /y / f7 G. L. Ritter, Manager Fuel Engineering & Technical Services Concurred by: v.f/M(/'>r h I v/ d '4<t f/f/ge J. N. Morgan, Manager Proposals & Customer Services Engineering 7

,,g ERON \UC_EA9 COV DA\Y INC.

t l

CUSTOMER OlSC1. AIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Exxon Nuclear Company's warranties and representations concermng the sabject metter of this document are those set forth in the Ayeoment between Exxon Nucteer Company, Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expreesty provided in sum Agreement, no:ther Exxon Nuclear Company, Inc. nor any person acting on its behalf makes any werranty or representation, expressed or implied, with roepect to the accuracy, completeness, or usefulness of the mformecon contamed in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privstoft* owned rights; or assumes any liabilitwo with respect to the use of arry informenon, apparatus, method or process discioeed in this document.

The information contained herein is for the soie use of Customer.

In order to avoid impearment of rights of Exacq nuleer Company, Inc.

in patants or invencons which may be included in tne information contained in this document, the recipient, by its acceptance of this document agrees not to publish or make public use (in the potent use of the term) of sud

, informenon unni so authorized in wrrting by Exxon Nuclear Company, Inc.

or unal after six (6) months following termination or expiration of the aforeseed Ayeoment and any extensson thereof, unions otherwise expressly provided in the Agreement. No rights or licenses in or to any potents are implied by the fumiehing of this domament.

usper-roo.7es

i XN-NF-86-149 TABLE OF CONTENTS I

Page

1.0 INTRODUCTION

................................................. 1 2.0

SUMMARY

...................................................... 2 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE..................... 3 4.0 GENERAL DESCRIPTION.......................................... 6 5.0 MECHANICAL DESIGN............................................ 11 6.0 NUCLEAR CORE DESIGN.......................................... 12 6.1 PHYSICS CHARACTERISTICS...................................... 13 6.2 POWER DISTRIBUTION CONTROL PROCEDURES........................ 15 6.3 ANALYTICAL METH000 LOGY....................................... 15 7.0 THERMAL-HYDRAULIC DESIGN..................................... 23 8.0 STANDARD REVIEW PLAN CHAPTER 15 EVENT ANALYSIS............... 24

9.0 REFERENCES

................................................... 46

f 11 XN-NF-86-149 LIST OF TABLES Table Page 4.1 H. B. Robinson Unit 2, Cycle 12, Fuel Assembly Design P a r ane t e r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 6.1 H. B. Robinson Unit 2, Neutronics Characteristics of Cycle 12 Compared with Cycle 11 Data........................... 17 6.2 H. B. Robinson Unit 2, Control Rod Shutdown Margin and Requ i re me n t s fo r Cyc l e 12. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 8.1 Plant Rated Operating Conditions............................... '39 8.2 Core and Fuel Design Parameters Used in Chapter 15 Event Analysis................................................. 40 8.3 Comparison of Bounding Physics Parameters for Reference Analyses and Cycle 12.......................................... 41 8.4 Bounding Power Peaking Factors and Augmentation Factors for Reference Analyses and Cycle 12.................... 42 8.5 H.B. Robinson Unit 2 Cycle 12 Results of the Analyses of CVCS Malfunction (Boron Dilution).................. 43 8.6 H.B. Robinson Unit 2 Cycle 12 Ejected Rod Analysis, HFP.................................................. 44 8.7 H.B. Robinson Unit 2 Cycle 12 Ejected Rod Analysis, HZP.................................................. 45 h

l iii XN-NF-86-149 l

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 11 Relative Power Distribution Map at 100.0% HFP and 6,473 mwd /MT, Bank D at 216 Steps. . . . . . . . 5 4.1 H. B. Robinson Unit 2 Cycle 12, Reference Loading Pattern for an E0C11 Exposure o f 11,485 mwd /MTU. . . . . . . . . . . . . . . . . . . . . . . . 9 4

4.2 H. B. Robinson Unit 2 Exposure Distribution, EOC11 =

11,485 mwd /MT.................................................. 10 6.1 H. B. Robinson Unit 2, Cycle 12, HFP Critical Boron Concentration versus Cycle Exposure for 2,300 MWt. . . . . . . . . . . . . . 19 6.2 H. B. Robinson Unit 2 Cycle 12, Assembly Relative Power Distribution, 100 mwd /MTU, 2,300 MWt, 3-D XTGPWR............... 20 6.3 H. B. Robinson Unit 2 Cycle 12, Assembly Relative Power Distribution, 1,000 mwd /MTU, 2,300 MWt, 3-D XTGPWR............. 21 6.4 H. B. Robinson Unit 2 Cycle 12, Assembly Relative Power Distribution, 12,650 mwd /MTV, 2,300 MWL, 3-D XTGPWR............ 22

)

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l l

XN-NF-86-149

(

1.0 INTRODUCTION

Results of the safety evaluation for Cycle 12 of the H. B. Robinson Unit 2 nuclear plant are presented in this report. The Cycle 12 analysis reflects plant operation at 2,300 MWt. The topics addressed herein include operating history of the reference cycle, power distribution considera-tions, control rod reactivity requirements, temperature coefficient con-siderations, setpoint analysis, and Standard Review Plan Chapter 15 Event analysis.

The Cycle 12 design requires the loading of forty-eight (48) fresh Exxon Nuclear Company (ENC) supplied fuel assemblies. The forty-eight (48) fresh XN-9 Region 15 fuel assemblies utilize natural uranium axial blankets (NUABs) and gadolinia-bearing fuel rods.

I 4

XN-NF-86-149 1

2.0

SUMMARY

Cycle 12 of the H. B. Robinson Unit 2 nuclear plant is designed to operate at 2,300 MWt starting in the spring of 1987. The characteristics of the fuel an,d of the reload core result in conformance with required shutdown margins and thermal limits. This document provides the safety analysis for the plant during Cycle 12 operation.

The ENC fuel mechanical design is presented in References 1 and 2. The generic Control Rod Ejection Analysis is provided in Reference 23. These analyses are all applicable to the Cycle 12 operating conditions at 2,300 MWt.

l The thermal-hydraulic design of the Cycle 12 core is discussed in Section 7.0 of this document. A review of the applicability to Cycle 12 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 2,300 MWt within the licensed limits. The LOCA/ECCS analysis of record is discussed in Section 8.0.

H. B. Robinson Cycle 11 has been chosen as the reference neutronics cycle due to the close resemblance of the overall neutronic characteristics of Cycle 12 to Cycle 11.

t

Xfi-NF-86-149 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE H. B. Robinson Cycle 11 has been chosen as the reference neutronics cycle due to the close resemblance of the overall neutronic characteristics of Cycle 12 to Cycle 11.

l The measured power peaking factors have remained within the Technical Specification limits for Cycle 11. The total nuclear peaking factor, F Tg, and the radial nuclear pin peaking factor FAH, have remained below 2.32 and 1.65, respectively. Cycle 11 operation has typically been rod free with Control Bank D positioned in the range of 215 steps to AR0; 228 steps being the fully withdrawn position. It is anticipated that similar control bank insertions will be seen in Cycle 12.

