ML17305A349

Cycle 3 Reload Analysis Rept.

ML17305A349
Person / Time
Site:	Palo Verde
Issue date:	10/24/1989
From:	ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
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NUDOCS 8911060109
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ATTACHMENT PVNGS UNIT 2, CYCLE 3 RELOAD ANALYSIS REPORT 891i060109. 891024 PDR ADQCN, 05000529 PNU

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0 RELOAD ANALYSIS REPORT FOR PALO VERDE NUCLEAR. GENERATING STATION UNIT .2 CYCLE 3 TABLE OF CONTENTS PAGE

1. INTRODUCTION 'AND

SUMMARY

2. OPERATING HISTORY OF THE REFERENCE CYCLE 2'-1

3. GENERAL DESCRIPTION 3-1

4. FUEL SYSTEH DESIGN 4'- I 5-1

6. THERMAL-HYDRAULIC DESIGN 6-1

7. TRANSIENT ANALYSIS 7-1

8. ECCS ANALYSIS 8-1

9. REACTOR PROTECTION 'AND HONITORING SYSTEM 9-1 10'. TECHNICAL SPECIFICATIONS 10-1

11. STARTUP 'TESTING

12. REFERENCES 12-1

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NTRODUCTION AND

SUMMARY

This, report provides an evaluation of the design and performance of

.Palo Verde Nuclear Generating Station Unit 2 (PVNGS-2) during,i.ts third cycle of operation at 100%%u'ated core power of 3800,'MWt and,

- -- NSSS power of 3822 'MWt'. '-'-Operating:conditions':for'.Cycle'3";.have:.been

,, assumed;to be, cons.istent. with -'those of the; previous cycle and:;are-- .

summarized as full power operation under base load conditions. The core will consist of irradiated Batch B, C and D assemblies, along with fresh Batch E assemblies. The Cycle 2 termination burnup has been assumed to be between 394 and 446 EFPD (Effective Full Power Days).

The second cycle of operation will hereafter be referred to in this report as the "Reference Cycle." Reference 1-2 presented analyses for the Reference Cycle.

The safety criteri'a (margins of safety, dose limits, etc.)

applicable for the plant were established in Reference 1-1.. A review of all postulated accidents and anticipated operational occurrences has shown that the Cycle 3'ore design meets these safety criteria.

The Cycle 3 reload core characteristics have been evaluated with respect to the Reference Cycle. Specific differences in core fuel loadings have been accounted for in the present analysis. The status of the postulated accidents and anticipated operational occurrences for, Cycle .3 can 'be summarized as, follows:

1. Transient data are less severe than those of the Reference Cycle analysis; therefore, no reanalysis is necessary, or

2. Transient data are not bounded by those of the Reference Cycle analysis, therefore, reanalysis is required.

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For. those transients requiring reanalysis (Type '2), .analyses are

,presented in Sections. 7 and,8. showing;.'results..that meet .the establ.ished safety criteria.

The Technical Specification .changes;needed-for Cycle 3 .are

summarized.,in. Section,;l0:.and,described:.in-.'greater,;detai;l,.in..separate

,; license amendment. applications.

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2.0 OPERATING HISTORY OF THE REFERENCE CYCLE The Reference Cycle began with initial criticality"on Hay 15, 1988.

Power ascension began on Hay 23, 1988, and on Hay 30; 1988, the unit reached full power.

It is presently estimated that Cycle 2 will terminate :on or about February 14, 1990. The Cycle -2 termination point can vary"between;-394 and 446 EFPD to accommodate .the plant.= schedule and .stil,l..be- wi,thin .the assumptions of the Cycle 3 analyses.

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2.0 OPERATING HISTORY OF THE REFERENCE CYCLE The Reference Cycle began with initial criticality on Hay 15, 1988. -Power ascension began on Hay 23, 1988, and on Hay 30, 1988, the unit;reached full power.

It is presently estimated that Cycle 2 will terminate on or about February 14, 1990. The Cycle 2 termination point can vary between 394 and 446 EFPD to accommodate the plant schedule and still be within the assumptions of the Cycle 3 analyses.

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3. 0 'GENERA DESCRIPT ON

, -The "Cycle;-.3"core will consist -of. those-assembly"types"and" numbers -,

" listed. in Table 3-1.':':Sixty-nine",Batch B.'ssembl.ies'.;and".twenty=.eight

,=;"~Batch:,C,'wi,ll'. be~removed.>from.the. Cycle:.,'2:.'core..to'make:-way,.for,.-.-,

. - "ninety-six,fresh,".,Batch..E:."assembl.ies.'"Thirty,:-:s.i'x Batch..C'nd,all.,-

.Batch.D:-assemblies, now .in .the;:,core-wil.l.'be-retained. One .Batch B...

assembly discharged at EOCl will be reinserted into the core.

Figure 3-1 shows the poison shim and zoning configuration for those assemblies.

The reloadbatch will consist of 24 type EO .assemblies, 16 type El assemblies with 16 burnable poison. shims per assembly, 24 type with 16 burnable poison shims:per assembly, 8'ype E3 E2'ssemblies assembl;ies with 16 burnable poison..shims per assembly, 4 type E4 assemblies"with'8 burnable. poison"shims per assembly,, and .20 type E5

".assembl'ies .with 12 'burnable poi'son rods per assembly. These sub-batch types are fuel zone-enriched and their configurations are shown, in Figure 3-2'.

, The, loading pattern for. Cycle, 3,,showing fuel type. and location, is

,displayed in Figure 3-3.

Figure 3-4 displays the begi'nning of Cycle 3 assembly average burnup distribution. The burnup distribution is based on a Cycle 2 length of 420 EFPD.

,Control element, assembly,,patterns .and. in-core,i.nstrument locations-will. remain unchanged from the, Reference Cycle and are shown in Figures 3-5 A & B and Figure 3-6, respectively.

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~O TABLE 3-1 PALO VERDE 'NUCLEAR 'GENERATING STATION 'UNIT' Cycle 3"Core Loading Initial Total Number Assembly Fuel Rods Initial Number Shim of Desig- ..Number of per Enrichment Shims/ Loading -

Fuel- Shim nation Assemblies 'Assembly U-235) Assembly

'w/o

(gm BIO/in) .Rods Rods 208 2.78 16 .018 208 16 12 1.92 12 36 224 3.30 8064 12 2.78 432 DO 32 184 4.02 0 5888 52 3.57 1664 Dl 20 168 4.02 16 .022 3360 320 52 3.57 1040 168 4.02 16 .020 1344 128 52 3.57 416 3 16 168 3.57 16 .022 2688 256 52 3.09 832 D4 172 3. 57 12 .008 688 48 52 3.09 208 DS 28 172 3.57 12 .020 4816 336 52 3.09 1456 EO 184 4.03 0 4416 52 3.70 1248 El 16 168 4.03 16 .016 2688 256 52 3.70 832

'E2 24 168 3.70 16 .020 4032 384 52 3.40 1248 E3 168 3.70 16 .016 1344 128 52 3.40 416 176 4.03 .012 704 32 52 3.70 208 20 172 3.70 12 .020 3440 240

52. 3.40 1040 TOTAL 241 54732 2144 3-2

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FIGURE 3-1 LOADING OF ASSEMBLIES SHUFFLED FROM PREVIOUS CYCLE Ini ti al Assembly Fuel Rods Initial Number Shim Desig- Number of per Enrichment Shims/ Loading nation Assemblies Assembly (w/o U-235) Assembly (gm 810/in) 208 2.78 16 .018 12 1.92 36 224 3.30 12 2.7&

DO 32 184 4;02 0 DO 52 3.57 01 20 168 4.02 16 .022 52 3.57 02 168 4.02 16 .020 52 3.57 D3 16 168 3.57 16 .022 Dl 52 3.09 D2 D3 D4 172 3.57 .008 52 3.09 05 28 172 3.57 12 .020 52 3.09 D4 D5

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Ih FIGURE 3-FRESH FEED ASSEMBLY L LOADINGS WATERHOLE AND SHI CEMENT~~

SUB-BATCH EO -, 24 Assemblies SUB-BATCH E3 8 Assemblies Q 4.03 w/o U-235 0 3.70 w/o U-235 8 3.70 w/o U-235 H 3.40 w/0 U-235 g 84C-. L2 0;016 gm 3

Shim Pin 8-10/in SUB-BATCH El - 16 Assemblies SUB-BATCH E4 4 Assemblies

%s 0 4.03 w/o U-235 0 4.03 w/o U-.235 H 3.70 w/0 U-235 H 3.70 w/0 U-235 84C-AL2 Shim Pin 0.016 gm 3

8-10/in ~ 84C-AL203 Shim Pin 0.012 gm 8-10/in SUB-BATCH E2 24 Assemblies SUB-.BATCH E5 20 Assemblies Q 3.70 w/o U-235 0 3.70 w/o U-235 H 3.40 w/0 U-235 IH 3.40 w/0 U-235 g 0.020 84C-AL203 Shim Pin g 0.020 84C-AL203 Shim Pin gm 8-10/in gm 8-10/in

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FIGURE 3-3 PVNGS UNIT 2.CYCLE 3 .FUEL,6MANAGEMENT F C EO 00 01 C

,02 EO C

EO C

'lEO El C 00 El 00

.05 EZ D3

'E3 05 E4 05 E5

'EO 01 El Dl E5, D3 E2 05 C 00 El E5 04- E5 C 00 EO 00 05 E2 03 'E5 05 E2 02 00 03 E3 05 E2 C EZ Dl DO 01 E4 05 E5 05 00 02 00 Pin Enrichments 8,Zoning Shim Avg.

Loading, No. of Assy.

