ML17304A977

Reload Analysis Rept for Palo Verde Nuclear Generating Station Unit 1 Cycle 3

ML17304A977
Person / Time
Site:	Palo Verde
Issue date:	01/18/1989
From:	ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
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ML17304A976	List: ML17304A975 ML17304A977
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NUDOCS 8901260085
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Text

ATTAQMNT UNIT I, CYCLZ 3 RELOAD ANALYSIS REPORT 85'Oi260085 890ii8

'DR AGOCK 05000528 P

PDC

RELOAD ANALYSIS REPORT FOR PALO VERDE NUCLEAR GENERATING STATION UNIT 1 CYCLE 3 TABLE OF CONTENTS PAGE 1.

INTRODUCTION AND

SUMMARY

j 2.

OPERATING HISTORY OF THE REFERENCE CYCLE 2-1 3.

GENERAL DESCRIPTION 3"1 4.

FUEL SYSTEM DESIGN 4-1 5.

NUCLEAR DESIGN, 6.

THERMAL"HYDRAULICDESIGN 5-1 6-1 7.

TRANSIENT ANALYSIS 8.

ECCS ANALYSIS 8"1 9.

REACTOR PROTECTION AND MONITORING SYSTEM 10.

TECHNICAL SPECIFICATIONS 9-1 10-1 11.

STARTUP TESTING 12.

REFERENCES 12"1

INTRODUCTION AND

SUMMARY

This report provides an evaluation of the design and performance of Palo Verde Nuclear Generating Station Unit 1 (PVNGS-1) during its third cycle of operation at 100'I rated core.power of 3800 NMt and NSSS power of 3822 NMt.

Operating conditions for Cycle 3 have been assumed to be consistent 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 304 and 356 EFPD (Effective Full Power Days).

The second cycle of operation will hereafter be r eferred to in this report as the "Reference Cycle."

The safety criteria (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 core 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 oc'currences 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.

1-1

For those transients requiring reanalysis (Type 2), analyses are presented in Sections 7 and 8 showing results that meet the established safety criteria.

The Technical Specification changes needed for Cycle 3 are summarized in Section 10 and described in greater detail in separate license amendment applications.

i I

e 2.0 OPERATING HISTORY OF THE REFERENCE CYCLE The Reference Cycle gegan with initial criticality on March 5, 1988.

Power Ascension began on March 9, 1988, and on March 19, 1988, the unit reached full power.

It is presently estimated that Cycle 2 will terminate on or about

March, 21 1989.

The Cycle 2 termination point can vary between 298 and 350 EFPD to accommodate the plant schedule and still be within the assumptions of the Cycle 3 analyses.

2-1

3.0 G NERAL DESCRIPTION The Cycle 3 core will consist of those assembly types and numbers listed in Table 3-1.

Ninety-seven Batch B assemblies and twelve Batch C will be removed from the Cycle 2 core to make way for 108

fresh, Batch E assemblies.

Fifty-two Batch C and all Batch D

assemblies now in the core will be retained.

One Batch B assembly discharged at EOC1 will be reinserted into the core.

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

The reload batch will consist of 24 type EO assemblies, 20 type El assemblies with 16 burnable poison shims per assembly, 12 type E2 assemblies with 16 burnable poison shims per assembly, 12 type E3 assemblies with 16 burnable poison shims per assembly, 32 type E4 assemblies with 16 burnable poison shims per assembly, and 8 type E5 assemblies with 4 burnable poison rods per assembly.

These sub-batch types are 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 beginning of Cycle 3 assembly average burnup distribution.

The burnup distribution is based on a Cycle 2 length

.of 330 EFPD.

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

TABLE 3-1 PALO VERDE NUCLEAR GENERATING STATION UNIT 1

Cycle 3 Core Loading Assembly Fuel Rods Desig-Number of per nation Assemblies Assembly Initial Enrichment (w/o U-235)

Number Shims/

Assembly Initial Shim Loading (gm B10/in)

Total Number of Fuel Shim Rods Rods C/

p*

Dx EO E1 E2 E3 E4 E5 TOTAL 36 16 36 28 12 20 12 12 32 241 208 12 224 12 208 12 184 52 172 52 216 12 216 12 184 52 168 52 168 52 168 52 168 52 180 52 2.78 1.92 3.30 2.78 3.30 2.78 4;05 3.36 3.36 2.78 3.36 2.78 I

3.36 2.78 4.03 3.90 4.03 3.90 3.90 3.60 3.90 3.60 3.90 3.60 4.03 3.90 16 16 16 16 16 16

.01842

.01151

.008

.008

.020

.024

.024

.026

.016

.012 208 12 8064 432 3328 192 6624 1872 4816 1456 2592 144 864 48 4416 1248 3360

'1040 2016 624 2016 624 5376 1664 1440 416 54892

~ 16 256 0

224 96 32 320 192 192 512 32 1872

1

FIGURE 3-1 B,

C 5 0 ASSEM L LOADINGS WATERHOLE AND SH CEMENT ASSKMSI.Y TYFK NUMSKR OI'SSEMS LIES fUEL ENRICHMENT W/Ta U224 No. OF FUEL ROOS FKR ASSKMbLY No. Of SHIM ROOS I ASSEMbLY yn b10 IIIL SUB.BATCHD. 38 ASSEMBLIES 0 4.05 wlo U-235 Q 335 wlo U-235 292 272 11 3N 0.01442 36 12 224 12 204 14 IL01141 SUBZATCH Do 28 ASSEMBLIES 0 338 w/o U-235 8 2.78 wlo U.235