The Cycle 11 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 P0Q model is compared to measured values shown in Figure 3.2. A representative power distribution for cortparison is made at a Cycle 11 exposure of 6,473 mwd /MTU with D-bank inserted to 216 steps for a core power level of 100.0% of 2,300 MWt.

1 i

h

1200 . . . . ..... ....

n .

2000 - ~

X X XCALCULATED A A AMEAStRED .

n x . .

a.

n.

v g a 5

r . .

g . .

s 800 _ _

z l w - -

l u . .

o

. . P g

g 400 -

O - .

m . .

200 - -

g '....i....i....i....i.... ....

. . . ._t . . . . i . . . . i . . . . i . . . . i . _ . . l 0 1 2 3 4 5 6 7 8 9 10 11 12 5 CYCLE EXPOSURE (GWD/MTU) M.

C7 m

Figure 3.1 H. B. Robinson Unit 2 Critical Baron Concentration .L versus Exposure for Cycle 11 at 2300 MWt 2 e

e

~-

<x M

XN-NF-86-149 H G F E D C 8 A l l I l l 1.354 0.953 0.926 1.301 0.960 0.279 '

1.003 l 1.072 l 1.307 0.964 1

0.278 8 0.995 l 1.072 1.361 0.951 0.923

-0.7 0 0.5 -0.3 -0.4 0.4 0.5 -0.6 1

1.071 0.881 0.950 1.187 1.117 1.001 1.175 0.246 9 1.070 0.878 0.951 1.189 1.114 0.998 1.177 0.243

, -0.1 1 -0.3 -0.02 0.2 -0.3 -0.3 0.1 -1.3 l 'l l l

,r 'I l l ,

l 1.352 1 0.946 1.274 0.991 1.168 l 1.153 1.061 l

< 10 1.357 0.941 1.268 0.980 1.160 1.152 l.066 1 ,

-0.7 -0.1 0.5 0.4 -0.6 -0.5 -1.1 l

l \ l 0.952 1.180  ; 0.975 1.174 1.185 1.267 0.819 l 11 0.952 1.179 l 0.963 1.164 1.176 1.275 0.831 l l

0 l -0.1 l -1.2 -1.0 -0.7 0.6 l 1.5 I

- 1 ( l 0.932 1.114 1.163 l 1.186 1.201 0.497 12 l 0.927 1.107 1.149 l 1.185  ! 1.214 . 0.503

-0.3 l -0.6 -1.2 -0.1 1.1 l 1.3 l

I l l l

l l 1.301 1.000 1.153 1.271 L 0.508 l 13 l 1.305 l 0.997 1.153 1.289 0.515 l l 0.3 -0.3 0 1 1.4 1 1.5 l I . I l l 1

0.960 1.176 l 1.062 0.822 Measured 14 0.964 1.180 l 1.069 0.833 PDQ 0.4 0.3 l 0.7 1.4 M-C/C*100 l

i l \ l

l I 0.280 l 0.247 l 15 0.280 1 0.246 l I

l 0 l -0.3 l l l l

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Figure 3.2 H. B. Robinson Unit 2 Cycle 11 Relative Power Distribution, Map at 100.0% HFP and 6,473 mwd /MT, Bank D at 216 Steps s

XN-NF-86-149

,z 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, 20 RCC guide tubes, )

and one instrumentation tube. The RCC guide tubes and the instrumentation l

tube are made of zircaloy. Each ENC assembly contains seven zircaloy l l spacers with Inconel springs; six of the spacers are located within the f active fuel region. The fuel rods consist of slightly enriched 002 pellets i inserted'into zircaloy tubes.

The Cycle 12 t'esign reflects the loading of 48 fresh ENC supplied fuel assemblies. This core design contains the third reload of axially blanketed fuel and the fourth relodd of gadolinia-bearing fuel pins for H.

B. Robinson Unit 2. Forty of the 48 fresh Region 15 (XN-9) fuel assemblies contain gadolinia-bearing pins.

The design for Batch XN-9, Region 15, includes natural uranium axial blankets in the top and bottom six inches of the active fuel region of the

[c non-gadolinia-bearing pins. In the 384 gadolinia-bearing fuel pins, l extended natural uranium blankets have been incorporated. The additional f six inches of natural uranium has been addeo to the gadolinia-bearing fuel I pins to flatten the axial power distribution.

The batch average enrichment for the blanketed assemblies is 3.54 w/o U-235. This average enrichment is achieved by using a central axial zone enrichment of 3.85 w/o in fuel pins which contain no gadolinia and 2.70 w/o in the gadolinia-bearing fuel pins. In 16 Region 15 assemblies, 12 fuel pins per assembly will contain 4 w/o gadolinia. In eight (8) Region 15 assemblies, 12 fuel pins per assembly will contain 6 w/o gadolinia. In

XN-NF-86-149 another eight (8) assemblies, eight (8) pins of 4 w/o gadolinia per assembly will be utilized. In addition, in eight (8) Region 15 assemblies, four (4)' pins per assembly will contain 4 w/o gadolinia. In the. remaining eight (8) Region 15 (XN-9) fuel assemblies no gadolinia pins will be used.

Thus, the total number of gadolinia pins required for Cycle 12 is 384.

The projected Cycle 12 loading pattern is shown in Figure 4.1 with the assemblies identified by assembly fabrication ID and by their core location in the prev'ious cycle or by fresh fuel region. BOC12 exposures, based on an E0C11 exposure of 11,485 mwd /MTU, along with Region ids, are

- 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.

i l

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Table 4.1 H. 8. Robinson Unit 2. Cycle 12 fuel Assembly nesign Parawlers Reginn PLSA PDAS 12C "* 13 13** 13 "

• 14 14* 14 " 14'* 14"" 15 l '.* 15 " 15 "

• 15 " "

XN Number 7 6 6 7 7 8 8 8 0 8 8 9 9 9 9 9 No. of Assemblics 12 4 1 8 24 12 8 8 8 16 0 11 ft il il n Pellet Density 94.0- 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 94.0 ij4.0 *j4.0 94.0 94.0 94.0 Pellet to Clad Olametral 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 7.5 Initial fnriclunent (w/o U-235)

Upper 6 inches U02 0.71 2.85 2.85 0.71 0.71 0.71 0.71 0.71 0.71 0.71 0.71 0 /l 0./l 0.71 0.71 0.71 2.20 UO2-Cd 023 Central 132 inches 002 1.24+ 2.85 2.85 3.34 3.34 3.34 3.73 3.73 3.73 3.73 3.73 1. tir. 3.ns 3.05 3.n5 3.h5 UO2-Cd 023 --- --- 2.20 --- 2.37 2.37 --- 2.60(1) 2.(.n(1) 2.60(1) 2.uill) --- 2./nll) 2./n(l) 2./n(I) 2. loft)

Lower 6 Inches UO7 304SSTL 2.85 2.85 0.71 0.71' O.71 0.71 0.71 0.71 0. 71,. 0.71 0./l 0.71 0.71 0.71 0.71