EO 0 236 184 4.03 52 3.70 WW 24 3.957 El 16 220 168 4.03 52 3.70 .016 16 3.952 E2 .16 220 168 3.70 52 3.40 .020 24 3.629 E3 16 220 168 3.70 52 3.40 .'016 8 3.629 228 176 4.03 . 52 3.70 .012 4 3.955 E5 12 224 172 3.70 52 3.40 .020 20 3.630 96

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Figure 3-4 PALO VERDE U2C3

'4 30,'387' . "0 '10; 508:21,,902'0 29,853 16,056 13,589 22,227 12 13 14 '5 '6

.21,672 22,365 22,289 0 20,885 18 19 20 21 22 23 24 , 25 29,840 30,589 0 24,069 22,271 26 27 28 29 30 , 31 32 33 0 22,293 23,.1'99 0 22,782 0 21)565 34 36 37 38 39 40 41 30,374, 16,058 '0 '24'., 060 0 21,216 24,795, 11,577 45 47 50 51 0 13,583 22,277 22,703 20,881 0 21,792 52 53 54 56 58 59 60 10,505 22,213 22,244 0 24,788 21,916 18,,144 61 62 63 65 66 , 67 68 69 21,949 0 20,886 21,559 11,578 21,'497'18,147 19,925 Assembly Average Burnup at BOC3 (HWD/T) 3-6

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- LEAD REGULATING BANK 4 - SECOND REGULATING BANK 3 - 'THIRD REGULATING BANK 2 - FOURTH REGULATING BANK I - LAST. REGULATING 'BANK B - SHUTDOWN BANK B A - SHUTDOWN BANK A P2

- PLR GROUP 2 PI . PLR GROUP I 5

S - SPARE. CEA LOCATIONS 2

S 3 S 10 12 13 ,14 15 16 17 18 A 1 A 19 20 22 23 24 25 26 27 28 29 30 31 P2 3 P;a .2 4 32 33 35 36 37 38 39 40 41 42 43 45 46 8 8 47 48 49 50 51 52 53 54 55 56 57 , 58 60 61

,P1- P1 76-62 , 63 A'4 65 8

66 67 68 69 A

70 71 A

72, 73 4

74 75 8

77 A

78 79 80 82 83 84 85 86 87 89 90 91 92 93'2 94 95 S P2 S 96 97 98 99 100 101 102 103 104 105 106 107 108 'I 09 110 111 112 1 8 A 3 '3 A 113 114 115 116 117 118 119 120 121 122 123 '124 125 126 127 128 129 V

3 5 P1 5 130 131 32 133,134 135 .136 137 138 139 140 141 '42 143 144 145 146 8 A A 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163

'S P2 P2 S 164 165 '1 66 167 168 169 170 171 172 173 174 175 176 177 178 179 180 A 8 4 ,A A 181 182 183 184 185 186 187 188 189 190 191 192 193 194 195 2 P1 P1 196 197 198 199 200 201 202 203 204 205 206 207 208 209 21 0 8 8 8 8 '8 ,8 211 212 213 214 215 216 217 218 219 220 221 222 223 P2 Pz 4 224 225 226 227 228 229 230 231 232 233 234 A 1 1 A 235 236 ~ 237 238 239 240 241 3 S AR IZONA Figura

. Pala Varda Nudaar Ganaraiinl CEA 8ANK IDENT1F1CATIQN 3.6A Qadoa

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ARIZONA . Figure Palo Venfe Nuolear Generating CEA PATTERN 368 Station

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17 a i 94 's" 13 14 70 71 72 '2 74 7$ ,7$

19 a al 'ao .$7 a9,90' l 92 .54 20 21 22' 23 24

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II 101 10$ 107 110 11'1 112 27 '02 112 114 11$ 'l l 117 119 120,:121 122 124 127i - 12$

29 30: '31 1%I '121 122 1 124 131 12$ 'l27 1& 140 14'1 '142 142 144- 14$ '144 35 36 38 159159 14 149 39 1 1$ 1 152 154 15$ 150 1$ 7.

'11$ 1 152 194 15$ 1$ $ 1$ 7,1 $ 9 170 171 172 172 174 - 175 17$ '17$ ' 179

'44 46 1$ 1 1$ Z 1$ 2 1$ 4 1$ $ 7 1$ $ 199 190 '191 192 194 185 47 48,,'$ 49 50 19$ '197 '19$ 207 210 51 52,, 53 211 212 212 21$ i 217 2'1$ 219 ~ '21

57 221 .222 ZXL 58 59 ZW 229 240 241. '

61 AR IZOMA Figure Palo Vanfa INSTRUMENT LOCATIONS Nudaar Ganaratinl Station

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4.0 FUEL SYSTEM DESIGN 4.1 'MECHANICAL DESIGN The mechanical design-of the. Batch E "reload-fuel 'assemblies, is .

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identical'to the design of the Reference Cycle Batch D reload fuel assemblies except for,a modification to the poison rod assembly design, lower end fitting,,and center guide tube design. .No changes in mechanical design. bases have, occurred since the original fuel designs A design feature was .incorporated into Batch E to improve the burnup .--

capability of the poison rods. The poison rod assembly design was

, modified, by,increasing the overall,.ilength from 160".918 inches-to:

161.168 inches. This provides greater internal void volume which enables higher burnups with poison rods with higher B-10 loadings while reducing end of. life .internal pressure. In addition,. this change makes the fuel and poison rods equal .in length.

The 1'ower end fitting design was changed from a two piece assembly to a single piece casting with a recess for the center guide tube to fit within the flow plate.

The length of the center guide tube was increased from 163.715 inches to 163.965 inches in order to fi.t within; the new lower end f'itting ~

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t4 GUIDE TUBE WEAR Twenty of the 'fuel assembl.i'es that;had CEA'.s located .in, them during

-:'Cycle 1"at Palo'Verde Unit'1".:were"inspected"for."guide-,tube. wear.. =-='-. ",-

That",inspection: was':part: of,therequired .licensing;procedures.

'"-'-"'requi'red:;by the"NRC.for all,-.pl'ants- a ter ".the -,first:cycle of

..operation (References, 4-.1,, 4-7, and-4-8..),;,.-A:simi".lar;:program: was..

".* ..also. performed",on'nit 2'uring the .firs 'refueling, outage 'C '

'(:Ref erence 4-2, and 4-.6-):. The=number -.of..assembl;;i es-.:.inspec-ed-.=:, or .:..;.=-:

':,guide, tube.~wear:was::determined"based.- on:the resul'ts of the Unit 1 inspection. The inspections revealed that guide tube wear was minor and wi;ll, not adversely affect the fuel assembly performance throughout its:,expected life in the core. Thus no auide, tube"wear inspections are necessarv.

4.3 THERMAL DESIGN The thermal 'performance of composi,te'uel pins that envelope the pins, of fuel batches B, 'C', '0'nd E present -in Cycl'e 3:have been --

evaluatedusing he FATES3A,version of he C-E fuel evaluation model (References \4-'3 .and 4-4) as approved by the .NRC .(keference 4-5).

-,-.FATES3A is, the version of FATES3,that incorporates the arain .size

-" "'restriction aiven 'in" Reference 4-5'.- -The'analysis"was"performed

',using.,a power. history thaL .enveloped the power.and .burnup 'j evels representative of the peak pin at each burnup interval', from beginning, of cycle to end of cycle burnups. The burnuo ranae analyzed. is in excess of that expected at the end of Cycle "..

CHEMICAL DESIGN The metallurgical requirements of the fuel',"cladding..and the fuel assembly'tructural members for the Batch E fuel are essen ially identical to those of the fuel batches included .in Cycle 2. The exception being the reduction of the tin conten in the fuel cladding for improved corrosion resistance. .thus, the chemical or metal:.lurgical performance of the Batch E fuel is enhanced from the performance of the Cycle 2 fue'1 and hereby, providing the poten-ial for higher,burnup., Although the,tin content, has .been'educed, it remains within the standard specification.

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4.5 SHOULDER GAP ADE UACY

.Measured .shoulder,gap data'.aequi,red .from. post 'Cycle,l inspection;,.of...

- '...," fuel";assemblies,.at., PVNGSUnits 'l,and, 2 "confirmed the,'conservatism;of ."

.the .shoul,der gap evaluation;.technique for. the PVNGS-;,fuel -(.references,.

1 and 4-2). 'his-'evaluation technique'redicts-adequate-shoulder- -.-

gaps for all fuel. operating in Cycle 3.

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5.0 NUC AR DESIGN

'5. ' 'PHYS ICS'"CHARACTERISTICS 1

5.1. 1 Fuel Mana ement

":-.The;Cycle. 3 core:makes -:use. of a -low-leakage .fuel management -scheme:;

in which previously burned Batch C assemblies are placed on the core periphery. Most of the fresh .Batch E assemblies are located throughout the interior of the core where they are mixed with the

.previously burned fuel in a pattern that minimizes power peaking.

With this loading and a Cycle 2 endpoint at 420..EFPD, the Cycle 3

reactivity lifetime,for full power oper ation is expected to be 430,.

EFPD. Explicit evaluations have been performed to assure 0 applicability of al.l analyses to a Cycle,2 termination burnup of between 394 and 446 EFPD and for..a Cycle 3 length up to 456 EFPD.

.Characteristic physics parameters for Cycle 3 are compared to those "

of the

Reference:

Cycle in Table 5-1. The values in this .table are

-" 'ntended'o represent'nominal core parameters. Those-values used .in the safety analysis (see Sections 7 and 8) contain appropri ate uncertainties, or incorporate values to bound future operating cycles, and in all cases are conservative with respect to the values reported in Table 5-1.

Table 5-.2 presents a summary of CEA reactivity worths and allowances for the end of Cycle 3 full power steam line break transient with a comparison to the Reference Cycle data. The full power steam line break was chosen to illustrate differences in CEA reactivity worths for the two cycles.

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i The CEA:core locations and, group -identifications, remain the same as in the Reference Cycle:. The..power dependent, inserti.on limit-.,(PDIL).

..for regulating .groups.,and part. length.-.CEA",groups ..i,s .:shownin Figures

'-',.=.:,5;l,.and 5-:2,;respecti.vely.'..Table..5-3.'shows'--the"react'ivi.ty,worths:of,.

,various 'CEA. groups cal,cul'ated.'..at'.,ful.l,=power,.':condi,tions,.for, Cycle,-3,,

and the Reference Cycle.

5. 1.2 Power Distribution Figures 5-3 through 5-5 illustrate the calculated All Rods,0ut (ARO) relat'ive assembly'.power densities during Cycle 3. ,-The, one-pin planar radial power.,peaks (Fxy) presented in these figures represent

.the'maximum over-..the,,mid .eighty;;percent of. the core axially. Time po'ints .at, the beginning,.-,middle,,and .end,of..cycle .were .chosen.to

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display the variation'. in'-'assembly .and 'maximum .planar, radial peaking" as a function of .burnup.

.'-." Rel'ativeas'sembly'-power:densi,ties I

for rodded configurations are g'iven for BOC and. EOC in"Figure's '5-6'through. 5-11,.;, The rodded.,

. configurations shown are those allowed by the: PDIL at full power:

..::. ;,part l.ength CEAs (PLCEAs),,Bank 5,.:.,and Bank 5..pl.us the.,PLCEAs.

The radial'ower dis'tributi'ons described in this section, are calculated data which do not include any uncertainties or allowances. The calculations performed to determine these radial power peaks explicitly account for augmented power, peaking, which is characteristic of .fuel; rods- adjacent to the 'water. holes.

Nominal axial peaking factors are expected to range .from 1.21 at

'BOC3 to l. 11 at EOC3.

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PHYSICS ANALYSIS METHODS

.Anal tical In ut. to In-Core Measurements In.-;core,,detector measurement'.,constants,,to be.;.usedin..eval.uating the

,reload, cycl'e power.distributi'ons wil;1';be.'.calcul'ated:;i'n;accordance Reference 5-1. ROCS-DIT. with, the.,HC module .w'i,l'l.'- be':used;

.'with ROCS-DIT. and the HC-module'.have-.been".approved:.-.for+this;.:application~--.

in. Reference 5-2.

5.2.2 ,Uncertainties in Measured Power Distributions

,'The- planar,;radial power; distribution, measurement uncertainty of 5.3/, based on Reference 5-1, will be applied to the Cycle 3 COLSS

.: ",.--,:and-.CPC",on-1~i'ne;calculations';which" use pl'anar"radial:power"'peaks.;; ..