~ g4C - AL2O3 SHIM PIN.008 0m 8-'IO/IN P

ANO SUB.BATCH DX-12 ASSEMBLIES 0 338 w/o U-235 Kb 2.78 wlo U.235 8 B4C - AL203 SHIM PIN.008 yn 8-10/IN Q

WATERHOLE S

LOWER ENRICHED FUKLFIN Q

HIGHER KNRICHKOI'UKL~IN Q

SHIM FIN SUB BATCHD/-4 ASSEMBLIES 0 338 w/o U-235 IN 2.78 w/o U-235

~ B4C - AL2O3 SHIM PIN.020 gm 8-10/IN

~

~,

t

FIGUR

.. -2 CURRENT CYCLE ASSEtNBLY FUEL LOADINGS VfATERHQLE AND BHIM PLACElVlENT 0

SUB-EATCH EO-24 ASSB8 IES RS-EATCH E3-l2 ASSENBLIES 0

4 03 v/o u-236 H

3.so v/0 u-235 0

3.00 V/0 U-235 8

3.60 V/0 U-235 B4C-AL203 SHIN PIN 026 gm B-IO/IN SUB-EAKH El-20 ASSMLIES SUB-EATCH E4-32 ASSBSLIES 0

4.03 V/o u-235 8

3.90 V/0 U-235

'4C-AL203 SHIH PIN 024 gnl B-,IO/IN 0

3.80 V/0 U-235 8

3 60 V/0 U-235 84C-AL203 SHIH PIN

-OI6 gm B-IO/IN RB-EAKH E2-l2 ASSEHOLIES 0

3.oo V/0 u-235 8

3 60 V/0 U-235 B4C-N 203 SHIN PIN 024 em 8-10/IN RS-EATN E5-8 8SSBSLIES 0

4.03 V/0 U-235 H

3.90 V/0 U-235 B4C-AL203 SHIH PIN 012 gm 8-10/IN

FIGURE 3-3 PVNGS UNli 0 CYCLE 3 FUEL MANAGEMENT C/

C/

E4 D

E4 EO D

C/

EO C

EO C

EO ES C

C C

C E4 D

El Dx D>>

D El D>>

E3 C

El Dx C/

E4 E4 D>>

EO ES D

E2 El C

D>>

E3 E4 C

D D/

D/

Pin Enrichment' Zoning Shin hssy 0

Fuel

Loathing, IE-T re Shim@

Pins 0 Pins V/0 0 Pins M/0 8-10/ in hvgo No. of

hssy, Assv.

Enzkehmene EO 0

236 184 4.03 52 3.90 El 16 220 168 4.03 S2 3.90

.024 E2 16 220 16&

3.90 52 3.30

.024 E3 16 220 16&

3.90 52 3.60

.026 E4 16 220 168 3.90 52 3.60

.016 ES 4

232 180 4.03 52 3.90

.012 24 20 L2 L2 32 8

108 4.001 3.999 3.829 3.829 3.829 4.001

II

(

Figure 3-4 PALO VERDE U1C3 30,173 27,144 11,068 14,047 10 18 24,530 19 12 20 13 23,818 21 14 22 0

15 16,134 23 16 24 17 26,697 25 30,116 27,946 11,428 16,630 16,447 26 27 28 29 30 31 32 33 34 0

23,817 35 36 27,945 37 26,183 38 39 0

13,258 40, 41 16,.032 42 27,140 43 44 0

0 45'1,398

.46 47 0

16,369 48 49 16,392 50 51 11,068 52 53 0

16,131 54 16,637 55 13,180 56 57 0

24,506 58 26,815 60 61 0

14,030 62 63 16,444 0

16,376 66 67 11,955 68 16,063 69 0

26,704 16,334 0

26,350 16,270 19,925 Assembly average burnup at BOC3 (MWD/T) 3-6

,t I

8 LEAD REGULATING BANK 4 'SECOND REGULATING BANK

.3 THIRD REGULATING BANK 2

FOURTH REGULATING BANK 0, - LAST REGULATING BANK 8 - SHUTDOWN BANK8 A

-SHUTDOWN BANKA P2 PLR GROUP 2 Pq -PLR GROUP l.

S SPACE CEA LOCATIONS 10 12 4

14 IS 1$

P2 5

2$

27 24 8

25 27 P2 41 29 8

21 4

47 SO 51 Pq 52 57 5$

Pg

$ 1 8

$7 4

~9 A

70 71 A

72 T4 '5 77 A

7$

$0 P2

$2

$2

$4

$7 90 91 92 P2 94 9$

9$

112 147 97 9$

114 115 121 22 14$

149 P2 99 11$

100 10'I 117 t1$

119 124 125129 1$ 1 152

'IS2 102 3

127 154 104 105 10$

121 122

'g 12$

129 140 15$

1S4

'!57 107 124 141 10$

109 110 125

~

12$

'127 142 142 8

159 1$ 0 141 Pp 11 1 112 125129

'I45 144 1$ 2 1$2 1$ 5 1$ $

1$ 7 1$$

1$ 9 170 171 172 172 A

74 175 4

17$

177 17$

8 179 150 1$ 1 1$2 1$4 1$5 I $$

P) 1$7 1$ $

1$9 190 191 192 Pq 1$ 2 194 1$ 5 199197 19$

8 199 200 8

201 8

204 8

205 20T 209209 8

210 211 4

2'12 212 214 215 P2 21 ~

217 21 ~

21$

Pp 220 22'1 224 225 225227 229 220 221 22 A

225 22$

227 22$

3 229 240 ARIZONA o Verde G5neretiny Statian CEA BANK IDENTIFICATION Figure 3%A

h II

P R

RlZONA lo Verde ar Genaratiny Station CSA PWTTaRN Figure 3kB

f I

0

47 10 BOX INSTR.