--- --- 2.20 --- --- --- --- --- --- --- --- --- --- -- --- --- a, UO2-Gd 023

• Average 0.86 2.85 2.81 3.12 3.09 3.02 3.48 3.45 3.43 3.41 3.41 3.M 3.57 3.54 3.52 3.52 Initial Cd73 0 , w/o --- --- 4 --- 4 4 --- 4 4 4 6 --- 4 4 4 6 Batch Averaqe liti at BOCl2, HWJ/Hi 8,931 21,342 26,320 21,908 25,368 26,137 9,098 12,840 12,327 13,479 14.424 0 0 0 0 0 Peak Assembly BU at EOCl2, HWd/HT 14,229 33,838 37,602 34,798 37,414 36,936 24,019 27,496 26,517 29,441 28,713 9,910 14.075 11.027 16.591 16,49:1

• 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 a w/o gadolinia per assembly -

• • • • Fuel with 12 pins of 6 w/o gadolinia per assembly i

+ Lower 36 inches of central 132 inches contains 304 $5TL ,

4 PLSA Part length Shield Assembly /..

PDAS Prematurely Discharged Assembly (Region 12) T I Six additional incnes of natural uranium axial blanket fuel is utilized in the top and bottom of gadolinla-tu*aring fuel sitns. ",'

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1 XN-NF-G6-149 R P N M L K J N G F E D C 8 A P.iA PtsA PtsA l G15 N15 J15 1 1 45 ,h49 Nf3 I

i 15 H3 15 2

$01 s33 sat P30 542 534 $02 *

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(3 E2 J5 G! 82 83 3 409 . so 33 %C3 s*3 , hC4 83 $11 Ni3 1

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!! P! i.2 *2 ;I3 35  ?! 5 s;5 515 P:5 M:! si! ;:20 N19 P21 s26 M6 P6 $13 s06 i

P6 M6 L5 G12 J12 E5 06 B6 6

$37 P37 N35 $29 N37 N21 $21 N29 N38 530 N36 P38 533 PLSA

• FLIA A9 L7 P7 N5 09 KS N6 F6 M9 C5 87 ET A9 7 N56 545 NO7 P45 P10 432 PZ6 P17 PZ5 h22 P11 P46 NC8 S46 N46 a

p(g4 ... ... ... ...

p,gA A8 NS J8 PS KS N8 78 BS G8 . c3 R3 8 N52 P25 516 N26 NIS 523 P16 M53 P15 524 NZ3 N25 $19 P29 450 PLs4

• p.5A AT L9 P9 N11 07 K10 N10 F10 M7 C11 69 E9 47 9 N48 547 N06 P48 P9 N24 P31 P19 P27 N30 P24 P47 N5 $48 N!4 P10 M10 L11 G4 J4 Ett D10 310 10 s39 P40 N34 $31 N40 N31 522 N23 N39 532 N33 P39 $40 e

... pong .... .... pggs ...

15 P11 L13 H14 E13 ++ 811 15 11 ,

307 S17 P08 MOS 527 P23 NIT P22 528 MOT P07 S20 S08 l l

N10 04 L14 K12 J14 H9 G14 F12 E14 v4 clo 12 j N16 P15 PO4 N42 P44 N25 P43 N41 P03 P14 N15 l

... ... ... 6 K13 K14 J11 G11 F14 F13 13 N12 S12 P36 NO2 S13 N01 P35 S14 N11 -

15 H13 15 14 503 S35 543 P32 544 536 S04 4 Pins of 4 w/o Gadolin'a PLSA PLSA PLSA l Cycle 11 location 8 Pins of 4 w/o Gadotinia , G1 M1 J1 - or g region number 15

• • • 12 Pins of 4 w/o Gadolinia N55 N51 N47 FabricotIon 30 PLSA Part Length shield &tsembly

• • • • 12 Pins of 6 w/o Gadolinia

+ inert Zirc Rods (+=1)

PDAS Prematurely Discharged Assembly Region 12 Figure 4.1 H. B. Robinson Unit 2 Cycle 12 Reference Loading Pattern for an E0C11 Exposure of 11,485 mwd /MT, Revision 0

f XN-NF-86-149 H G F E D C B A

• • • *** PLSA l l l l l l 26.320 l 14.991 l 0.000 l 24.804 l 25.090 0.000 l 14.335 l 7.032 8 l 37.602 l 29.441 l 16.575 l 36.930 l 37.266 16.569 l 26.450 l 10.076 l l 12 l 14 1 15 l 13 l 13 , 15 l 14 l 13 ,

I I i l 1 1 I

• PLSA l l l l l l I 14.945 i 14.527 l 25.297 13.220 l 12.846 l 21.963 l 0.000 l 5.973 9 29.408 l 28.713 l 37.293 27.926 l 27.491 l 34.798 l 14.007 l 8.682 l 14 i 14 l 13 1 14 l 14 l 13 l 15 l 13 l l l l 0.000 25.242 1 25.409 I 0.000 l 25.617 l 11.821 l 0.000 l 10 l 16.588 l 37.250 l 37.414 l 16.490 l 38.017 l 26.506 l 13.015 l l 15 l 13 l 13 l 15 l 13 l 14' l 15 l l I l I- l l 1 l l l l l PDAS l l l l t 24.801 l 13.248 l 0.000 21.342 l 9.093 1 0.000 l 0.000 l 11 l 36.936 l 27.959 l 16.498 33.838 l 24.004 l 14.517 l 9.918 l l 13 l 14 l 15 l 12 l 14 l 15 l 15 l l l l 1 l l l l l l l l \ l l g i 25.088 l 12.833 l 25.556 l 9.103 l 12.480 l 26.810 l 12 l 37.227 l 27.496 l 37.974 l 24.019 l 24.353 l 32.610 l l 13 l 14 l 13 l 14 l 14 l 13 l l I I I I I I l l l l I l l 0.000 l 21.852 l 11.816 1 0.000 l 26.799 l 13 l 16.593 1 34.720 l 26.517 l 14.526 I 32.603 l l 15 l 13 l 14 l 15 l 13 l l l I I I l l 1 i l l l 14.307 I 0.000 l 0.000 1 0.000 l 80Cl2 Exposure (GWd/MTV) 14 l 26.441 l 14.025 l 13.027 l 9.924 1 E0Cl2 Exposure (GWd/MTV) l 14 l 15 l 15 l 15 l Region ID I l l l 1 l PLSA l PLSA l l 7.031 l 5.973 l 15 l 10.079 l 8.685 l

}

l 13 l 13 l l l 1 4 pins of 4 w/o gadolinia per assembly

• • 8 pins of 4 w/o gadoliaia 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 PDAS Prematurely Discharged Assembly Figure 4.2 H. B. Robinson Unit 2 Exposure Distribution, E0Cll = 11,485 mwd /MTV 1

XN-NF-86-149 5.0 MECHANICAL DESIGN The forty-eight (48) Batch XN-9 reload fuel assemblies are mechanically identical to the previous reload assemblies with the exception of the

.l gadolinia-bearing fuel rods. Each fresh gadolinia-bearing fuel rod  !

contains an enriched gadolinia-bearing region of 120 inches with 12 inches l of natural uranium at each end of the rod. For non-gadolinia-bearing fuel I rods, the 144 inch fuel column includes a six (6) inch column of natural UO2 pell6ts 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 I-analysis of the Batch XN-9 fuel and of the resident XN-7 and XN-8 fuel is contained in Reference 2. This reference is being supplemented for submittal to the utility on about January 31, 1987 to incorporate methodology updates approved in Reference 34. The reference analyses of those performed for the supplement to Reference 2 encompass the expected conditions and 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-8 assemblies, with the exception of the bottom 42 inches of the fuel column, which contains stainless steel inserts.

f

XN-NF-86-149 6.0 NUCLEAR CORE DESIGN

.The H. B. Robinson Unit 2, Cycle 12, XN-9 (Region 15) reload design has been -

developed in accordance with the following reg'uirements:

1. The Cycle 12 reload shall contain 48 new fuel assemblies.

2. The length of Cycle 12 shall be maximized.

3. The rated power for Cycle 12 shall be 2,300 MWt.

4. The length' of Cycle 12 shall be determined based on a projected E0C11 I

exposure of 11,485 mwd /MTU.