The axial and three, dimensional, power.,distribution measurement

" uncertainties are" determined in: conjunction with .other monitoring =".

-. ", .'and.:protection:system.measurement-;.uncertainties, as was done for Cycle 2.

5.2.3 Nuclear Desi n Methodolo

-'- The .Cycle 3.nuclear, design was performed .with two and three dimensional core models using the ROCS and HC computer .codes employing DIT calculated cross sections.. ROCS, HC, and DIT were described in Reference 5-2.

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TABLE 5-1 PVNGS-'2 CYCLE 3 NOMINAL,'PHYSICS CHARACTERISTICS.

Reference Dissolved Boron  ;.Uni ts 'Cele .~Ccl e 3 Dissolved Boron Concentration for Critical.ity, CEAs Withdrawn, Hot Full Power PPH 1116 1092 Equi'librium,Xenon, BOC Boron Worth Hot Full Power, BOC PPH/%hp 120 121 Hot Full .Power, EOC 'PPH/%h,p 95 91 Moderator Tem erature Coefficients

-: Hot.:;Ful.l,;Power,,Equi.l.ibri:um:."Xenon Beginning of Cycle 10-4hp/ F -0.5 -0.6 End of Cycle 10-4b,p/ F -2.0 -3.0 Hot Zero Power, Beginning of Cycle 10-4hp/'F +0.2 Do ler Coefficient Hot Zero Power, BOC 10-Shp/ F -1.8 -1.9 Hot Full Power, BOC 10-5hp/ F -1.4 -1.5 Hot Full Power, EOC 10-5hp/ F -1.7 -1.7 Total Dela ed Neutron Fraction eff BOC .0063 .0063 EOC .0052 .0051 Prom t Neutron Generation Time 1*

BOC 10-6,sec 21.7 20.7 EOC 10-6 sec 27.7 28.1 5-4

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TABL'E 5-2 PVNGS-2 CYCLE 3:L'IMITING VALUES OF

'REACTIVITY WORTHS'AND.'ALL'OWANCES "FOR;":HOT FULL..POWER:,STEAM L'INE. BREAK, IAp, ',END-OF-CYCLE .(EOC)

-:Reference

~0c1 e ~Cele 3

'Worth of al.l CEAs Inserted -16.0 -17.6

2. Stuck CEA Allowance +4.3 +4.0 3.' --~.Worth'f'"all.:CEAs"Less,.Highest Worth CEA Stuck Out -11.7 -13.6 Full Power Dependent,Inserti'on Limit -CEA Bite +0.2 +0.2

5. 'Cal cul ated,.Scram Worth -11.5 -13.'4.

6. ,'Phys,i cs,Uncertainty +1.2. +1.3 7-..: ~

.Other Allowances .(losses due to voiding) +O.l +0.1

8. Net Available Scram Worth -10.2 -.12. 0

9. Scram Worth Used in Safety Analysis -10. 0: -10;2

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TABLE 5.-3 PVNGS-2 CYCLE 3

.REACTIVITY WORTH. OF, CEA. REGULATING ..GROUPS AT HOT FULL 'POWER, leap Be innin of C cle ~Ed f Regulating Reference 'Reference CEAs ~Cc1 e ~Cele 3 ~Cc1 e ~Cele 3 Group 5 -.25 -.28 -.'29 -.29 Group 4 -.39 -.41 -.46 -,'41 Group 3 -.67 -.91 -.74 -.88 Note:

Values shown .assume sequential group insertion.

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I'AllT I.KNOTIICEA NlSITION. INCIIES WITIN)flhWN

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Figure 5-3 PALO VERDE U2C3 F 1. 532 BQ 50

,;: -.0."3027':= 0. 6978 0.6328 . 0. 57.71 10 0.3515 0.8511 0.8627 0.9748 0.9041 1.1504 12 13 15 16 17 0.5154 1.0368 1.0256 1.1985 0.9851 1.1657 1.0342 18 .19 20 , 21 22 23 25 0.3521 1.0379 0.8198 1.2307 0.9396 1.1768 1.0308 1.2433 26 27 28 29 30 31 32 33

0. 8525 1'."0277'.'231'9 '1':1'450 '.2928 1.0827 1.2181 1.0808; 34 35 36 37 38 39 40 41 0.3034. 0.8644, 1.2007 0.9411 1.2939 1.1465 1.3249 1.0055 1.2895 43 44 50 51 0.6996 0.9771 0.9873 1.1790 1.0849 1.3256 1.1597 1.3437 1.2813 53 54 56 57 59 60 0 '349 0.9068 1.1685 1.0332 1.2203 1.0068 1.3442 1.2654 1.2468 61 62 63 64 66 67 68 69 0.5812 1.1540 1.0368 1.2459 1.0828 1.2915 1.2854 1.2465 1.0703 ARO Assembly Relative Power Densities at Hot Ful.l Power with Eq. Xe BOC3 5-9

0 e.

4H d 0

Figure 5-4 PALO VERDE U2C3 F =1. 488 2 89 30 0.3045 0;6674 0'6261 . 0..5917

'10 0.3386 0.7955 0.8351 0.9413 0.9068 1.1646 13 ,15 16 17 0.4871 0.9461 0.9987 1.2672 1.0148 1.2779 1.0685 18 19 20 21 '2 23 25 0.3390 0.9467 0.8071 1.3053 0.9814 1.3034 1.0819 1.3194 26 27 29 30 31 32 33

,0."7,960 1.'0001 1.'3060', 1. 1626 p1,3412 1.1072 1.3071 1.0776 34 35 36 37 38 39 40 41 42 0.3046 0.8356 1.2682 0.9822 . 1.3418 1.1257 1.3355 0.9758 1.1765 43 45 46 48 49 50 51

'.6676 0.9418 1.0155 1.3045 1.1088 1'.3358 1.1132 1.3173 1.1445 52 55 58 59 60 0.6265 0.9074 1.2786 1.0829 1.3081 0.9763 1.3172 1.1151 1.0483 61 62 63 66 67 68 69 0.5936 1.1655 1.0690 1.3202 1.0783 1.1774 1'. 1478 1.0477 0.8866 ARO Assembly Relative Power Densities at Hot Full Power with Eq. Xe MOC3 5-10

4 pg 2

'IP c

,K

Figure 5-5 PALO VERDE U2C3 F 1.467 89 28 0;3383 --0.6990 -'0.6616 0.6317

'0 0.3684 0.8127 0.8561 0.9494 0.9192 1.1646 12 13 14 15 16 17 0.5203 0.9530 1.0055 1.2998 1.0254 1.2943 1.0597 18 19 20 21 22 23 25 0.3688 0.9536 0.8325 1.3238 0.9964, 1.3397 1.0800 1.3174 26 27'8 '29 30 31 32 . 33 0;81'28' '1'.'0063'1';3239 '1".1'370, '1.3209 1.0888 1.3184 1.0536 35 36 37 38 39 40 41 '2 '4 0.3381 0.8557 1.2993 0.9963 1.3208 1.0814 1.2984 0.9534 1.1034 43 45 46 49 50 0.6983 0 '487 1.0246 1.3388 1.0891 1.2981 1.0647 1.2750 1.0651 52 53 54 56 57 58 59 60 0.6609 0.9183 1.2927 1.0785 1.3174 0.9531 1.2745 1.0368 0.9651 61 62 63 64 66 67 68 69 0.6322 1.1633 1.0581 1.3150 1.0526 1.1032 1.0675 0.9645 0.8254 ARO Assembly Relative -Power Densities at Hot Full Power with Eq. Xe EOC3 5-11

0

~ *4 I

Figure 5-6 PALO VERDE U2C3 F 1.512 BQ 30

'0'.3162 -'0.7384 .'0.'6659 -.'0 6034

'6a~ M t 4 0 5 8 10 0.9130'.9215 7'.3694 1.0370 0.9461'.1982 12 13 14 15 16 1241'.1003 .-1. 2630 17'0.5522 1:. 1..0209 1.1679 1.0289 18 19 20 21 22 '23 24 25

'0.3701; 1.1253 0.8678 '.2984 '0.9601 1.1671 0.9637 1.0914

, 34 26 35 '6 27 28 37 29 38 '9 30 0.'91'47 '~1;:1028 -.'1'--.'.2997'-'-1':;1 976 '-1'. 3195 31 40 1.0562 32 1.0539

'41 33 0,.6350 0'.3170 ;0'.9236 1.'2657 0'.9618 '.3207 1.1598 1.2937 0.'9153 1.1403 43 44 '5 46 47 "

48 '9 50

" 0.'7405 *':0236 '1.'1697 1'. 0585 1'.2443 1.0399 '. 1. 2945 1.1371 1.3010 52 53 54 56 57 58 59 60 0.6683 0.9494 1..1713 . 0.9664 1.0559 0.9164 1.3015 1.2622: 1,.2549 61 62 63 64 65 66 .. 67 68 69 0.6078 1.2027 1.0322 1.0942 0.6363 1.1420 1 '480 1.2546 '.0421 0 Assembly Relative Power Densities at Hot Full Power with Eq. Xe BOC3 with Bank 5 Inserted 5-12

0 WO 1 l$'

0

Figure 5-7

'PALO 'VERDE U2C3

.F 1.630 3 BQ 50

'0.'29 1'1 ."0.'6875 0..'6309 '0. 57.79, 6 10 0.3370 0..8345 0.8434 .0.9553 0 '042 1.1660 12 13 14 15 16 17

".. 0','.5029 1..0244 1.0059 1.1548 :0.8868 1.1515 1.0508 18 19 20 21 22 '3 25 0.3376 1.0256 0.7880 1.. 1784.,0. 9008 1. 1'496 1.0432 1.2731, 26 . 27 '8 29 30  :

31 32 33 "0".8362: '1'."0082 1'. 1'797 1'0226,"'. 2703 1. 0981 1.2519 1.1308 34 35 36 37 38 39 40 41- 42 0'.'2919 0.8453 '1.1573 '0 9024 1'71'5 1.1695 .1.3739 1.0459 '1;3859 43 45 '46 47 50 51 0'. 6894 0.:9581 0.8891 '1.1521 '1'*.