9 10 so r'I 11 3

24 6

53 14 12 12 13 14 4

ld 41 15 15 17 16 31 9

61 17 81 20 18 21 100 10'1 69 70 71 103 104

'105 72 73 74 91 92 24 106109 75 19 110 78 25 111 112 113

'114 29 115 11 117 11$

119

'l20 121 30 31 124 125 12d

. 32 128 129 34 131 147 148 164 165 1$6 42 1

134

~

135 36 150 151 152 39 157 168 153 170 137

.138 154 155 171 172 44 15$

173 174 141 l42 159 160 175 17$

144 1$ 1 41 145 14$

182 163 179 46 181 47 196 197 1S3 184 185 198199 200 51 186 48 187 188 1SQ

.49 191 50 192 193 207 52 53 210 211 212 213 214 215 55 238 216 217 218 56 219 231 59 241 60 RlZDNA lo Verde er Generating Station INSTRUMENT LOCATIONS Figure 34

4.0 FUEL SYSTEM DESIGN MECHANICAL DESIGN The mechanical design of the Batch E reload fuel assemblies is identical to the design of the Reference Cycle Batch D reload fuel a semblies except for a modification to the poison rod assembly design.

No changes in mechanical design bases have occurred since the original fuel design.

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 length 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 lif'e internal pressure.

In addition, this change makes the fuel and poison rods equal'in length.

4.2 GUIDE TUBE WEAR Twenty of the fuel assemblies 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 the required licensing procedures required by the NRC for all plants after the.first cycle of operation (References 4-1, 4-7, and 4-8).

A similar program was also performed on Unit 2 during the first refueling outage (Reference 4-2).

The number of assemblies inspected for guide tube wear was determined based on the results of the Unit 1 inspection.

The inspections revealed that guide tube wear was minor and will not adversely affect the fuel assembly performance throughout its expected life in the core.

Thus no guide tube wear inspections are necessary for EOC2.

II

4 3 THERMAL DESIGN The thermal performance of composite fuel pins that envelope the pins of fuel batches 8,

C, D and E present in Cycle 3 have been evaluated using the FATES3A version of the C-E fuel evaluation model (References 4-3 and 4-4) as approved by the NRC (Reference 4-5).

The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak pin at each burnup interval, from beginning of cycle to end of cycle burnups.

The burnup range analyzed is in excess of that expected at the end of Cycle 3.

4.4 CHEMICAL DESIGN

. The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch E fuel are identical to those of the fuel batches included in Cycle 2.

Thus, the chemical or metallurgical performance of the Batch E fuel will remain.

unchanged from th'e performance oi'he Cycle 2 'fuel.

4.5 SHOULDER GAP ADE UACY Measured shoulder gap data (references 4-1 and 4-2) acquired from post Cycle 1 inspection of fuel assemblies at PVNGS Units 1 and 2

indicated that the fuel had adequate shoulder gap for Cycle 2

operation.

Although the present shoulder gap is projected to be adequate for Cycle 3 operation, additional shoulder gap inspections will be performed at EOC2.

4-2

l

5.0 e

NUCLEAR DESIGN PHYSICS CHARACTERISTICS 5.1.1 Fuel Management 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 330 EFPD, the Cycle 3

reactivity lifetime for full power operation is expected to be 475 EFPD.

Explicit evaluations have been performed to assure applicability of all analyses to a Cycle 2 termination burnup of between 298 and 356 EFPD and for a Cycle 3 length up to 500 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 intended to represent nominal core parameters.

Those values used in the safety analysis (see Sections 7 and 8) contain appropriate 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.

I i

The CEA core locations and group identifications remain the same as in the Reference Cycle.

The power dependent insertion limit (PDIL) for regulating groups and part length CEA groups is shown in Figures

'5-1 and 5-2 respectively.

Table 5-3 shows the reactivity worths of various CEA groups calculated at full power conditions for Cycle 3

and the Reference Cycle.

Power Distribution Figures 5-3 through 5-5 illustrate the calculated All Rods Out (ARO) relative 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 points at the beginning, middle, and end of cycle were chosen to display the variation in assembly and maximum planar radial peaking as a function of burnup.

Relative assembly power densities for rodded configurations are given for BOC and EOC in Figures 5-6 through 5-11.

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

part length CEAs (PLCEAs),

Bank 5, and Hank 5 plus the PLCEAs.

The radial power distributions 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 1.23 at BOC3 to 1.12 at EOC3.

)

t

5.2 PHYS CS ANALYSIS METHODS 5.2.1

~

~

Anal tica1 In ut to In-Core Measurements In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in accordance with Reference 5-1.

ROCS-DIT with the MC module will be used.

ROCS-DIT and the MC 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 5.31.,

based on Reference 5-1, will be applied to the Cycle 3 COLSS and CPC on-line calculations which use planar 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 MC computer codes employing DIT calculated cross sections.