5. Cycle 12 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 step change bases as set forth in the FSAR.

I

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 F T g, including uncertainties, shall be i

less than or equal to 2.32 at 2,300 MWt. The integrated peak to l

average pin power, FAH, including measurement uncertainties, shall be less than or equal to 1.65 at 2,300 MWt.

'XN-NF-86-149 1

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

I 6.1 Physical Characteristics The neutronic characteristics of the Cycle 12 core are compared to those of Cycle 11 in Table 6.1. The data presented in the table indicates the neutronic similarity between Cycles 11 and 12. The reactivity coef-ficients of the Cycle 12 core are bounded by the coefficients used in the safety analysis. The safety analysis for Cycle 12 is based on physics characteristics representative of those expected for Cycle 11 lengths of

-500 mwd /MTU and +500 mwd /MTU about the expected length of 11,485 mwd /MTV.

{

The boron letdown curve for Cycle 12 operation at 2,300 MWt is shown in Figure 6.1. As shown, the BOCl2, no xenor., hot full power (HFP) critical boron concentration is predicted to be 1,301 ppm. At 100 mwd /MTd, equilibrium xenon, the critical boron concentration at HFP is 971 ppm. The Cycle 12 length is projected to be 12,650 mwd /MTU (364 EFPD) with no boron at E0C at a power level of 2,300 MWt. j i

6.1.1 Power Distribution Considerations i At a power level of 2,300 MWt at equilibrium xenon conditions,100 mwd /MTU, the calculated XTG peak' FAH is 1.56 including a 4% measurement uncertainty.

At the same exposure, the peak FT Q is 2.13 including a - 3% engineering factor, a 5% measurement uncertainty, K(Z) considerations (32), and an 8%

- allowance for operatior. with PDC-II for +3% target bands.

The peak FaH in Cycle 12 is calculated to occur at a cycle exposure of 1,000 mwd /MTU. The predicted value of FAH at this exposure is 1.56 including the 4% measurement uncertainty. The corresponding F T g, includin5 V(Z) and K(Z) considerations, and appropriate uncertainties, is 2.11.

4 m.,,-,-, , m , , , , . _ - - - , - _

, . , . . - , , ~

XN-NF-86-149 T

The peak F g in Cycle 12 is calculated to occur at a cycle exposure of 5,000 mwd /MTU. The calculated peakTF g at this exposure, again including V(Z),

K(Z) considerations and appropriate uncertainties, is 2.13 with +3% target bands. 'The predicted value of FAH at this exposure is 1.56 including the 4% measurement uncertainty.

The quarter-core radial power distributions are presented in Figures 6.2 through 6.4 for Cycle 12 exposures of 100 mwd /MTU,1,000 mwd /MTU, and (E0C) 12,650 mwd /MTU, respectively.

6.1.2 Control Rod Reactivity Requirements Detailed calculations of shutdown margins for Cycle 12 are compared with Cycle 11 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 12 analysis indicates excess shutdown margins of 1,572 pcm at the BOC and 606 pcm at the E0C. The Cycle 11 analysis indicated excess shutdown margins for that cycle of 2,420 pcm at the 80C and 508 at the EOC.

The control rod groups and insertion limits for Cycle 12 will remain-unchanged from Cycle 11. The control rod shutdown requirements in Table L

6.2 allow for a HFP O Bank insertion equivalent to 600 pcm for B0C and 400 pcm for EOC, to bound the Cycle 12 control rod worths.

6.1.3 Isothermal Temperature Coefficient Considerations The Cycle 12 isothermal temperature coefficients are shown in Table 6.1 for HFP and HZP conditions at both B0C and EOC. At BOC12, following a nominal E0C11 shutdown exposure of 11,485 mwd /MTU, the HFP isothermal temperature coefficient is projected to be -4.2 pcm/oF at a critical boron concentra

XN-NF-86-149 tion of 1,301 ppm. The corresponding HZP critical boron concentration is 1,460 ppm with the isothermal temperature coefficient being -0.6 pcm/0F.

The Technical Specification for the moderator temperature coefficient for Cycle 12 allows a +5.0 pcm/0F (Reference 15) at HZP conditions. With a calculated value of + 1.1 pcm/0F at HZP, the Technical Specification is expec.ted 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 contro' cf the core power distribution is accomplished by following the procedures for " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase II"(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 12 core analyses are 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 chles was accounted for in the calculations.

The PDQ/ HARMONY (13,14) code package will be utilized to monitor the power distribution.

4 4

, ., . - - --- , , - - ,,v. y.---. ...-, , - - - - , _ . - - , , , - -

XN-NF-66-149 Calculated values of FQ, Fxy, and FaH were studied with the 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.

ni -

i i

l.

6 i

)

I i

I XN-NF-86-149 Table 6.1 H. B. Robinson Unit 2, Neutronics Characteristics of Cycle 12 Compared with Cycle 11 Data Cycle 11 Cycle 12 BOC EOC BOC EOC Critical Boron HFP,AR0,(ppm) 1,132 0 1,301 0 HZP, ARO, No Xenon (ppm) 1,268 --

1,460 --

Moderator Temp. Coefficient HFP,(pcm/oF) -4.2 -28.9 -2.9 -29.8 HZP,(pcm/0F) +0.7 -22.6 +1.1 -23.5 Isothermal Temp. Coefficient HFP,(pcm/0F) -5.5 -30.4 -4.2 -31.3 HZP,(pcm/0F) -1.0 -24.4 -0.6 -25.3 (

, U-238 Atoms Consumed Per Total Atoms Fissioned 0.50 0.76- 0.59 0.76 Pressure Coefficient HFP,' 10-6 ao/ psi ----

+3.2 ----

+3.3 HZP,10-6 ap/ psi -0.08 ----

-0.12 ----

DopplerCoefficient(pcm/0F) -1.3 -1.5 -1.3 -1.5 Power Defect (Moderator + j Doppler),pcm 1,574 2,032 1,578 2,083 Boron Worth, (ppm /103 pcm) i HFP -117 -102 -126 -108 HZP -114 -98 -123 -102 '

Prompt Neutron Lifetime (usec) 23.8 23.'4 22.0 21.9  :

Delayed Neutron Fraction 0.0059 0.0052 0.0060 0.0053 Control Rod Worth' of All Rods In Minus Most Reactive Rod,HZP,(pcm) 6,938 5,956 6,000 6,121 Excess Shutdown Margin, (pcm) 2,420 508 1,572 606