1006 1. 3748 1. 2246 1.4255 1.3772 53 54 , 56 57 ,58 59 60

'0.6333 0.9074 1.1548 1.0461 1.2543 1.0473 1.4260 1.3522 1.3258

. 61 62 63 66 67 68 .69 0.5821 '.1703 1.0540 1.2762 1.1331 1.3882 1.3814 1'. 3256 1.0047 Assembly Relative Power Densities at Hot Full Power with Eq. Xe BOC3 with PLCEAs Inserted 5-13

0 t

(fF

>P

Figure 5-8 PALO VERDE U2C3 F 1.566 BQ 50

'0.'3146 '.7358 '*'0.'667,9 0.6079 '.'0 0.3720 0'.9134 0.9097 1.0124 0.9405 1.2031 12 13 15 16 17

.0.5569 1,1285 1.0952:.1.2331 0.9179 1.1505 1.0327 18 19 -21 22 23 24 25 0.3727 1.1297 0.8610 1.2673 0.9427 1.1489 0.9677, 1.1064 26 27 28 29 30 31 32 '3

-.9152 1 0977 1,;2687: 1.:0776 -'1.3010 1.0702 1.0820 0.6553 34 35 36 37 38 39 40 41 0.3155 0.'9118 '.2357 0.9443 '1.3022 1.1727 1.3342 0.9544 1.1936 43 45 46 48 50 51 0.7379, 1".0153 0.9204 1.1514 1.0726 1.3350 1.1860 1.3629 1.3037 52 53 .54 56 57 58 59 60 0.6704 0.9438 1.1539 0.9704 1.0841 0.9556 1.3634 1.3140 1.2919 61 62 ,63 64 65 66 67 68 69 0.6123 1.2077 1.0360 1.1092 0.6566 1.1955 1.3077 1.2916 0.9851 Assembly Relative Power Densities at Hot .Ful.l Power with Eq. Xe BOC3 with Bank 5 and PLCEAs Inserted 5-14

0

+ <<.c>

\

~ kI" og,

.Figure 5-9 PALO VERDE U2C3 F -1.555 BQ 21 20.:7766

.,0.'. 3629 ..;0:. 7.121: .'.'0.,6650,'0 4 8 t'8 IN

'5 0.40'44 0.9194 0.9300 1.0126 0.9508 1.2308 12 '3 15 16 17

"'0.5783 1' 0826 ','1'. 0903 1.4182, 1', 0506 1.:3126 1. 0324

,18 19 20 21 22 23 2'4 25 0.4046 1.0828 0.9023'.4443 1.0287 1.3538 0.9874 '.1550 26 27 28 29 31 '2 33

' 0'192'1'.'09 1'0' 1':"4'4'43.' '1';.'1 717'1..3504 1.0294 1.1393 '0.5890 34 35 36 37 38 39  : 40 42

'0".3628 0.'9295 :1.4176 ."'1 ..0285;.1".'.'3503 1.'0507 1. 2451: .'. 8373,0.'9131 43 '4 45 46 47 '9 50 51 0.7759 :1'. 0119 1 '500 1.3532 1.0299 1'. 2449 0.9915 1.1971'.9578 52 53 55 '6 60 0.7116 0.9503 1.3115 0.9870 1.1389 0.8372 1.1966 0.9564 0.8927

61 63 64 65 66 ,67 68 69

,0 '660 1.2300 .1.0314 1.1541 0 '888, 0.9130 0.'9598 0.8921 0'. 7637

" -Assembly"Relative 'Power-Densities at Full Power EOC3 wi,th: Bank 5 Inserted-5-15

oc w f

~-

Figure '5-10 PALO VERDE U2C3 F 1.466'Q 32

.0. 3343 '0.;7244..'0 6781 0.6410, 8 10, 0.3697 0.8405 0.8480 0.9269 0.9104 1.2033 12 13 14 15 1'6 ,17 0.5277 ,0.9869 0.9946 1.2867 '0.8921 1,.2898 1.0574 18 19 20 21 22 23 24 25 0'.3698 0.9871 0.8157 1.2991 0.9591'.3265 1.0686 1.3538 26 27 28 29 30 31 32 33

!.':.0': 8403':0 9952 '.";1', 2991( ',"'0;; 97.41;  ;.1%. 2925 1-.'0774 1. 3691: 1'. 0724 35 36 37 '8 39 40 42 I

,;0. 3342...0. 8475 .',; 2861 .:0.'9590 '." 2924 1,.3458 41'.0658 0;9799 1.1478 45 46 47 48 49 50 51 43';7238 0'.9263, 0.8916 1.3258 '..'0779 1.'3455 '..0961 1.3538 1.0971'2

'3 54 56 57 58 59 60 0.6776 0.9098 1.2887 1.0682 1'. 3684 0.9797 1.3533 1.0611 0.9738 61 62 63 65 66 67 68, 69 0.6419 1.2024 1.0564. 1.3528 1.0719 1.1476 1.0994 0.9731 '.7445,

. Assembly Relative:Power,:Densities. at Full Power EOC3 with PLCEAs inserted 5'-16

0

~

M~

i 4

4

<5~

Figure 5-11 PALO VERDE UZC3 F 1. 519 BQ 21

..-, ."'0.'3616 "0'. 7,7.57 "',0,.'7.1'78,.'.0.':6742.,

5 8 10 1

0.4086 '0.9204 0.9147 0.9814 0.9454 1.2408 12 '3 15 16 17

-,0. 5854 ..'1.'0881 - 1'.'0826 .':1'. 3732 .0..'9219 1..2875 '1.0385, 18 19 20 21 22 23 24 25 0'. 4087 1.0883,. 0.8917 1.3981, 1,;0037 1.3253 0.9927 1.1767

, 26 27 28 29 30 31 33

'; 0':9202 1. 0833 '.1".'3980 "1":0278 '.3235 ',.0471'21.1798 0.6140 34 35 36 37 38 39 '40 41 42

','0.3614 0.'9142, ,'1. 3726, .1,:0035 .,1..'3234.. 1.. 0660 ,.1;. 2998 --0.8875,, 0,.9735, 43 44 45 46 47 '8 49 50 51 0.7751.,'0.9808 0.9214 '.3247 1.0476 1.2996 1.0538 1.2825 1.0273 52 .53 54 56 57 58 , 59 .60 0.7173 0.9449 1.2865, 0.9923 1.1794 0.8874 1.2820 1.0192 0'.9373 61 '2 , 63 66'7 68 69 0.6753.. 1.2400 1.0376 1.1758 0.6137 0.9734 1.0294 0.9366 0.;7210

'ssembly'elative Power 'Densities at Full Power EOC3 with Bank 5 and PLCEAs Inserted

'5-17

0' g kit I'

0 i

6.0 THERMAL-HYDRAULIC DESIGN

6. 1 DNBR ANALYSIS Steady state DNBR analyses .of;Cycle:3';at the rated'ower level.:of "

3800. MWT. have,.been. performed. using".the .'TORC-computer;.',code'"described:

in,'Reference"6-.1.,- the.'.CE-'.1"."criti.cal"heat>',fl,ux.'.~correlat'ion.'-.:described .,"

,in..References 6-2 .and 6-8, and the CETOP code described in Reference 6-3.

Table 1 contains a list of pertinent thermal-hydraulic, design

~ .

~

"..<parameters..'.The.'Modi:fied'~Statistical=.'Combi'nat'ion:of Uncertainties (MSCU) methodol'ogy presented in .Reference 6-4 was applied. with: Palo Verde-2 specific data using the calculational factors listed in

.Table 6-1 and other, uncertainty factors to define overall uncertainty penalty 'factors to"be,'applied 'in the DNBR calculations

.- .:;,performed..iby"the'.lCore-.Protection<Calculators (CPC) and Core Operating Limit Supervisory System (COLSS)-which,-when:used,.wi:.th the Cycle 3 DNBR limit of 1..24,'.provide assurance at the 95/95 confidence/probabi.l;i.ty l,evel"that,.the hot,rod,:wil:.1 not,,experi'ence..; "

DNB. The 1.24 DNBR li~it was cal'culated using the methodology of "Reference 6-5 as,'was done for. the Reference Cycle.

This Cycl'e 3 DNBR limit includes the following, allowances:

1. NRC imposed 0..01 DNBR penalty .for,HID-1 grids as discussed in Reference 6-6.

2. 'Rod bow penalty as discussed in Section. 6.2 below.

Other penalties imposed by NRC in the course of their review o'f the Cycle 1 Statistical Combination of Uncertainties (SCU) analysis

-'discussed'n Reference 6-5 (i..e..; TORC, code uncertainty and CE-1 'CHF correlation cross validation uncertainty,, as discussed. in Reference 6-6) are incl'uded in the overall uncertainty penalty factors derived in the Cycle 3 MSCU analysis.

6-1

0

~ i ggi

6. 2 . FF CTS OF FUE ROD BOWING'N DNBR MARG IN Effects of fuel rod bowing on DNBR .margin have been incorporated in the .safety and setpoint analyses in the .manner discussed in Reference 6-7. The penalty used -for this analysis, 1.75/ MDNBR,, is

'"val'id for, bundle burnups.up to,30,'000.MWD/MTU.;..Th'is '.penalty.'i,s ...

included in the 1.24 DNBR limit.

':For assembl'ies with"burnup greater than 30-'GWD/T sufficient available margin exists to offset rod bow penalties due to the lower radial power peaks in these higher burnup batches. Hence the rod bow penalty based upon Reference 6-7 for 30 GWD/T is applicable for,

'.al.l .assembly burnups expected, for Cycle 3.

'6-2

0 Cp

\+

TABLE 6-1 PVNGS-2 Cycle 3 Thermal H draulic Parameters at Full Power

'Reference

~

General Characteristics Units ~Cele 3.

Total Heat Output (Core only) HW$ .3800 3800, 10 Btu/hr 12,970 12,970

.Fraction of Heat Generated in 0.975 Fuel Rod 0;975'rimary System Pressure psia 2250 2250 Nominal Inlet Temperature (Nominal) F 565.0 565.0 Total Reactor Coolant Flow gpss 423,300 423,300 (Minimum Steady'State)- 10 1 b/hr 155.8 155.8 Coolant:Flow Through Core (Minimum) 10 1 b/hr 151.1 151. 1 Hydraulic Diameter (Nominal: Channel) ft 0.039 0. 039 Average:Nass Velocity 10 'b/hr-ft 2.49 2.49 Pressure Drop Across Core (Minimum Psl 14.5 14.5 steady state flow irreversible

.hP over entire fuel assembly)

Total Pressure Drop"Across;"Vessel 51.3 ,, .

.(Based on .nominal dimensions and minimum steady state, flow) 'Psl'TU/hr- -;51..3'85,100"*

Core Average Heat Flux (Accounts ft '186,600*

for fraction of heat generated in fuel rod and axial densi.fica-tion factor)

Total Heat Transfer Area .(Accounts 67,700* 68,300**

for axial densification factor)

Film Coefficient at Average . BTU/hr-ft F 6100 6100 Conditions Average Film Temperature Difference F 31 30 Average Linear Heat Rate of Unden- kw/ft 5.5 5.4 sified Fuel Rod (Accounts for

'fraction"of heat generated in fuel rod) 6-3

0 f

4 i

ae >

ili

0 TABLE 6-1 (continued)

Reference General Char teristics Units ~Cc1'e ~Cele 3 Average Core Enthalpy Rise BTU/lb 85.9 . 85.9 Maximum Clad Surface Temperature 0F 656 656 Engineering Heat Flux Factor l. 03+ 1.03+

Engineer,ing,Factor on.Hot .Channel.. .=1.03+:.= - - . '1;03+-..:.-.