ROCS, MC, and DIT were described in Reference 5-2.

5-3

II

TABLE 5-1 PVNGS-I CYCLE 3 NOMINAL PHYSICS CHARACTERISTICS Dissolved Boron Units Refer ence

~Ccl e

~Cele 3

, Dissolved Boron Concentration for Criticality, CEAs Withdrawn, Hot Full Power Equilibrium 'Xenon, BOC PPM 1088 1223 Boron Worth Hot Full Power, BOC Hot Full Power, EOC PPM//8p PPM/Xhp 112 90 127 98 Moderator Tem erature Coefficients Hot Full Power,'qui'librium Xenon Beginning of Cycle 10-4hp/

F End of Cycle 10-4hp/

F

-.5

-2.4

.6 3 ~ 3 Hot Zero Power, Beginning of Cycle 10-4hp/'F Do ler Coefficient

+0.3

+0.3 Hot Zero Power, BOC Hot Full Power, BOC Hot Full Power, EOC 10-Mp/

F 10-5hp/

F 10-5hp/

F

-1.8

-1.5

-1.7

-2.1

-1.7

-1.9 Total Dela ed Neutron Fraction eff BOC EOC

.0060

.0051

.0069

.0046 Prom t Neutron Generation Time l

BOC EOC 10-6 sec 10-6 sec 5-4 24.0 29.9 20.7 27.3

I 1

TABLE 5-2 PVNGS-1 CYCLE 3 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT FULL POWER STEAN LINE BREAK; Qp, END-OF-CYCLE (EOC)

Refer ence

~Cel e

~Cele 3

Worth of all CEAs Inserted

-16.0

-18.0 Stuck CEA Allowance

+4.5

+5.5 Morth of all CEAs Less Highest Worth CEA Stuck Out

-11.5

-12.5 Full Power Dependent'nsertion Limit CEA Bite

+.2

+0.2 Calculated Scram Worth

-11.3

-12.3 Physics Uncertainty

+1.2

+1.2 Other Allowances (losses due to voiding)

+.1

+O.l Net Available Scram Morth

-10.0

-11.0 Scram Morth Used in Safety Analysis

-10.0

-10.2

0 TABLE 5-3 PVNGS-1 CYCLE 3 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER,

%hp Be innin of C cle End of C cle Regulating CEAs Reference

~Cc1 e

~Cele 3

Reference

~Cc1 e

~Cele 3

Group 5

-.24

-.31

-e27

~ 33 Group 4

~ 37

~ 37

-.42

-.39 Group 3

-.90

-.91

-.9.2'2 Note:

Values shown assume sequential group insertion.

l

1.0 CER INSERTION LIMITS VS.

THERMRL POWER 0.9 0.8 0.7 0.6 0.5 0.4 GN LI 0.3 0.2 O.t 0.0 BRNK 6 BANK 3 160 120 90 BO 30 0

160 120 90 BO I

I I-l-~~

I I

BRNK 4 160 120 90 BO 30 0

160 l~

BRNK 1

30 0

1 60 120 90 OO I

J I

I BRNK 2 120 90 BO 30 0

L 30 0

I CER WITHDRRWRL ( INCHES3

n Ql

~ ~

Xl O

tal

-I

~l lat 0.00 o.ao tL 0.70 cf uxl 0 GQ ul Q.60 OAQ 2.'.30 tg 0.20 0.10 n vl tA

'M O

UHACCEPTABLK OPERATtOH

~

~

RESTRICTEO OPERATNN M

M fa CI 4tfl Cl n

taltil CI tai IG0 140 l30 I20 II0 IOQ 00 SQ 70 60 60 40 30 20 l0 0

I I ni>r I.avonl ca@ I osn<no. wciics wnl<nnneat I

I tt Ij t

1

Figure 5-3 PALO VERDE U1C3 Fxy - 1.540 Box 38 0.293 0.642 3

0.957 0.996 0.290 0.736 0.893 1.111 10 1.157 1.232 12 13 14 16 17 0.385 0.985 0.759 1.155 1.075 1.260 0.884 18 19 20 21 22 23 24 25 0.291 0.986 1.226 0.780 1.185 1.119 1.086 1.257 0

26 0.736 35 27 0.759 36 28 29 37 38 0.780 0,808 30 1 ~ 274 39 31 1.278 40 32 1.298 41 33 1.109 42 0.293 0.893 1.155 1.186 1.275

. 1.091 1.275 1.081 1.219 43'4 45 46 47'8 49 50 51 0.642 1.111 1.074 1.118 1.279 1.276 0.933 1.256 0.879 52 54 56 57 58 59'0 0.957 1.157 1.258 1.083 1.295 1.079 1.252 1.196 1.037 61 62

. 63 64 65 66 67 68.

69 0.996 1.232 0,884 1.257 1.109 1.219 0.879 1.028 0.872 ARO Assembly Relative Power Densities at HFP BOC3 with Eq.