--s, w g -es-awm-r- - - + ---w---+m.wp=-px - o ~ m

'-M- - --

--tmnr ,,-r-,- wyy ,,-----y 4 m -

.. . .. . _. .=

XN-NF-86-149 Table 6.2 H. B. Robinson Unit 2, Control Rod Shutdown Margin and Requirements for Cycle 12 BOC 11 E0C 11 BOC 12 E0C 12 Control Rod Worth ARI 8,338 7,356 7,400 7,521 N-1 6,938 5,956 6,000 6,121 (N-1)

• 0.9 6,244 5,360 5,400 5,509 Reactivity Insertion Power Defect (Moderator

+ Doppler) 1,574 2,032 1,578 2,083

Full Power D Bank Insertion 600 400 600 400

Flux Redistribution 600 600 600 600 Void 50 50 50 50 Total Requirements 2,824 3,082 2,628 3,133 Shutdown Margin f

' (N-1)

• 0.9 - (Total Requirements) 3,420 2,278 2,572 2,376 Required Shutdown Margin 1,000 1,770 1,000 1,770 Excess Shutdown Margin 2,420 508 1,572 606 J

I

i'

(

l l

1400.0 l

l l

l 1200.O

\

) f 1000. O C ss N s j g 800.0

\ '

l E N m o ' N m

E '

i 600. 0 '

' g s A

N i

e s ,

G s i 400.0  % ,

s '

• -= '

s s.

200.0 ' '

N s

s.

0. 0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 5 l CYCLE EXPOSURE, GWd/HTU "n

=

, Figure 6.1 H. B. Robinson Unit 2 Cycle 12 HFP Critical Baron Concentration k I

versus Cycle Exposure at 2,300 MWt

• 1 1

XN-NF-86-149 H G F E D C B A 6 l 0.989 l 1.272 1 1.305 l 1.021 1.015 l 1.244 1 0.936 0.209 l l l l l l l PLSA I I I I I I I I I I I I I I I I I I I I I I I I 9 l 1.274 1.237 l 1.007 l 1.259 l 1.255 l 1.046 l 1.094 0.186 I I I I I I i l PLSA l l l 1 I I I I I I l I l I 10 1 1.307 l 1.009 l 0.989 l 1.304 l 1.028 l 1.201 1 0.977 I I I I I I I I I I I I I I I I I I I I I I I I I I l l l l l l l 11 l 1.022 l 1.261 l 1.305 l 1.024 l 1.250 1 1.067 l 0.768 l l 1 l i PDAS l l l l l l l l l 1 I I I I I I I i l i l i l I 12 1 1.017 l 1.258 l 1.031 1 1.251 l 0.965 1 0.432 I I I I I I I I I I l l I I I I I I I l l l .

l l l l l 13 l 1.248 l 1.051 l 1.204 l 1.069 l 0.433 I Assembly Power I I I I I I I i l I I I I

I I l 14 1 0.939 l 1.097 1 0.979 I 0.769 l F A

= 1.307 (10 H)

! I I I I I 1 I I I I FDH = 1.503 (11 F)

\ l l I

I I I F q = 1.820 (11 F 13)

I 15 l 0.209 l 0.187 I l PLSA I PLSA I

! I l I

• 4 pins of 4 w/o gadolinia per assembly

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

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

! **** 12 pins of 6 w/o gadolinia per assembly PLSA Part length Shielding Assembly PDAS Prematurely Discharged Assembly Figure 6.2 H. B. Robinson Unit 2 Cycle 12 Assembly Relative Power Distribution, 100 mwd /MTV, 2,300 MWt, 3-D XTGPWR

. -~ . . - . -

l XN-NF-86-149 H G F E D C B A I I li l I I I I i l l l l 8 l' O.992 l 1.272 1.319'l 1.023 1.016 l 1.254  ; 0.940 ll 0.212 l l l l PLSA i l l l l l l l l l 1 I I I I I 9 1.273 1.241 1.012 l 1.256 1 1.250 l 1.045 1.087 I 0.189 I I l PLSA l I 1 l l 1 l I 10 1.321 1.013 0.992 1.302 l 1.027 1.197 l 0.973 i- 1 I I l ll l l l

> . I I I I

.1 l l I l l L 11 1 1.025 1.257 1.303 .

1.024 l 1.246 l 1.072 l 0.760 l l l PDAS I l l 1 I I I I I I I I I I I 1 I i l 12 1 1.018 1.253 i '

1.030 l 1.247 l 0.969 l 0.438 I I  ; I I I I _I I I I I I

• • • *** j l l 13 1 1.257 1.049 1.200 1.074 1 0.438 Assembly Power I  ! l l l l l l

l l 14 0.943 1.090 0.975 1 0.761 l F = 1.321 (10 H)

A l I I l l F0H = 1.504 (10 H) l F g = 1.797 {l0 H 14) +

! 15 l 0.213 l 0.189 l 1 PLSA l PLSA l l l l 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 PDAS Prematurely Discharged Assembly Figure 6.3 H. B. Robinson Unit 2 Cycle 12 Assembly Relative Power Distribution, 1,000 mwd /MTV, 2,300 MWt, 3-0 XTGPWR

. - . . , - - . . . - - , . . - - . - - , ,m - - , , , ,

XN-NF-66-149 H G F E D C B A I I I I l l 8 1 0.892 l 1.104 l 1.317 I 0.970 1 0.972 1 1.328 l 0.989 I 0.282 l l l l l l l l l PLSA l l l l l l l l l l l l 1 I I I I i l i I I I I I I I I 9 l 1.105 1 1.091 1 0.968 l 1.151 1 1.137 l 1.025 l 1.139 I 0.248 I I I I I I I I l PLSA I I I I I I I I I I 10 j

1 1.317 j

1 0.968 l 0.994 j l l

1.379 l

l 1.003 l

1 1.155 l..

l 1.079 l

I I I I I I I I I I I I I I I I I I I I I l*** l 11 1 0.970 l 1.150 l 1.379 l 1.034 l 1.179 I 1.207 1 0.816 I I I l l PDAS l l l l l l l l l l 1 I I I I I I I I I I I I I I I 12 1 0.972 l 1.137 l 1.004 l 1.179 I 0.958 1 0.508 l 1 I I I I I I I I I I I I I 13 l-l 1.328 l l

1.026 l l

1.155 l l- 1.207 l l l 0.508 l Assembly Power l l l I l l l 1 1 I I I I

l l

14 1 0.989 l 1.139 l 1.079 0.816 l F A = 1.379 (11 F) l I I I I l l l l FDH = 1.467 (10 E) l I I I I I F g = 1.736 (10 E 21) 15 l 0.282 1 0.248 I I PLSA I PLSA I I I l 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 gadolinic per assembly PLSA Part length Shielding Assembly PDAS Prematurely Discharged Assembly Figure 6.4 H. B. Robinson Unit 2 Cycle 12 Assembly Relative Power Distribution, 12,650 mwd /MTV, 2,300 MWt, 3-D XTGPWR

XN-NF-86-149 7.0 THERMAL-HYDRAULIC DESIGN The Cycle 12 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 addition of backup dimples to spacer corner cells and a minor redesign of spacer sideplate corners were reviewed and found to have insignificant thermal hydraulic effect.