Heat Input Rod Pitch, Bowing and Clad Diameter 1.05+ 1.05+

Factor Fuel Densification Factor (Axial) 1.002 1.002 NOTES:

• Based on 2576 poison rods.

• • Based on 2144 poison rods.

+ -

These factors have"been combined statistically -with other uncertainty factors as, described .in .Reference- 6-4 to define overall uncertainty

-.penalty factors to. be-applied in the DNBR calculations. in. COLSS,,andCPC-.- >>

which, when used in conjunction with the appropriate DNBR limit for that cycl'e provide assurance at the 95/95 confidence/probability level that the hot rod will not experience DNB.

Tech. Spec. minimum flow rate.

6-4

0 Nl zr 4 t

7.0 ON- OCA SAFETY ANALYSIS 7.0.1 ntroduction This section presents the results of the Palo Verde Nuclear Generating Station Uni.t.,2 (PVNGS-2), Cycle 3.Non-LOCA'afety,

.analyses at 3800 HWt.

The Design Basis Events (DBEs) considered in the safety analyses are listed in Table 7.0-1. These events are categorized into three groups: Hoderate Frequency, Infrequent, and Limiting Fault events.

, =,For.'the..purpose of -this report, "the Hoderate Frequency and Infrequent, Events will,be termed..Anticipated Operational Occurrences. The DBEs were evaluated with respect to four criteria:

,Offsite .Dose, Reactor Coolant System (RCS) Pressure, Fuel Performance (DNBR and Centerline Helt SAFDLs), and Loss, of Shutdown Hargin. Tables:7,.0-2 through,7...0-5. present the lists of events analyzed for each criterion. ,All events were .re-evaluated to.,assure that they meet their respective criteria for Cycle 3. The DBEs chosen for analysis, for-,each cri.terion. are, the .l.imiting events with, respect to that criterion.

7.0.2 Hethods of Anal sis The analytical methodology used for PVNGS-2 Cycle 3 is the same as the Cycle 2 (Reference Cycle) methodology (References 7-1, 7-2 and

'7-9) unless otherwise stated in the event presentations. Only methodology that has previously been reviewed and- approved on the PVNGS dockets (References 7-10 and 7-11),. the CESSAR docket (Reference 7-2), or on other dockets is used.

0

<<l a

A'A ik~

7.'0.3 Mathematical Models

.The mathematical models and computer codes, used in .the Cycle 3

. Non-LOCA safety analysis are the same .as those used. in the Reference Cycle analysis (References .7-1, 7.-2,and .7,-9) . -.-,Plant, response for ..

Non-LOCA Events was simulated using the CESEC III computer code

...(Reference 7;3)...,.Simulation,"of the-..fl.u'id, condi.tions:wi;thin the:hot

. channel. of .the reactor core and calculation of DNBR was performed using the CETOP-D computer code described in Reference 7-4.

The TORC computer code was used to simulate the fluid conditions within the reactor core and to calculate fuel pin DNBR for the RCP Shaft Seizure and Sheared Shaft event. The TORC code is described in References 7-6 and 7-7.

The number of fuel pins predicted. to experience clad failure is

..-taken, as the number of-,pi.ns,whi.ch have a CE-1 DNBR value below 1.24.

The only exceptions are the.,Shaft .Seizure and Sheared Shaft, events for which the statistical convolution method, described in

.,Reference..7-8,,:was,used. 'Reference 7-8 has been .approved .by- the.NRC and has been used in References 7-1, 7-2 and 7-9.

The HERMITE computer code (Reference 7-5) was used to simulate the reactor core for analyses which required more spatial detail than. is provided by a point kinetics model. Reference 7-5 has been approved by the NRC and has been used in References 7-1, 7-2 and 7-9.

HERMITE was also used to generate input to the CESEC point. kinetics model by partially crediting space-time effects so that the CESEC calculation of core power during a reactor scram is conservative relative to HERMITE.

7-2

iO

'l il

7.0..'4 n ut Parameters and. Anal sis Assum tions

,'Tabl.e. 7..0-6, summarizes the core parameters'assumed in the Cycle 3

,transient. analys,is,.and..:compares.,them..to,.the.'.val,ues..used .in ..the

.Reference Cycle., Spec'ific..ini,ti.al., condi:tions. for-.each:.event.;are tabulated in:-the.,section; of.',the, report,:summarizing,,that,:-event. -Tech

" "Spec changes,-.are"desc'ribed;-:in,.Section".'10;."~",.The~effects "of':.these';,'~~..

,changes.,were"considered for each DBE and were included as appropri ate. For some of the DBEs presented, certain initial core parameters were assumed to,be .more limiting .than the actual I calcula'ted Cycle 3 values. Such assumptions resulted in more "r~'adverse~consequences;..;:Events,.which...have",credited'PC .trip

.protection have assumed'nstrument channel'esponse times which are conservative:rel.ative to the Cycle 3 Technical Specifications.

7.0.5

~ ~ Conclusion

. All',DBEs have been: evaluated for .PVNGS-2, Cycle .3 to,determine

-'- 'whether their resul'ts'are"bounded by the Reference Cycle.

7-3

0

,r,~+

0 P+ty ll i~

Table 7.0-1 PVNGS Unit 2 Desi n Basis Events Considered in the C cle 3 Safet Anal sis 7.1 .Increase in Heat Removal by the:Secondary,System 7.1.1 Decrease in .Feedwater Temperature 7..1,. 2 Increase in 'Feedwater Flow 7.1.3 Increased Main, Steam Flow 7.1.4 Inadvertent Opening of a Steam. Generator Safety Valve or Atmospheric Dump Valve 7.1.5* Steam System Piping Failures 7.2 Decrease in Heat Removal by the Secondary System 7.2.1 'Loss of External Load 7.2.2 Turbine Trip 7.2.3 .Loss. of Condenser, Vacuum 7.2.4 Loss of Normal AC Power

'.5 Loss of Normal Feedwater

7 7.2.6* Feedwater System Pipe Breaks

, Decrease i,n..Reactor,,Cool,ant .Fl.owrate 7.3.1 Total Loss of Forced Reactor Coolant Flow 7.3,.2* Single. Reactor Coolant'Pump Shaft Seizure/Sheared Shaft 7.4 Reactivity, and Power .Distribution 'Anomalies 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition 7 '.2 Uncontrolled CEA Withdrawal at Power 7 '.3 CEA Misoperation Events 7.4.4 CVCS Malfunction ( Inadvertent Boron Dilution) 7.4.5 Startup of an Inactive Reactor Coolant System Pump 7.4.6* Control Element Assembly Ejection 7.5 Increase in Reactor Coolant System 'Inventory 7.5.1 CVCS Malfunction 7.5.2 Inadvertent Operation of the ECCS During, Power Operation

• Categorized as Limiting Fault Events 7"4

0 0

I

Table 7.0'-1 (continued) 7,. 6 ,Decrease .in,,Reactor Coolant System Inven'tory

,7 . 6.1 Pressurizer, Pressure .Decrease .Events

.7 . 6 . 2* Small. Pri mary, .L i ne ,Break..Outs,i de Containment

,7.. 6 . 3~ =.Steam Generator Tube, Rupture 7.7 Miscellaneous

,7 . 7.. 1 Asymmetric .Steam Generator Events

• Categorized's Limiting Fault Events 7-5

0 Q/

4r Sy

>a ~~*

0

Table 7.0-2

.DBEs Evaluated with Res ect to Offsite Dose Criterion Section Event Results A) Anticipated Operational 'Occurrences

7. 1. 4' 1) - Inadvertent.'opening':;of-.a".Steam," " " .'".Bounded,"by."--":"'.

e,.Generator-Safety-Val.ve"or-Atmospher'ic. -"'Reference Cycle'-"-

Dump Valve 7.2.4 2) Loss of Normal AC Power Bounded by Reference Cycle B) Limiting Fault Events l)~ -Steam System Piping Failures: Bounded by Reference Cycle 7.1.5a a) Pre-Trip, Power Excursions 7.1.5b b) Post-Trip Return-to-Power 7.2 '

~ ~ 2) Feedwater System Pipe Breaks Bounded by Reference Cycle 7.3.2 3) Single Reactor Coolant Pump Presented Shaft Seizure/Sheared Shaft 7.. 4 . 6 4) - Control Element..Assembly. Ejection -. Bounded by Reference Cycle 7.6.2 5) Small Primary Line Break Outs i de Bounded by Containment Reference Cycle 7.6.3 6) Steam Generator Tube Rupture Bounded by Reference Cycle 7-6

I 0

J4.,

C

~ i

&A 4

~V

~%a 0

Table 7.0-3 DBEs Evaluated with Res ect to RCS Pressure Criterion Section Event Results A) Anticipated .Operational,,Occurrences 7.2.1 1),",Loss;of ;External:.:Load -.Bounded "by -;=...

Reference Cycle 7.2.2 2) Turbine Trip Bounded by Reference Cycle 7.2.3 3) Loss of Condenser Vacuum Bounded by Reference Cycle 7.2.4 4.) .Loss of Normal AC Power Bounded by Reference Cycle 7.2.5 5) Loss of Normal Feedwater Bounded by Reference Cycle V.,4.> ':6)"--Uncontrol-led, CEA-Wi:thdrawal from Bounded by Subcritical or Low Power .Condition Reference Cycle 7.4.2 7) Uncontrolled CEA Withdrawal at Power Bounded by Reference Cycle 7,.5,1 .8) CVCS Halfunction Bounded by Reference Cycle 7.5.2 9) Inadvertent Operation of the Bounded by ECCS During Power Operation Reference Cycle B) Limiting Fault Events 7.2.6 1) Feedwater System Pipe Breaks Bounded by Reference Cycle 7.4.6 2) Control Element Assembly Ejection Bounded by Reference Cycle 7-7

~'

.~

wr cg 41WA'

Table 7.0-4 DBEs Evaluated with Res ect to Fuel Performance Section Event Results

,A) Anticipated Operational, Occurrences

,7 ..1. 1 1) .Decrease:.in;:,Feedwater.".Temperature " =-

-.;Bounded';by Reference Cycle F 1.2 2) Increase in Feedwater flow Bounded by Reference Cycle 7.1.3 3) Increased Nain Steam Flow Bounded by Reference Cycle 7.1.4 4) Inadvertent Opening of a Steam Presented

• Generator Safety Valve or

'Atmospheric Dump Valve

6) Total Loss of Forced Reactor Bounded by Coolant Flow Reference Cycle 7.4.1 7) Uncontrolled CEA,,Withdrawal from a Bounded"by-Subcritical or Low Power Condition -Reference Cycle 7.4.2 8) -Uncontrol"led CEA Withdrawal Bounded-by at Power Reference Cycle 7.4.3 9); CEA Nisoperation Events Bounded by Reference Cycle 7.6.1 10) Pressurizer Pressure Decrease Bounded by Events Reference Cycle 7.7.1 11). Asymmetric Steam Generator Events Bounded by Reference Cycle B) Limiting Fault Events

1) Steam System Piping Failures: Bounded by Reference Cycle 7.1.5a a) Pre-Trip Power Excursions 7.1.5b b) Post-Trip Return to Power 7-8

il 0

Table 7.0-4 (continued):

Section Event Resul.ts 7'.3. 2 2) ,Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft Presented'Bounded 7'.4.6 3) Control Element,*Assembly'.Ejection, by-Reference Cycl,e.