Xe

f, Il il I

Figure 5-4 PALO VERDE'U1C3 Fxy - 1.512 Box 25 0.338 0.659 0.880

0. 922 10 0.280 0.733 1.017 1.212 1.110 1.265 12 13 14 15 16 17 18 0.280 19 20 0.833 1.029 0.340 0.832 0.721 21

0. 721 1.221 1.056 1.325 22 23 24 1.090 1.036 1.067 0.919 25 1.355 26 27 28 29 30 3.1 32 33 34 0.338 0.733 1.018 0.721 0.721 36 1.221 i.091 0:781 38 1.293 1.292 1.208 1.376 39 40 41 1.089 1.377 1.112 1.132 42 1.343 43 44 45 46 47 48 49 50 0.659 1.213 1.057 1.036 1.210 1.378 0.988 1.365 0.913 52 54 55 56 57 58 59 60 0.880 1.110 1.325 1.066 1.374 1.112 1.363 1.138 0.976 61 62 63

~

64 66 67 68 69 0.922 1.265 0.919 1.355 1.132 1.343 0.913 0.969 0.812 ARO Assembly Relative Power Densities at HFP MOC3 with Eq, Xe 5-10

,f I

Figure 5-5 PALO VERDE U1C3

.Pxy 1.468 Box 25 0.414 0.. 713 0.876 0.917 10 0.350 0.837 1,151 1.255 1.086 1.288 12 13 14 15 16 17 0.412 0.924 0.802 1.269 1.040 1.335 0.943 18 20 21 22 23 24 25 0.350 0.925 1.094 0.782 1.068 0.982 1.024 1.341 26 27 28 29 30 31

'2 33 43 44 0.713 1.255.

0.837 35

'.414 1.151 0.802 36 1.269 45 1.040 0.782 0.830 1.296 37 38 39

1. 069

1. 29'6

1. 0'42 46 47 48 0.983 1.117 1.304 1.116 40

'.303 49 0.924 1'.028 1.287 50 51 1.222 0.852 1.316 1.066 41 42 52 53

'5 57 58 60

0. 876

1. 087 1.335 1.025 1 ~ 317 1.029 1.222 0.998 0.869 61 62 63 65 66 67 68 69 0.917 1.288 0.943 1.341 1.066.

1.287 0.852 0.865 0.747 ARO Assembly Relative Power Densities at HFP EOC3 with Eq.

Xe 5-11

al il 4

I l

gt

Figure 5-6 PALO VERDE U1C3 Fxy - 1.556 Box 38 0.290 0.729 0.288 0.872 0.634 0.954 0.997 10 1.076.

1.143 1.229 13 14 15 16 17 0.386 0.984 0.751 1.119 0 '63 1.236 0.883 18 19 20 21 22 23 24 25 0.291 0 '85 1.217 0.764 1.166 1.103 1.091 1.275 26 27 28 29 30 31 32 33 34 0.288 0.730

'5 0.872 0.751 0.765 0.736 36 37 38 1.119 1.167 1.270

'1.270

'9 1.112 1.300-1.336 1.146 40 41 42 1.320 1.128 1.276 43 44 45 46 47 48 49 50 0.634 1 ~ 076 0.962 1.102 1.301 1.320 0.977 1.320 0.924 52 53 56 57 58 59 60 0.954 1.143 1.234 1.089 1.331 1.126 1.316 1.252 1.074 61 62 63 64 65 66 67 68 69 0.997 1.229 0.883 1.275 1.146 1.276 0.924 1.065 0.832 Assembly Relative Power Densities at HFP BOC3 w/ PLCEAs inserted w/ ARO Eq.

Xe 5-12

i1 t

Figure 5-7 PALO VERDE U1C3 Fxy 1.578 Box 28 0.320 0.698 1.036 1.078 0.325 0.814 0.973 1.193 1.230 10 1.308 12 13 14 15 16 17 0.433 1.101 0.833 1.238 1.122 1.277 0.883 18 19 20 21 22 23 24 25 0.326

. 1.102 1.359 0.844 1.237 1.117 1.006 1.085 26 27 28 29 30 31 32 0.816 0.835

-0.844 0.849 3.;293 1.230 1.102 0.'616 34 35 36 37 38 39 40 41 42 0.321 0.975 1.240 1.240 1.294 1.077 1.208 0.952 0.999 43 44 45 46 47 49 50 51 0.700 1.195 1.124 1.119 1.233 1.210 0.875 1.156 0.798 52 53 54 55 56 57 58 59 1.039 1.234 1.280 1.007 1.103 0.953 1.153 1.118 0.971 61 62 63 64 65 66 67 68 69 1.078 1.308 0.883 1.085 0.616 0.999 0.798 0.963 0.821 Assembly Relative Power Densities at HFP BOC3 w/ Bk5 inserted w/ ARO Eq.

Xe 5-13

l

Figure 5-8 PALO VERDE UlC3 Fxy - 1.585 Box 20 0.317 2

0.693 1.038 1.084 10 0.327 0.812 0.955 1.160 1.220 1.309 12 13 14 15 16 17 0.437 1.108 0.830 1.204 1.006 1.252 0.882 18 19 20 21 22 23 24 25 0.328 1.109 1.358 0.831 1.220 1 ~ 099 1.006 1.095 26 27 28 29 30 31 32 33 35 36

'&,8?4 0.831

0. 831 37 0.776 1.287 39

l. 249 40 1.130 41 0.635 42 0.318 0.957 1.206 1.223 1.289 1.096 1.250 0.993 1.046 43 46 47 48 49 50 51 0.695 1.163 1.008 1.101 1.252 1.252 0.917 1.217 0.841 52 53 55 56 57 58 59 60 1.041 1.223 1.255 1.008 1.131 0.994 1.214 1.173 1.009 61 62 63 64 65 66 67 68 69 1.084 1.309 0.882 1.095 0.635 1,046 0.841 1.000 0.786 Assembly Relative Power Densities at HFP k

BOC3 w/,Bk 5 and PLCEAs inserted w/ ARO Eq.