Therefore, the hydraulic design of the ENC fuel assemblies is essentially unchanged from previous cycles. No mixed core DNB penalty need be applied.(15) The results of thermal-hydraulic analyses at design hy-draulic 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 gadolinia does not affect the hydraulic character of the fuel.

Evaluation of the effect of axial blankets on the overpower aT and over-temperature 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 12 core under postulated transient and accident conditions is evaluated in Section 8.0 of this document.

l

I XN-NF-86-149 8.0 FSAR CHAPTER 15 EVENT ANALYSIS (Transient and Accident Analysis)

Extensive reanalysis of Chapter 15 events was perfo ved 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 l

methodology described in Reference 19. In Reference 33, :he impact of BOC I values of the effective delayed neutron f raction an.1 prompt neutron lifetime slightly smaller than had been used in the Cycle 10 analyses was evaluated. The LOCA/ECCS refernce analysis is descriaed in Subsection 15.6.5. The applicability of these reference analyses to Cycle 12 is reviewed in this section. Rated operating conditions are sunnarized in Table 8.1.

Significant fuel and core thermal-hydraulic design parimeters employed in the reference Chapter 15 event analyses are listed n Table 8.2. The values are unchanged for Cycle 12. With respect ta fuel design, the reference Chapter 15 event analyses are therefore app'icable to Cycle 12.

The reload fuel design and bounding cycle physics paraneters were reviewed to identify any significant changes from the reference analyses. Fuel design parameters affecting thermal and hydraulic performance are essen-tially the same as those considered in the Cycle 11 event review. Bounding physics parameters for Cycle 12 are compared in Tables 8.3 and 8.4 with the bounding values of those parameters employed in the reference analyses.

The Cycle 12 parameters are unchanged from the reference analyses.

The Cycle 12 core will contain three reloads of axially blanketed fuel.

The Batch XN-7 fuel has six inch natural uranium blankets, while Batches XN-8 and XN-9 have 12 inch natural uranium blan<ets in the gadolinia-bearing rods. These changes have no impact on tle reference Chapter 15 event analyses because these events have been evaluated at Technical Specification peaking limits.

XN-NF-86-149 Setpoint calculations verifying the adequacy of the over-temperature AT and overpower AT reactor trip reset function, f( AI), were performed for Cycle 10 and Cycle 11. Axial power profiles used in the Cycle 10 calculations were prototypic of the Cycle 10 core, which included one reload of axially blanketed fuel having 6-inch blankets. Axial power profiles used in the Cycle 11 calculations were based on a full core of Reload XN-8 /XN-9 type assemblies having 12-inch natural uranium axial blankets in the gadolinia rods. The aggregate of axial power profiles employed in the Cycle 10 and Cycle 11 analyses is considered prototypic of the transition to a full core of XN-8/XN-9 type assemblies, and of Cycle 12 in particular. The validity of the current f(AI) trip reset function has therefore been verified for power distributions prototypic of Cycle 12 in the preceeding Cycle 10 and Cycle 11 setpoint analyses.

The DNB local peaking and rod bow requirements (20) were reviewed for Cycle

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

The Chapter 15 events are reviewed on an individual basis below. Event numbering follows that employed in Reference 19. No event reanalysis is necessary for Cycle 12 because the reference analyses results are found for each event to bound the expected Cycle 12 result.

l XN-NF-86-149 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 resulting from the events. The evaluation is independent of cycle physics parameters and fuel design. The disposition of this event for Cycle 12 is thus unchanged from the reference cycle analysis.

15.1.2 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 physics parameters and fuel design. The disposition for subevent I 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

XN-NF-86-149 for Cycle 12 will remain more positive than -35 pcm/0F (Table 8.3). The reactivity insertion rate employed in the Reference 17 evaluation of subevent 2 is therefore bounding of Cycle 12, and the disposition of the subevent is unchanged from the reference analysis.

15.1.3 Increase in Steam Flow A 10% Increase in Steam Flow event was analyzed for Cycles 10(18) and 11(33). The event results are controlled by the magnitude of the steam flow increase, the moderator and doppler feedbacks, and the control bank worth. The reference analyses considered a range of physics parameters which bounds values expected for Cycle 12 (Tables 8.3 and 8.4). The MDNBR of 1.30 reported in Table 2.2 of Reference 18 is bounding of Cycle 12.

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 compariscn 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 j is thus unchanged from the reference cycle analysis.

l Subevent 2 was determined to be bounded by the Main Steam Line Break event l (15.1.5) based on a comparison of the magnitude of the steam flow for the two ' events. The evaluation is independent of cycle physics parameters and fuel design. The event 15.1.5 analyses (21,22) demonstrate results in accordance with Condition II acceptance criteria.

XN-NF-86-149 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 tank). 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 bounding values in the reference analyses. The reference analyses employed a steam flow rate approximately 20% larger than that predicted by the conservative Moody critical flow model (saturated steam), leading to a significantly conservative predic-tion of moderator cooldown and thus an exaggerated peak power level. Power peaking is cycle specific; the essential similarity af the Cycle 10 and Cycle 12 cores 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 speci.fic variations in power peaking are bounded. The reference analysis is thus applicable to Cycle 12.

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 MDNBR SAFDL, the other to assess the challenge to vessel ,

pressurization limits. The cases were reanalyzed in Cycle 11(33) to assess

XN-NF-86-149 the effect of minor changes in cycle physics parameters. The reference analyses considered ranges of cycle physics parameters which bound Cycle

12. The MDNBR of 1.19 and the peak pressurizer pressure of 2661 psia reported in Table 2.2 of Reference 18 therefore bound Cycle 12.

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 (18,33) in a manner 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 12.

15.2.6 Loss of Non-emergency 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.

The short term phase of the event, during which a challenge to the DNB 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 12 is f unchanged from the reference.

I 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

XN-NF-86-149 15.2.6 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.

15.2.7 Loss of Hormal 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 12.

I 15.2.8 Feedwater System Pipe Breaks 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 12 is thus unchanged from the reference.

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 M2NBR, and a second to assess the challenge to vessel pressurization limits. The reference analysis shows the pressurization case to be clearly bounded by the Loss of External Load (15.4.2), a result not dependent on cycle specific parameters. The MDNBR case was reanalyzed for Cycle 11(33) to assess the impact of minor changes in cycle physics parameters. The reference analyses considered ranges of fue! jesign and cycle physics parameters which bound the Cycle 12 values (Tables 8.2, 8.3 and 8.4) . The MONBR of 1.25 reported in Reference 18 for this event is therefore bounding of Cycle 12.

b I

XN-NF-86-149 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 18. 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(33) to assess the effects of minor changes in cycle physics parameters. The reference analyses considered ranges of fuel design and cycle physics parameters which bound the Cycle 12 values (Tables 8.2, 8.3 and 8.4). The MONBR of 0.90 reported in Reference 18 for this event is therefore bounding of Cycle 12.

15.3.4 Reactor Coolant Pump Shaft Break 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 12 is thus unchanged from the reference analysis.