• The Base Case is bounded'y Reference Cycle. -

Results of the Event with Loss of Offsite Power is presented.

7-9

0 C

4 il Il

Table 7.0-5 DBEs Evaluated with Res ect to Shutdown Mar in Criterion Section Event Results A) Anticipated Operational'Occurrences

7. 1.4 .

1");..'Inadvertent.;Opening'.of"a;"Steam

" .Generator Safety Valve or

' '" "Bounded: by ""=

Reference Cycle Atmospheric Dump Valve 7.4.4 2) CVCS Hal function (Inadvertent Bounded by Boron Dilution) Reference Cycle 7.4.5 3) Startup of an Inactive Reactor Bounded by Coolant System 'Pump Reference Cycle

.B), Limi,ting..Faul,t Events

1) Steam System Piping Failures: Bounded by Reference Cycle 7..1.5b

~ ~ a') .Post-Trip;Return-to-Power 7-10

ll I

U I

Table 7.0-6 PVNGS Unit 2 C cle 3 Core Parameters In ut to Safet Anal ses Reference Cycle Safet Parameters Units ~C1 11 Total.'RCS" Power MWt 3898 '"3898 (Core Thermal 'Power

+ Pump Heat)

Core Inlet Steady State F 560 to 570 560 to 570 Temperature (90/ power and (90/ power and above) above) 550 to 572 550 to 572 (below 90% power) (below 90% power)

Steady, State ,psia 2000 - 2325 2000 - 2325 RCS Pressure Minimum Guar anteed gpm 423,320 423,320 Del ivered Vol umetri c Flow Rate Axial Shape Index LCO ASI -0.3 to +0.3 -0.3 to +0.3 Band Assumed for Units All Powers Maximum CEA Insertion  % Insertion 28 28 at Full Power of Lead Bank

% Insertion 25 25 of Part-Length Maximum Initial Linear KW/ft 13.5 13. 5 Heat Rate Steady State Linear KW/ft 21.0 21.0 Heat Rate for Fuel Center Line Melt CEA Drop Time from sec 4.0 4.0 Removal of Power to Holding Coils to 90%

Insertion N

'7-11

J 4l 0

Table 7.0-6 (continued)

,Reference Cycle Safet Parameters Units Hinimum 'DNBR CE-1'SAFDL) 1.24 .l.. 24 Hacbeth (Fuel failure 1.30 1.30 1;imit for, post-trip SLB. wi,th 'LOAC-References 7-5 and 7-6)

Initial Hoderator 10 hp/ F . Figure 7.0-1 Figure .7.0-1 Temperature Coefficient Shutdown Hargin (Value '%hp -6.5 -6.5 Assumed in Limiting Hot 2ero'Power- SLB) 7-,12

0

-jgt

0 '

7 1 "'INCREASE IN"HEAT REMOVAL""'BY"THE SECONDARY SYSTEM 7.1.1 . Decrease in Feedwater Tem erature The results are bounded, by the Reference .Cycle.

7. 1.'2 Increase in Feedwater Flow The results are bounded by the Reference Cycle.

7.1.3 Increased Main Steam Flow The results .are bounded by the'eference Cycle.

7. 1.4 Inadvertent 'Ooenin of a Steam Generator Safet Valve or Atmos heric The amount of predicted failed.fuel:has increased for the '

inadvertent.,Opening of a Steam Generator, Atmospheric Dump Valve wi.th

- Loss of Offsite Power after, Turbine Trip '( IOSGADV+LOP) from:.8%. to

.12%. The increase in failed fuel was the result of more adverse

-nuclear power distributions.. All other refer ence cycle data related to the IOSGADV remain applicable.

7-13

It tt 0

7. 1 ~

'5 Steam 'S stem Pi in Failures"

7. 1.5a Steam S stem Pi in Failures: Inside and Outside Containment Pre-Tri Power Excursions The results are bounded..by. the, Reference Cycle.

7. 1.5b Steam S stem Pi in Failures: Post-Tri Return to Power The results are bound'ed by the Reference Cycle 7.2 " '.'DECREASE IN'EAT 'REMOVAL 'BY THE SECONDARY SYSTEM 7.2.1 Loss of External Load' The resul.ts are bounded, by the Reference Cycle..

.2.2 ~bi

~

The r esul ts are, bounded "by"the,Reference .Cycl e.

7.2.3 L'oss of Condenser Vacuum The results are bounded by the Reference Cycle.

7.2.4 Loss of Normal AC Power The results are bounded by the Reference Cycle.

7-14

I 0

I Agan It ib

~AC ql pVtt'<

4V ll l

7.2.5 Loss of Normal Feedwater The results are bounded by the Refer ence Cycle.

7.2.6 Feedwater S stem Pi e Breaks

,The results are bounded 'by.'the .Reference.:Cyc~le.

7.3 DECREASE IN REACTOR COOLANT FLOWRATE 7.3. 1 Loss of Forced Reactor Coolant The results are bounded by the Reference Cycle.

Sin le Reactor Coolant Pump Shaft Seizure/Sheared Shaft The amount of predi.cted 'failed. fuel. has increased for the Single Reactor Coolant" Pump Shaft 'Seizure/Sheared Shaft from 3.79% to'4.5%%d.

As with section 7.1.4, the increase in failed fuel was the result of =

.-mor e adverse nucl ear power distributions.

The resul tant

.-.radiological consequences are a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site boundary thyroid dose of less than 240 Rem. This is within 10CFR100 guidelines.

7 4

~ REACTIVITY AND POWER DISTRIBUTION ANOMALIES 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition The results are bounded by the Reference Cycle.

7-15

I[ l h

0

'.4.2 Uncontrolled CEA Withdrawal at Power The results are bounded by the Reference Cycle.

7.4.3 CEA Miso eration Event

- The results are bounded by the Reference Cycle.

7.4.4 CVCS MALFUNCTION INADVERTENT BORON DILUTION)

The results are bounded by the Reference Cycle.

7.4.5 'Startup of .an Inactive Reactor Coolant Pum 'Event The results are bounded by the. Reference Cycle.

7.4.6 Control Element Assembl'E ection*

The. resul,ts,,are, bounded, by. the Reference Cycle.

7.5 , ; ., INCREASE IN REACTOR COOLANT SYSTEM INVENTORY 7.5.1 CVCS Malfunction The results are bounded by the Reference Cycle.

7.5.2 Inadvertent 0 eration of the ECCS Durin Power Operation The results are bounded by the Reference Cycle.

7-16

Il 0

F 6 DECREASE IN'EACTOR COOL'ANT SYSTEM INVENTORY 7.6.1 Pressurizer .Pressure Decrease Events The results are, bounded, by "the, Reference .Cycle.

7.6.2 -- Small Primar Line Pi e Break Outside Containment The results are bounded by the Reference Cycle.

7..6.3. . -Steam Generator. Tube Ru ture The results'-are bounded'by the 'Reference Cycle.

MISCELLANEOUS 7.7. 1's mmetric Steam Generator Events'The resul.ts are bounded 'by the-Reference Cycle.

7-17

4I~

il~

0

FIGURE RLLOIIIJRBLE MTC MODES 1 RND 2 LL O

0.5 O (0%,0.5) (100%,0.0) 0.0

-0.5

-1.0

-1.5 RLLOWRBLE MTC

-2.0

-2.5 (0%,-2 8)

-3.0 ( 100%, -3.

-3.5 0 l0 20 30 40 50 60 70 80 90 100 CORE POMER LEVEL, % OF RRTED THERHRL POWER

~'

Al Cy 8 P

4 0

8.0 CCS ANALYSIS 8.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT 8.1.1 Introduction And Summary An ECCS performance analysis of the limiting break size was performed f'r PVNGS-2 Cycle 3 to demonstrate compliance with 10CFR50.46 wnich presents the HRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled reactors (Reference 8-1). The,ana'lysis justifies. an allowable Peak Linear Heat Generation Rate (PLHGR) of 13.5 kw/ft. The method of analysis and

-detailed results which. support this value are presented herein.-

8.1.2 Hethod Of Analysis The ECCS performance analysis for PYHGS-2 Cycle 3 consisted of an evalua.ion of the differences between Cycle 3 and PVNGS-2 Cycle 1.

, .For, this reason PVNGS-2, Cycle 1 shall be referred to as the Reference Cycle in Section 8: Acceptable ECCS performance was demonstrated for the Reference, Cycle in Reference 8-2 and approved by the NRC in Reference 8-3. As in the Reference Cycle, the calculations performed for this evaluation used the HRC approved C-E large break ECCS performance evaluation model which is described in Reference 8-4 incl.uding the use of a more conservative axial power shape. The blowdown hydraulic calculations;. refill/reflood hydraulics calculations, and, steam cooling heat transfer coefficients of the Reference Cycle apply to PVNGS-2 Cycle 3 since there have been no significant adverse changes to RCS or ECCS hardware characteristics, or to core and system parameters.

Therefore, only fuel rod clad temperature and oxidation calculations are required to re-evaluate .ECCS performance with respect to the changes in fuel conditions introduced by Cycle 3. The HRC approved STRIKIN-II (Reference 8-5) code was used for this purpose.

8-1

0 l5 fV-<

Jb il

Burnup dependent calculations were performed with STRIKIN-II to determine the limiting conditions for the ECCS .performance analysis.

The fuel performance data was generated with the FATES-3A fuel evaluation model .(References,'8-6 and:,8-7),wi.th,the NRC grain size.

restriction (Reference 8-8). It was demonstrated that the burnup

,.with the h.ighest initial .fuel; stored:energy.,was l.,imiting...This ,-.-

.,occurred: at a..low burnup for. the hot.,rod.

The Unit 2 Cycle 3 analysis considered a reduction of 470 gpm in LPSI runout flow relative to the Reference Cycle. The evaluation confirmed that there is adequate safety injection flow to maintain a

'ful.l downcomer with the reduced flow. Therefore, this reduction in

.LPSI,flow,will,,not affect the .results.

The acceptable performance of Unit 2 Cycle 3 has also been confirmed ";

with up .to 400 plugged tubes per. steam generator and with a reduction in system flow rate to 155'.8 X 10 ibm/hr and a reduction in core flow rate to 151..1,X, 10 6 .ibm/hr,. For the Unit 2.,Cycle, 3, analysis nuclear flux augmentation factors were set to unity. The-allowable 'PLHGR of= 13.'5= kw/ft for Cycle 3 is a reduction of 0 5-kw/ft from the Reference Cycle.

The temperature and oxidation calculations were performed for the 1.0 Double-Ended Guillotine at Pump Discharge (DEG/PD) break. This break size is the limiting break size of the Reference Cycle and, as the hydraulics are identical,, is the limiting break size for Cycle 3.