Xe 5-14

Figure 5-9 PALO VERDE U1C3 Fxy - 1.496 Box 25 0.405 0.701 '.872 0.918 0.348 0.825 1.115 10 1.202 1.069 1.285 12 13 14 15 16 17 0.412 0.918 0.787 1.212 0.903 1.302 0.944 18 19 20 21 22 23 24 25 0.349 0.919 1.076 0.754 1.039 0.961 1.034 1.372 26 27 28 29 30 31 32 33 0.825 0.787 0.755 0.731 .283'.145'1.375.'.120 35 36 37 38 39 40 41 42 0.405 1.115 1.212 1.040 1.284 1.068 1.371 1.096 1.379 43 44 45 46 47 48 49 50 51 0.701 1.202 0.903 0.962 1 ~ 146 1.372 0.988 1.316 0.918 52 53 54 55 56 57 58 59 60 0.872 1.069 1

~ 302 1.035 1.376 1.097 1.316 1.068 0.917 61 62 63 64 65 66 67 68 69 0.918 1.285

.0.944 1.372 1.120 1.379 0.918 0.913 0.708 Assembly Relative Power Densities at HFP EOC3 w/ PLCEAs inserted w/ ARO Eq. Xe

1

Figure 5-10 PALO VERDE U1C3 Fxy - 1.515 Box 54 0.459 0.784 0.955 0.996 0.395 0.937 1.270 1.363 1.161 10 1.368 12 13 14 15 16 17 0.468 1.042 0.889 1.376 1.095 1.356 0.941 18 19 20 21 22 23 0.396 1.042 1.221 0.850 1.122 0.981 0.941 1.146 34 26 27 35 36 0.937 0.890 28 0.851 37 29

.0.873 38 30 31 32 39 40 1.311 1.061 1.991 33 0.572 0.459 1,270 1.377 1.123 1.312 1.019 1.216 0.884 1.024 43 44 46 47 48 49 50 51 0.784 1.364 1.096 0.982 1.063 1.217 0.849 1.096 0.752 52 53 54 55 56 57 58 0.956 1.162 1.358 0.942 1.093 0.8&5 1.096 0.907 0.791 61 62 63 64 65 66 67

, 68 69

'.996 1.368 0.941 1.146 0,572 1.024 0.752 0.788 0.684 Assemb'ly Relative Power Densities at HFP EOC3 w/ Bk 5 inserted w/ ARO Eq.

Xe 5-16

Figure 5-11 PALO VERDE U1C3 Fxy - 1.481 Box 54 0.454 0.778 0.958

'l. 005 0.399 0.933 1.242 1.315.

l.'148 10 1.371 12 13 14 16 17 0.474 1.049 0.882 1.326 0.954 1.322

0. 941.

18 19 20 21 22 23 25 0.399 1.049 1.214 0,827 1.095 0.957 0.943 1.164 26 27 28 29 30 31 32 33 0.934 0.8'83 34 35 36 0.454 1.243 1.326 43 44 45 0 '78 1.316 0,955 53 54 0.959 1.149 1 ~ 323.

0.828 37 1.097 46 0.958 55 0.945 38 39 40 1.298 1.042 1.277 47 1.087 48 49 1.278 0.908 56 57 58 1.136 0.944 1.183 0.771

-1:.297

. 1.085 1.134 41 0.942 50 1.183 59

\\

0.974 0.598 42 1.098 51 0.812 60 0.837 61 62 63 64 66 67 68 69 1.005 1.371 0.941 1.164 0.598 1.098 0.812 0.833 0.651 Assembly Relative Power Densities at HFP EOC3 w/ Bk 5 and PLCEAs inserted w/ ARO 'Eq.

Xe

jrl

6.0 THERMAL"HYDRAULICDESIGN

'NBR ANALYSIS Steady state DNBR analyses of Cycle 3 at the rated power level of 3800 MHT have been performed u ing the TORC computer code described in Reference 6-1, the CE-1 critical heat flux correlation described in References 6-2 and 6-8, and the CETOP code described in Reference 6-3.

Table 6-1 contains a list of pertinent thermal-hydraulic design parameters.

The Modified Statistical Combination of Uncertainties (MSCU) methodology presented in Reference 6-4 was applied with Palo Verde-1 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 by the Core Protection Calculators (CPC) and Core Operating Limit Supervisory System (COLSS) which, when used with the Cycle' DNBR limit of 1.24, provide assurance at the 95'795 confidence/probability level that the hot rod will not experience'NB.

The 1.2$

DNBR limit was calculated using the methodology of Reference'6-5 as was done for the Reference Cycle.

This Cycle 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.

e Other penalties imposed by NRC in the course of their review of the Cycle 1 Statistical Combination of Uncertainties (SCU) analysis discussed in Reference 6-5 (i.e.,

TORC code uncertainty and CE-1 CHF correlation cross validation uncertainty, as discussed in Reference 6-6) are included in the overall uncertainty penalty factors derived in the Cycle 3

MSCU analysis.

6-1

6.2 EFFECTS OF FUEL ROD BOWING ON DNBR MARGIN 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%

NDNBR, is valid for bundle burnups up to 30,000 NWD/MTU.