15.4.1 Uncontrolled 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 I

XN-NF-86-149 Cycle 11(33) to determine the effect of slightly changed cycle physics parameters. The reference analyses considered ranges of fuel design and cycle physics parameters which bound the values for Cycle 12 (Tables 8.2, 8.3 and 8.4). The MDNBR of 1.26 reported in Reference 18 for this event therefore remains bounding for Cycle 12.

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 B0C and E0C conditions, at a spectrum of possible reactivity insertion rates. The limiting case for Cycle 10 was reanalyzed for Cycle 11(33) to assess the effects of slightly changed cycle physics parameters. The reference analyses considered ranges of fuel design and cycle physics parameters which bound the values for Cycle 12(Tables 8.2, 8.3 and 8.4). Therefore, the MONBR of 1.19 reported for this event 'in Reference 18 is bounding of Cycle 12.

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 factors in excess of the rated power design limit is thus deemed unnecessary.

15.4.3 Control Rod Misoperation This event includes the single RCCA withdrawal, the static RCCA mis-alignment, and the Dropped RCCA or RCCA Bank events. The events were reanalyzedt for Cycle 10 in Reference 18 and are addressed individually for t Safety issued raised in References 29 and 30 are addressed in the reference analysis.

p

• ),

' a ,

XN-NF-86-149 Cycle 12 below. Bounding radial peaking augmentation factors for these events in Cycle 12 are .the same or smaller than those employed in the reference analyses (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 augmentation factor and the limiting case of event 15.4.2 are unchanged for Cycle 12 relative to the r,eference analysis, no Cycle 12 specific analysis is required.

The static rod misalignment event results are cycle specific only through the radial peaking augmentation factor. The bounding value of 1.17 calculated for the reference cycle bounds Cycle 12 also (Table 8.4). Thus, the Cycle 10' analysis is bounding of Cycle 12.

The radial peaking augmentation factors for the dropped RCCA/ dropped RCCA bank event eriployed in the Cycle 10 analyses bound Cycle 12 (Table 8.4).

The effect of slightly changed cycle physics parameters were evaluated by reanalysis of limiting cases for Cycle 11(33). These reference analyses considered ranges of fuel design and cycle physics parameters which bound l the values for Cycle 12 (Tables 8.2, 8.3 and 8.4). Therefore, the results l . reported for this event in the reference analyses bound Cycle 12.

i l

l 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.

1 i

i

i f XN-NF-86-149 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. Robinson Unit 2.

-15.4.6 Chemical and Volume Cor. trol System Malfunction that Results in a Decrease 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 80C boron worth coefficient, and the B0C critical boron 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 12 to assess the impact of increased B0C critical baron concentration and a slightly decreased boron worth coefficient. The results of the Cycle 12 analysis are reported in Table 8.5. The maximum reactivity addition due to this event is sufficiently 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 12 indicated that a maximum undetected FaH of 1.72 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 12.

XN-NF-86-149 15.4.8 Spectrum of Rod Ejection Accidents (PWR)

A Control Rod Ejection Accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a Rod Cluster Control Assembly (RCCA) and drive shaft. The consequence of this 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 the procedures developed in the ENC Generic Rod Ejection Analysis, Reference 23. The ejected rod worths and hot pellet peaking factors were calculated using the XTGPWR code. No credit was taken for the power flattening effects of Doppler or moderator feedback in the calculation of ejected rod worths or resultant post-transient peaking factors. The pellet energy deposition resulting from an ejected rod was conservatively evaluated explicitly for 800, MOC and E0C conditions at HFP and B0C and E0C at HZP conditions. The HFP pellet energy deposition was calculated to be 155 cal /gm at BOC, and 164 cal /gm at E0C. The HZP pellet energy deposition was calculated to be less than 40 cal /gm for both BOC 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 ar.d 8.7.

l L 15.4.9 Spectrum of Rod Ejection Accidents (BWR)

This event is not applicable to pressurized water reactors.

I

XN-NF-86-149 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, Uncentrolled RCCA Bank Withdrawal at Power.

The consequenc'es of volumetric addition and effect on the pressure boundary during all operational modes have been earlier addressed in Section 15.5 of the H.8. Robinson Unit 2 FSAR.(24) The evaluation is determined by pressurizer PORV capacity, and is thus independent of cycle.

15.6.1 Inadvertent Opening of a Pressurizer Pressure Relief 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 12.

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.(24) The analysis was performed for breaks ranging in size from a 6

XN-NF-86-149 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) bounds the results of this event.

15.6.3 Radiological Consequences of Steam Generator Tube Failure ,

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 12 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 12 which would change these controlling parameters. The reference analysis is thus applicable for~ Cycle 12.

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

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,

i XN-NF-86-149 i

these analyses are applicable to Cycle 12. Analyses to support operation with a 2.32 Fg are reported in Reference 32.

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

15.7.3 Postulated Radioactive Releases due to Liquid-Containing Tank ratture This event has been reviewed and resuits documented in the H.B. Robinson Unit 2 FSAR. The results of this evaluation are unchanged by the planneu licensing actions and therefore bound Cycle 12.

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 12. Fission product inventories are determined by exposure and power level. The reference analysis employed a maximum design burnup of 44,000 mwd /MIU, o power level of 2300 MWt, and a radial peaking factor of 1.65. These parameters are unchanged; thus, the reference analysis is bounding or Cycle 12.

15.7.5 Spent Fuel Cask Drop Accidents The results of this accident are unchanged from those documented in the H.B. Robinson Unit 2 FSAR. The results of the analysis therefore bouno Cycle 12.

m XN-NF-86-149 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

• 97.3

• 106.lb/hr Active Core Flow

• 92.9

• 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.

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XN-NF-26-149 1 Table S.2 . Core and Fuel Design Parare:ers Used in ,

Chapter 15 Event Analysis .

Number ef-fusi assemblies of all types 157 fr. c:re:

.k :er Of ar: leng:h shieicir.g 12 fuel assemblies

-Fuel assembly pitch 8.466 in. ,

. Fuel assembly design type. 15x15 ,

Fuel rods per assembly 204 Guide tubes per assembly 20 Instrument tub'es per assembly 1, Fuel rod pitch .563 in.

-Fuel rad 0.0. .424 in.

Guide and' instrument tube 0.0. .544 in.

(abovedashpot)

Active fuel length 144 in.

Fuel rod length ,

152 in.

Number of spacers 7.

Max'imum spacer span length 26.2 in.