8. 1.3 Results The ECCS performance analysis for PVNGS-2 Cycle . showed that the reference analysis results conservatively apply. The peak clad temperature, maximum local clad oxidation, and core wide oxidation values of 2091'F, 9.0% and < 0.80%, respectively, for the reference analysis are below the corresponding 10CFR50.46 acceptance criteria of 2200'F, 17%, and 1%, respectively.

8-2

~O

~ 'l fp,.

AE JI 1'

Conclusion Conformance to the ECCS criteria .is demonstrated by the analysis resul.ts. Therefore, oper ation of PVNGS-,2 Cycle 3 at a core power level of 3876 MWt (102%%d of 3800 MWt) and a,PLHGR of 13.5 kw/ft, is in compliance with 10CFR50.46 ~

8.2 SMALL BREAK LOSS-OF-COOLANT 'ACCIDENT A review of Cycle 3 fuel and core data confirmed that the reported small break loss-of-coolant accident results (Reference 8-9) for PVNGS-2 Cycle 1 bounds PVNGS-2 Cycle 3. Therefore, acceptable small break LOCA,ECCS performance is demonstrated at a peak linear heat generation rate of 13.5 kw/ft and a reactor .power leVel of 3876 MWT

'(102%%d of 3800 MWT):. This acceptable performance has been confirmed '-

.with up to 400 plugged tubes per steam generator.

The reduction, in,del,ivered low, pressure safety injection flow (see Reference 8-10) does not impact the small break loss-of-coolant analysis. The fuel cladding temperature excursion is either

.,terminated by the high, pressure safety .injection pump flow, or, by the discharge of the safety injection tanks.

8-'3

0 E

0 0

9.0 REACTOR PROTECTION AND MONITORING SYSTEM

9.1 INTRODUCTION

The, Core. Protection Calculator, System (CPCS)...is .designed,to,provide.-

the low'.DNBR, and'high, Local, Power..',.Dens,i.ty',(,LPD)".trips;;to. (.1'),ensure "

.that the"speci.fied.-.,:acceptable .fuel..'design. limits on; departure 'from nucleate boiling and centerline fuel melting are not exceeded during Anticipated Operational Occurrences (AOOs) and (2) assist the Engineered Safety Features 'System in limiting the consequences of certain postulated accidents.

The., CPCS .in conjunction with. the remaining,.Reactor Protection, System

'(RPS) must be capable of providing protection for certain specified basis events, provided that at the initiation of these 'esign

.occurrences the Nuclear Steam Supply System, its subsystems, components and parameters are maintained within operating limits and Limiting, Conditions for Operation (LCOs).

9.2 'CPCS SOFTWARE MODIFICATIONS The algorithms associated with the CPC Improvement Program (References 9-1, 9-2 and 9-3) which were implemented in Cycle 2, are applicable to this cycle. The values for the Reload Data Block constants wi+1- be evaluated for applicability consistent with the cycle design, performance and safety analyses.'-"Any necessary change to the RDB constants wi.l.l be instal,led in:accordance with Reference.

9-4.

9-1

P 0

9.3 ADDRESSABLE CONSTANTS Certain CPC constants are addressable so that they can be changed as required during operation. Addressable constants include (I) constants that are measured. during startup .(e.g.,:shape anneal.ing,.

matrix, boundary point power correlation, coefficients, .and

,.-,adjustments for planar .radial.:.peaking..factors')., (2) uncertai.nty

, factors. to account for.:processing;;and,".measurement-uncertainties:i.n;.-

'DNBR'nd "LPD calculations (BERRO through BERR4), '(3) trip setpoints and (4) miscellaneous items (e.g., penalty factor multipliers, CEAC penalty factor time delay, pre.-trip setpoints, CEAC inoperable flag, calibration constants, etc.).

Trip setpoints, uncertainty'actors and other addr'essable constants

.will, be. determined for,thiscycle. consistent with the software and..

- methodology established in the CPC Improvement Program and the cycle design, performance and safety;analyses. As .for the Reference

Cycle,;uncertainty factors;wi;l,l .be determined using the modified

9. 4 ', statistical- combination"of uncertainties method (Reference 9-5).

DIG ITAL'MONITORING SYSTEM'OLSS

.'he Core Operating Limit Supervisory System (COLSS), described in Reference 9-6, is a monitoring system that initiates alarms if the LCO's. on DNBR, peak linear heat rate, axial shape index, core power, or core azimuthal tilt are exceeded. The COLSS data base and uncertainties wi.ll be updated, as required, to reflect 'the reload core design.

9-2

0 0

0

T CHNICAL SPECIFICATIONS

-This section provides a summary of the-proposed"changes to the

Technical'pecifications for-PVNGS-2=Cycle 3. The"changes are -~

arranged in numerical, order. Detai.led change, pages for-.the Technical Specifications:.are.presented:;:el,sewhere:,.

10-1

0

,)

0

Section Ti tl e Nature of Chan e

3. 1 ..1,. 2, - =

Shutdown; Margin Revise Shutdown-Margin .Requirements Figure 3. 1-. 1A T greater than ,consistent wi,th Cycle 3 ..analyses.

210'F

.3 ..l..2...7.-, ..Bor,onD.i,,l.ut,i,on, .'Revi'se 'Tables.'.3.,1-,2 :3=..1;:3, and .3.,1-5, on.

Tables 3.1-2, Alarms monitoring frequency with 1 or 2 boron

3. 1-3, and dilution alarms inoperable, due to 3.1-5 increased BOC critical boron concentrations for Cycle 3.

3,. 1'.3.,6 Regulating CEA Revise Figures 3.. 1-3 and 3..1-4 to reflect Figures 3. 1-3,, Insertion Limit Cycle 3 core characteristics.

3.1-4

3. 2:,4 DNBR Margin 'Revise -Figures 3.2-3 and 3.2-2A to reflect Figures 3.2-2 Cycle 3 'core characteristics.

and 3.2-2A 3.2.7 Axial Shape Index Change-'COLSS ASI limits to +0.27.

10-2

Cl 4

S 6 'j &-

0

11.0 STARTUP T STING

=-, The:planned,startup test program associated with core performance. is

.outlined below. The described"tests,.veri.fy"that;coreperformance =.is

. consistent,.with: the .engineering,-. design,,and -safety .analysis.. The:

.:,program .conforms::to.";ANSI/ANS=19':6..-1-::1985',.=="..Reload:-"Startup-=Physics

'-Tests, for, Pressurizedi."'Water,.....Reactors" --.'and..;supplements...:normal-surveillance tests which are required by Technical Specifications (i.e., CEA drop time testing, RCS fl,ow measurement, HTC verification, etc).

11. 1.  ;" LOW POWER 'PHYSICS TESTS 11.1.1 Ini ti al Cri ti cal i t Ini.tial criticality will- be achieved by, one of two methods. By the first method, all 'CEA" groups woul'd'e fully .withdrawn with the

.exception of the lead regulating group which. would be .posi.tioned at.

.approximately mid-.core. The boron. concentration of the reactor coolant-would.'then 'be"reduced unti'1"'critical-ity 'is "attained'..- By"the second, method, the boron concentration is adjusted to the expected

~

-'critical concentration -with the shutdown .and- Part-Length CEA,groups fully withdrawn. The regulating CEA groups would be withdrawn to achieve criticality.

11. 1-.2- Critical "Boron'Concentration CBC The CBC will .be .determined for the'nrodded configuration and for. a.

partial.ly rodded configuration. The measured CBC values will be verified to be within +1% hk/k of the predicted 'values.

11-1

k 0

11.1.3 Tem erature Reactivit Coefficient

. The isothermal temperature coefficient (ITC) wi.ll:be",measured at the Essentially All Rods Out, (EARO) configuration, and at a, partially rodded .configuration. The coolant temperature .wi.l.l,'be ,varied and:

the resulting .reactivi.ty,.change wi:l,l ,be.:measured..--.The.:measured'4 values wi;ll be verified, to. be:within +0.3 x 10' ak/k/ F-.of'he-.

pr edicted values.

11.1.4 CEA Reacti vi t Wor th CEA group worths will be measured using .the. CEA Exchange technique.

This technique consists of measuring the worth of a "Reference

'.Group" via standard boration/dilution techniques and then exchanging this group with other groups to measure their worths. Al.l full-length CEAs will be included in the measurement. Due to the large, differences in CEA -;group 'worths, two reference groups ,(one with high worth and'ne with medium. worth) may be used. The groups to .be measured ,will,. be. exchanged with the appropriate reference

,, group. Acceptance criteria .will .be as .specified .in::Reference .11-2.

11.1 ~ 5 Inverse Boron Worth IBW The IBW will be calculated using results from the CBC measurements and the CEA group worth measurements. The calculated IBW value wi'll be verified to be within +15 ppm/% ak/k of the .predicted value.'1.2 Power Ascension Testin Following completion of the Low Power Physics Test sequence, reactor power will be increased in accordance with normal operating

- procedures. The power ascension will be monitored through use of an off-l.ine 'NSSS performance and data processing computer algorithm.

This computer code will be executed in parallel with the power ascension to monitor CPC and COLSS performance relative to the 11-2

41 0

ta e-

processed .plant data againstl which they, are normally calibrated. If necessary., the power ascension will be suspended- while necessary data reduction and equipment calibrations are .performed. The

.:foll,owing measurements will be performed, during, the program.

11.2. 1 Flux S mmetr Verification

,....Core..power...distribution,,.as -determined...from fixed."incore=.detector data, will be examined prior to exceeding 30% power to verify that no detectable fuel misloadings exist. Differences between measured powers in symmetric,, instrumented assemblies .will be verified to be within 10% of the symmetric group average.

11.2.2 Core Power Distribution Core power distributions derived from the fixed ,incore neutron detectors wil.l be compared to predicted distributions at two power plateaus. -These ,comparisons,,serve .to further verify proper fuel loading and verify. consistency between the as-built core and the engineering design models. -'Compliance with the acceptance criteria

. 'at .the -intermediate -power .:plateau :(between: 40%,.:and -70% ..power)

. provides reasonable assurance that the power distribution will

'remain wi,thin the design limits while reactor power is increased to 100%, where the second comparison will be performed.

The measured results will be compared to the predicted values in the following ,manner for both the intermediate and ,the full power analyses:

A. The root-mean-square (RHS) of the difference between the measured and predicted relative power density (axially integrated) for each of the fuel assemblies will be verified to be less than or equal to 5%.

11-3

l 1r 0

kC ky

B. The RHS of the I difference between ,the. measured and

,,predicted core .average. axial power"..distribution for .:each

.axial node will be verified .to be: less than or .equal to

'5%.

C. The measured, values..of. planar radial peaking factor (Fxy),

integrated .radial .peaking .,factor,.(Fr),'ore. average;...axi,al;

...,peak (F .);.,and the:.3-.D:power peak. (Fq) wi:1:1 be veri,fied'o.

be, within +10% of their predicted values.