This penalty is included in the 1.24 DNBR limit.

For assemblies with burnup greater than 30 GWD/T sufficient available ma~gin 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 all assembly burnups expected for Cycle 3.

TABLE 6-1 PVNGS-1 Cycle 3 Thermal H draulic Parameters at Full'ower General Characteristics Total Heat Output (Core only)

Fraction of Heat Generated in Fuel Rod Primary System Pressure Nominal Inlet Temperature (Nominal)

Total Reactor Coolant Flow (Minimum Steady State)

Coolant Flow Through Core (Minimum)

Hydraulic Diameter (Nominal Channel)

Average Mass Velocity Pressure Drop Across Core (Minimum steady state flow irreversible hP over entire fuel assembly)

Total Pressure Drop Across Vessel (Based on nominal dimensions and minimum steady state flow)

Core Average Heat Flux (Accounts for fraction of heat generated in fuel rod and axial densifica-tion factor)

Total Heat Transfer Area (Accounts for axial densification factor)

Units MW$

10 Btu/hr psia 0F

.Qpg 10 lb/hr 10 lb/hr ft 10.lb/hr-ff psi psi BTU/hr-ft Reference

~Cc1 e 3800 12,970 0.975 2250 565.0 423,300 155.8 151.1 0.039 2.49 14.5 51.3 185,700*

68,100*

~Cele 3

3800 12,970 0.975 2250 565.0 423,300 155.8 151.1 0.039 2.49 14.5 51.3 184,200**

68,600**

Film Coefficient at Average Conditions BTU/hr-ft F

6100 6100 Average Film Temperature Difference F

30 30 Average Linear Heat Rate of Unden-sified Fuel Rod (Accounts for fraction of heat generated in fuel rod) kw/ft 5.4 5.4

jll

't

General Characteristics TABLE 6-1 (continued)

Units Reference

~Cc1 e

~Cele 3

Average Core Enthalpy Rise Maximum Clad Surface Temperature Engineering Heat Flux Factor Engineering Factor on Hot Channel Heat Input Rod Pitch, Bowing and Clad Diameter Factor BTU/lb OF 85.9 656 1.03+

1.03+

1.05+

85.9 656 1.03+

1.03+

1.05+

Fuel Densification Factor (Axial) 1.002 1.002 NOTES:

Based on 2288 poison rods.

~

~

~

e

~

~

Based

.on 1872 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 and CPC

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

Tech.

Spec.

minimum flow rate.

NON-LOCA SAFETY ANALYSIS Introduction This section presents the results of the Palo Verde Nuclear Generating Station Unit 1 (PVNGS-1), Cycle 3 Non-LOCA safety analyses at 3800 NWt.

The Design Basis Events (DBEs) considered in the safety analyses are listed in Table 7.0-1.

These events are categorized into three groups:

Moderate Frequency, Infrequent, and Limiting Fault events.

For the purpose of this report, the Moderate 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 Melt SAFDLs'),

and Loss of Shutdown Margin.

Tables 7.0-2 thr'ough /.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

'hosen for analysis for each criterion are the limiting events with respect to that criterion.

7.0.2 Methods of Anal sis The analytical methodology used for PVNGS-1 Cycle 3 is the same as 0

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),

o'r on other dockets is used.

7-1

I l

I

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 fluid conditions within 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 p'redicted to experience.clad failure i.s taken as the number of pins which have a CR-1 DNBR value below 1.24.

The only exceptions are the Shaft Seizure and Sheared Shaft events

.for which the statistical'onvolution

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

7.0.4 In ut Parameters and Anal sis Assum tions

~

~

Table 7.0-6 summarizes the core parameters assumed in the Cycle 3 transient analysis and compares them to the values used in the Reference Cycle.

Specific initial conditions for each event are tabulated in the section of the report summarizing that event.

Tech Spec changes are described in Section 10.. The effects of these changes were considered for each DBE and were, included as appropriate.

For some of the DBEs presented, certain initial core parameters were assumed to be more limiting than the actual calculated Cycle 3 values.

Such assumptions resulted in more adverse consequences.

Events which have credited CPC trip protection have assumed instrument channel response times which are conservative relative to the Cycle 3 Technical Specifications.

Conclusion All'BEs have been evaluated for PVNGS-1, Cycle 3 to determine whether their results are bounded by the Reference Cycle.

7"3

Table 7.0-1 PVNGS Unit 1 Desi n Basis Events Considered in the C cle 3 Safet Anal sis 7.1 Increase in Heat Removal by the Secondary System 7.2 7.1.1 7.1.2 7.1.3 7.1.4 7.1.5~

7.1.6 Decrease Decrease in Feedwater Temperature Increase in Feedwater Flow Increased Main Steam Flow Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve Steam System Piping Failures Excess Load With a Single Failure 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 7.2.5 Loss of Normal Feedwater 7.2.6*

Feedwater

'System Pipe Breags I

Decrease in Reactor Coolant Flowrate 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 7.4.2 7.4.3 7.4.4 7.4.5 7.4.6~

Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition Uncontrolled CEA Withdrawal at Power CEA Misoperation Events CVCS Malfunction

( Inadvertent Boron Dilution)

Startup of an Inactive Reactor Coolant System Pump Control Element Assembly Ejection 7.5 Increase in Reactor Coolant System Inventory 7.5.1 7.5.2 CVCS Malfunction Inadvertent Operation of the ECCS During Power Operation