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l XN-NF-86-la9 Table 8.3 C'emparison of Bounding Physics Parameters for Reference Analyses and Cycle 12 Itir. Referancs Analysis Cycle 12 ECC i;C ECC E;C

1. Mcdera :r Ten eraturs +5.0 -35.0 +5.0 -35.0 Coefficii. , IC-; 10;;F

2. Moderator Pressure -0.56 +3.9 -0.56 +3.9 Coefficient, 10-0 op/ psia

3. Dopoler Coefficient, -1.0 -1.7 -1.0 -1.7 10-5 ap/oF

4. Delayed Neutron Fraction 0.0070 0.0045 0.0070 0.0045

5. Rod Worth, (N-1)*.9, 0.036 0.036 0.036 0.036

6. 8/E*, sec-1 304.3 204.5 304.3 204.5

7. U-238 Atoms Consumed per 0.50 0.76 0.50 0.76 Total Atoms Fissioned Is the effective neutron lifetime G

l- XN-NF-86-149 l:

i Table 8.4 Bounding Power Peaking Factors and Augmentation Factors for Reference Analyses and Cycle 12 Reference Analysis Cycle 12 Nuclear Enthalpy Rise Factor, FAH, 1.65 1.65

(<100% rated power)

Axial Peaking Factor- 1.65 1.65 Total Heat Flux Peaking Factor

• 2.32 2.32 Fsraction of Power Deposited in Fuel 0.974 0.974 FAH Augmentation Used in Single RCCA 1.27 1.27 Withdrawal Event Fag Augmentation used in Static 1.17 1.17 Misalignment Event FdH Augmentation Used-in Rod / Bank Drop Analysis Large Bank Worth cl.30 <1.30 Small Rod Worth - <1.10 <1.10

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

-TABLE 8.5 .H. B. ROBINSON UNIT 2 CYCLE 12 RESULTS OF THE ANALYSES OF CVCS MALFUNCTION Control Rod Time to Reactor Bank Critical Boron Loss of Shutdown Conditions Configuration Concentration (ppm) Margin (min.)

Refueling ARI 982 63.0 Cold Shutdown Case 1 (3%AP Shutdown) SA & 58 out 1,183 20.4 (1 pump) SA & SB out 1,183 21.5 Case 2 SA & 58 out 1,183 18.1-Hot Shutdown SA & SB out 1,170 17.4 Startup BOC HZP PDIL* 1,267 16.5 Power Operation Bounded by analysis in Sections 15.4.1 and 15.4.2

• 5 PDIL - Power Dependent Insertion Limit a

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Table 8.6 H.B. Robinson Unit'2, Cycle 12 Ejected Rod Analysis' HFP.

BOC EOC

. Contribution (a) to Contribution (a) to Energy Deposition,. Energy Deposition, Value (cal /gm) Value (cal /gm)

A. Initial Fuel Enthalpy .

(cal /gm) 74.5 ----

92.7 ----

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

40.8 ----

C. Delta Initial Fuel Enthalpy (cal /gm) '33.7 33.7 51.9 51.9 D. Maximum Control Rod Worth (pcm) 100 121 128 125 E. Doppler Coefficient (pcm/DF) -1.3(e) 0.99(b)- -1.5(e) 0.89(b)

F. Delayed Neutron Fraction,6 0.0060 1.01(b) 0.0053 ~ 1.03(b)

G.

H.

Power Peaking Factor Power Peaking Factor Used(c) 2.4 4.0 4.2-5.0 f

Total 154.7(d) 164(d)

(a) The contribution to. the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient,.and delayed neutron fraction. The energy deposition contribution values and factors are derived from data calculated in Reference 23.

(b) These values are multiplication factors applied to (C + D).

(c) The energy deposition due to maximum control rod worth is a function of the power peaking factor. F z

(d) Total pellet energy deposition (cal /gm) calculated by the equation 7' '

Total (cal /gm) = (C+D)(E)(F) y (e) For this Doppler coefficient, conservative values of -1.1 and -1.4 were assumed at BOC j and EOC, respectively.

Table 8.7 H.B. Robinson Unit 2, Cycle 12 Ejected Rod Analysis, HZP B0C E0C Contribution (a) to Contribution (a) to Energy Deposition, Energy Deposition, Value (cal /gm) Value (cal /gm)

A. Initial Fuel Enthalpy (cal /gm) 16.7 ---- 16.7 ----

B. Generic Initial Fuel Enthalpy (cal /gm) 16.7 ---- 16.7 ----

C. Delta Initial Fuel Enthalpy (cal /gm) 0.0 0.0 0.0 0.4 Maximum Control Rod Worth (pcm) 600 33 600 33 D.

-1.7(e) 1.04(b) _1,g(e) 0.73(b)

E. Doppler Coefficient (pcm/0F) 0.0053 1.13(b) I F. Delayed Neutron Fraction,8 0.0060 1.00(D)

G. Power Peaking Factor 4.9 ---- 9.0 ----

H. Power Peaking Factor Used(c) 13.0 ---- 13.0 ----

TOTAL 34.3(d) 27.4(d)

(a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy deposition contribution values and factors are derived from data calculated in Reference 23. 5 (b) These values are multiplication factors applied to (C + D). y (c) The energy deposition due to maximum control rod worth is a function of the power peaking factor. y (d) Total pellet energy deposition (cal /gm) calculated by the equation

• Total (cal /gm) = (C+D)(E)(F)

(e) For this Doppler coefficient, conservative values of -1.0 and -1.4 were assumed at 80C and E0C, respectively.

XN-NF-86-149

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 Pressurized Water Reactors", Exxon Nuclear Company, June 1975.

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

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

7. XN-75-27 A, Supplement 3, November 1980.

8. XN-75-27 A, Supplement 4, December 1965.

9. XN-NF-77-57(A), " Exxon Nuclear Power Distribution Control for Pres-surized 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, September 1981.

12. XN-CC-28(A): Revision 3, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing (PWR Version)", Exxon Nuclear Company, January 1985.

13. WAPD-TM-678, "PDQ7 Reference Manual", Westinghouse Electric Corpora-tion, January 1967.

14. WAPD-TM-478, " HARMONY: System for Nuclear Reactor Depletion Computa-tion", Westinghouse Electric Corpo'ation, 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.

i XN-NF-86-149

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

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

18. XN-NF-84-74, Rev.1, " Plant Transient Analysis for H.B. Robinson Unit 2at2300MWtwithIncreasedF$H" April 1986.

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 Transjent Analysis for H.B. Robinson Unit 2 at 2300 MWt with Increased FaH, 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 Boron Injection Tank Removal or Dilution 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 Tran-sient for Pressurized Water Reactors", Exxon Nuclear Company, Feb-ruary 1979.

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

25. XN-NF-84-72, "H.B. Robinson Unit 2 Large Break LOCA/ECCS 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 Analy-sis", 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.

XN-NF-86-149

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 Docket No. 50-261/ License No. DPR-23.

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

Subject:

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

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

31. XN-NF-84-68(P), "H.B. Robinson Unit 2 Radiological Assessment of Postulated Accidents", Exxon Nuclear Company, June 1984.

32. Letter, S.R. Zinnerman (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. DPR-23,

33. XN-NF-85-103, Rev. 2, "H. B. Robinson Unit 2, Cycle 11 Safety Analysis Report", Exxon Nuclear Company, January 1986.

34. XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, Revision 1, "Qualifi-cation of Exxon Nuclear Fuel for Extended Burnup", October 1986.

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XN-NF-86-149 Issue Date 1/7/87 H. B. Robinson Unit 2, Cycle 12 Safety Analysis Report DISTRIBUTION FT Adams GJ Busselman LJ Federico RC Gottula JS Holm JW Hulsman JN Morgan TW Patten FB Skogen IZ Stone (2)

MS Stricker CJ Volmer GN Ward RT Welzbacker HE Williamson DocumentControl(5)

CP&L (20)/HG Shaw

.. - _