11.2.3 Sha e Annealin Hatrix SAH and Boundar Point Power Correlation Coefficients BPPCC Verification The SAM and BPPCC -values will be'determined from a'inear regression analysis of the measured excore detector readings and corresponding'ore power distribution determined from incore detector signals.

0 Since these values must be representative for a rodded and unrodded-

core throughout 'the, cycle, i.t.is .desirable to use as wide a range of axial shapes as is available to establish their values. The spectrum

of axial shapes encountered during the power ascension has

..; . been;.demonstrated ..to,'; be..adequate.,for.,the, calculation..of .,the;matrix" elements. The necessary data will be compi.led and analyzed through the power- ascension by the off-line NSSS performance and data processing algorithm. The results of the analysis will be used to modify the appropriate CPC constants, if necessary.

11.2.4 - Radial Peakin Factor RPF and CEA Shadowin Factor RSF Verification The RPF and RSF values will be determined using data collected, from the fixed incore detectors and the excore detectors. Values will be determined for unrodded as well as rodded (lead regulating group and part-length group only) operating conditions. .Appropriate CPC and/or

'OLSS constants wil.l be modified based upon the calculated values.

The rodded .portions of this measurement may be deleted from the test program if appropriate adjustments are made to CPC and COLSS constants.

11-4

4b d ~+Wl, "

S Q

II C

1 F

rc I'1.2.5 Tem erature Reactivitv Coefficients at Power P

The i'sothermal temperature coefficient. (ITC) will be measured at approximately full power. The ITC will" be- measured by changing-coolant temperature, compensat,ing-.wi;th .CEA',motion,. and,.maintaining power steady. The ITC. will,be- verified to be .within, +0.3. x 10 Lk/k/'F of the .predicted value.

II.2.6 Critical Boron Concentration The CBC will be determined for conditions of'ull .power, equilibrium

xenon. The ..measured CBC will be verifi,ed to be within -:50 ppm of

.the'.,predicted value after adjustment for the bias .observed between measured and predicted CBC values at zero power.

11.3 ,PROCEDURE IF ACCEPTANCE CRITERIA ARE NOT MET j The results of all tests wi,ll be reviewed by the plant's reactor engineering group. If the acceptance criteria of the startup

,physics, tests,.are,,not ,met, an '.evaluation ,will, be performed with assistance from the fuel vendor as needed.

11-5

!I E

f 4 . -,I q&'

12.0 R F RENC S 12.1 ,SECTION'. 0 REFERENCES (1-1) "Palo Verde Nuclear Generating Station-Unit No. 2, Final' Safety..'.Analys,is ':Report,""Arizona;..Publ.ic".;Servi.ce.:Company;,

Docket. No. 50-529.

(1-2) 161-00683-EEVB/LJH, "Palo Verde Nuclear Generating Station Unit No. 2 Docket No. STN 50-529, Submittal of the Reload Analysis Report for Unit 2 Cycle 2," December 3, 1987.

12".2, SECTION

2.0 REFERENCES

None 12.3j SECTION

3.0 REFERENCES

None 12.4 SECTION

4.0 REFERENCES

(4-1) 161-00730- EEVB/LJH, "Final Surveillance Test Results for PVNGS-1, Cycle 1," January 8, 1988.

(4-2) 161-01102-EEVB/PGN, "Fuel Survei1.l ance Test Results.. for PVNGS-2 Cycle 1," June 9, 1988.

(4-3) CENPD-139-P-A, "C-E Fuel Evaluation Hodel," July, 1974.

CEN-161(B) -P, "Improvements,to Fuel Evaluation Hodel, "

July, 1981.

12-1

~O "t ~

Ik 0

(4-5) R. A. Cl ark (NRC) to A. E. Lundval 1, Jr. (BGE E), "Safety

-Evaluation of CEN-,161 (FATES3)," March 31, 1983.

(4-6) 161-00453-,JGH/SGB, "Fuel Assembly Guide Tube Wear Program for PVNGS Unit 2," August 20, 1987.

"Palo Verde Nuclear"Generating Station Unit No. 1, 'inal Safety Analysis 'Report,'"';Arizona- Publ;ic Servi,ce.-. Company,.

Docket No. 50-528, Section 4.2.4.

(4-8) CESSAR SSER2, Section 4.2.5, "Guide Tube Wear Surveillance".

SECTION

5.0 REFERENCES

CENPD-153-P, Rev. 1-P-A, "INCA/CECOR Power Peaking

,Uncertainty," May, 1980.

CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Desi,gn," April, 1983.

SECTION.

6.0 REFERENCES

(6"1) CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core", April, 1986.

CENPD-162-A, ".Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," September, 1976.

161-01867-DBK/JRP, "Generic Applicability of the CETOP-D Code for 'PWR Core T-H Analysis," April'6, 1989.

(6-4) CEN-356(V)-P-A, Rev. 01-P-A, "Modified Statistical Combination of Uncertainties", May, 1988.

12-2

~ I C

~O

Enclosure 1-P to LD-82-054, "Statisti cal'ombination of System Par ameter Uncertainties, in Thermal: Margin Analyses for System 80", submitted by letter from,A. E. Scherer (C-E) to D. G. Eisenhut (NRC), May 14, 1982.

CESSAR .SSER 2 Section 4.4...6,.'Statisti ca'1.',Combination of Uncertainties.

'(6-7) CENPD-'225-P-A, '"Fuel and Poison Rod Bowing," June 1983.

CENPD-207-P-A, "Critical Heat Flux" Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-uniform Axial Power Distribution," December, 1984.

SECTION

7.0 REFERENCES

(7-1) "Palo Verde Nuclear Generating Station Unit No. 2, Final Safety Analysi's;Report,"'. Arizona Public Service Company, Docket No. 50-529.

.(7-2) ."CESSAR,. Combustion. Engineering Standard Safety,,Analysis.

Report," Docket No. 50-470.

(7-3) "CESEC, Digital .Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981, Enclosure 1-P to LD-82-001, January 6, 1982.

(7-4) 161-01867-DBK/JRP, "Generi c Appl i cabi,l i ty of the CETOP-D .

Code for PMR Core T-H Analysis," April 26, '1989.

(7-5) CENPD-188-A, "HERMITE Space-Time Kinetics," July, 1975.

(7-6) CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor 'Core," April, 1986.

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CENPD-206-P-A, "TORC Code Veri fi cati on and S imp 1 i fi ed Modeling Methods," June 1981.

(7-8) CENPD-183-A, "Loss of Flow - C-E Methods for Loss of Flow Analysis," June 1984.

(7-9) 161-00683-EEVB/LJM,.-"Palo"Verde Nuclear Generating 'Station (PVNGS) Unit .2 Docket,.No. STN, 50-.529, Submi.ttal..of the.

Reload Anal.ysi,s,Report for Unit 2 Cycle 2,"'ecember 3, 1987.

(7-10) USNRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 24 to Facility

'Operating License No. NPF-41, Arizona Public Service Company, et. al. Palo Verde Nuclear Generating Station, Unit No. 1 Docket No. STN50-528," October, 21, 1987.

USNRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 19 to Facility Operating License No. NPF-51, Arizona Public Service Company, et. al. Palo Verde Nuclear Generating Station,

>>.Unit No. 2 Docket No. STN50-529," .May .5,, 1988.

(7-12) Letter to D. G. Eisenhut from A. E. Scherer, Letter No.

LD-82-040, March 31, 1982.

SECTION

8.0 REFERENCES

-ECCS ANALYSIS (8-1) "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," Federal Register, Vol. 39, No. 3, Friday, January 4, 1974.

(8-2) ANPP-33584-EEVB/KLM, "Limiting Large Break LOCA Analysis Results - Chapter 15 Reanalyses", September 27, 1985.

ANPP-33650-EEVB/KLM, "Lar ge Break LOCA Evaluation Model Reanalysis Results", October 3, 1985.

12-4

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(8-3) PVNGS Safety Evaluation Report; NUREG-0857, Supplement 9, dated December, 1985, Section 6.3 (8-4) CEHP0-132-P, "Calculational Methods for the.C-E Large Break LOCA Evaluation Model",, August 1974.

CEHPD-132-:P, Supplement 1, '"Cal:culationalMethods for the

C-E Large. Break--LOCA .Evaluation. Model", February 1975.

CEHPD-132-P, Supplement 2-P, "Calculational Methods for the C-E Large Break LOCA Evaluation Model", July 1975.

'Letter 0. D. Parr (HRC) 'to F. M. Stern '(C-E), dated June 13, 1975 (HRC. Staff Review of the Combustion Engineering ECCS Evaluation Model).

Let.er 0. D. Parr (HRC) to A. E. Scherer (C-E) .dated December 9, 1975-'(HRC, Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model Changes).

(8-5) CEHPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1974.

CEHPD-135, Supplement 2P, "STRIKIH-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (modification)",

February 1975.

CEHPD-135-P,- Supplement 4P. "STRIKIN-II,,A Cyl.indrical Geometry Fuel Rod Heat Transfer Program", August 1976.

CEHPD-139-P-A, "C-E Fuel Evaluation Model", July, 1974.

(8-7) CEN-161(B)-P, "Improvements to Fuel Evaluation Model",

July, 1981.

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' Letter from R. A. Clark (NRC). to A. E. Lundvall, Jr ,

(BG&E), "Safety Evaluation of CEN-161 (FATES 3),"

March 31, 1983.

ANPP-33609-EEVB/KLM, dated. September 30; '1985, "Limiting-Small Break LOCA-Analysis '-:Additional Information".

(8-10) =.161-00890-EEVB/BJA; =,dated March-'1'6; '1988,""! Pr oposed'-

Technical Specifications Change - L'PSI Flow Requirements". =

161-01155-EEVB/BJA, dated July 6, 1988, "LPSI Flow Requirements".

Letter from the NRC, dated October 17, 1988, "Amendment to PVNGS-1, 2 & 3 Operating License for LPSI 'Flow".

12.9 ,SECTION 9..0 REFERENCES (9-1) CEN-304-P, Rev. 01-P, "Functional Desi'gn Requirement for a Control Element Assembly Calculator," May, 1986.

.(9-,2) .. CEN.-305-P, Rev.. 01-P, "Functional Design Requirement for a Core Protection Calculator," May, 1986.

(9-3) CEN-330-P-A, "CPC/CEAC Software Modifications for the CPC Improvement, Program Reload Data Block," October, 1987.

CEN-323-P-A, "Reload Data Block Constant Installation Guidelines," September, 1986.

(9-5) CEN-356(V)-P-A, Rev. 01-P-A, "Modified Statistical Combination of Uncertainties," May, 1988.

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(9-6) CEN-312-P, Rev. OI-P, -"Overview Description of the Core Operating Limit "Supervisory System .(COLSS)", November, 1986.

SECTION

10.0 REFERENCES

NONE SECTION

11.0 REFERENCES

(11-1) ANSI/ANS-19.6.1-1985, "Reload Startup, Physics, Tests for Pressurized Mater Reactors".

(11-2) CEN-319-A, "Control Rod Group- Exchange Technique," Qune 1986.

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