• Categorized as Limiting Fault Events 7-4

Ol

Table 7.0-1 (continued) 7.6 Decrease in Reactor Coolant System 1nventory 7.6.1 7.6.2" 7.6.3~

Pressurizer Pressure Decrease Events Small Primary Line Break Outside Containment Steam Generator Tube Rupture 7.7 Miscellaneous 7.7.1 Asymmetric Steam Generator Events

• Categorized as Limiting Fault Events 7-5

Table 7.0-2 DBEs Evaluated with Res ect to Offsite Dose Criterion Section 7.1.6 7.2.4 Event A)

Anticipated Operational Occurrences 1)

Excess Load Mith a Single Failure 2)

Loss of Normal AC Power B)

Limiting Fault Events Results Presented Bounded by Reference Cycle 7.1.5a 7.1.5b 7.2.6

~

~

7.3.2')

Steam System Piping Failures:

a)

Pre-Trip Power Excursions b)

Post-Trip Return-to-Power 2)

Feedwater System Pipe Breaks 3)

Single Reactor Coolant Pump.

%haft Seizure/Sheared Shaft Bounded by Reference Cycle Bounded by Reference Cy'cle Bounded by Reference Cycle 7.4.6 4)

Control Element Assembly Ejection Bounded by Reference Cycle 7.6.2 5)

Small Primary Line Break Outside

'Containment Bounded by Reference Cycle

7. 6.3 6)

Steam Generator Tube Rupture

'I Bounded by Reference Cycle 7"6

l 1

Table 7.0-3 DBEs Evaluated with Res ect to RCS Pressure Criterion Section Event A)

Anticipated. Operational Occurrences Results 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

~

~

7.4.l 7.4.2 5)'oss of Normal Feedwater 6)

Uncontrolled CEA'Mithdrawal from Subcritical or Low Power Condition 7)

Uncontrolled CEA Withdrawal at Power Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle 7.5.1 7.5.2 8)

CVCS Malfunction 9)

Inadvertent Operation of the ECCS During Power Operation 8)

Limiting Faul t Events Bounded by Reference Cycle Bounded by Reference Cycle 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

l l

Table 7.0-4 DBEs Evaluated with Res ect to Fuel Performance Section 7.1.1 7.1.2 Event A)

Anticipated Operational Occurrences 1)

Decrease in Feedwater Temperature 2)

Increase in Feedwater flow Resul ts Bounded by Reference Cycle Bounded by Reference Cycle 7.1.3 3)

Increased Main Steam Flow Bounded by Reference Cycle 7.1.4 4)

Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve Bounded by Reference Cycle 7.1.6 7.4.1 5)

Excess Load With a Single Failure 6)

Total.L'oss of Forced Reactor Coolant Flow 7)

Uncorftrolled CEA Withdrawal from a Subcritical or Low Power Condition Presented Bounded by Reference Cycle Bounded by Reference Cycle 7.4.2 7.4.3 7.6.1 8)

Uncontrolled CEA Withdrawal at Power 9)

CEA Misoperation Events 10)

Pressurizer Pressure Decrease Events Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle 7.7.1 11)

Asymmetric Steam Generator Events Bounded by Reference Cycle B)

Limiting Fault Events 7.1.5a 7.1.5b

~

~

1)

Steam System Piping Failures:

a)

Pre-Trip Power Excursions b)

Post-Trip Return to Power Bounded by Reference Cycle 7-8

l

,l, t

Section Event Table 7.0-4 (continued)

Results 7.3.2 2)

Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft Bounded by Reference Cycle 7.4.6 3)

Control Element Assembly Ejection Bounded by Reference Cycle 7"9

Table 7.0-5 DBEs Evaluated with Respect to Shutdown Mar in Criterion Section Event

'esults A)

Anticipated Operational Occurrences 7.1.4 1)

Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve Bounded by Reference Cycle 7.4.4 2)

CVCS Malfunction (Inadvertent Boron Dilution)

Bounded by Reference Cycle 7.4.5 3)

Startup of an Inactive Reactor Coolant System Pump Bounded by Reference Cycle

~

~

7.1.5b

.')

Limiting Fault Events 1)

Steam System Piping Failures:

II a)

Post-Trip-Return-to-Power Bounded by Reference Cycle 7"10

Table 7.0-6 PVNGS Unit 1.

C cle 3 Core Parameters In ut to Safet Anal ses Safet Parameters Total RCS Power (Core Thermal Power

+ Pump Heat)

Units MMt Reference Cycle Value 3898 C cle 3 Value 3898 Core Inlet Steady State F

Temperature 560 to 570 (90% power and

,above) 550 to 572 (below 90% power) 560 to 570 (90% power and above) 550 to 572 (below 90% power)

Steady State RCS Pressure Minimum Guaranteed Delivered Volumetric Flow Rate Axial Shape Index L'CO Band Assumed for All Powers psia gpm ASI Units 2000 " 2325 423,320

-0.3 to +0:3 2000 - 2325 423,320

~ -0.3 to +0.3

~

'aximum CEA Insertion at Full Power

% Insertion 28 of Lead Bank

% Insertion 25 of Part-Length 28 25 Maximum Initial Linear Heat, Rate Steady State Linear Heat Rate for Fuel Center Line Melt CEA Drop Time from Removal of Power to Holding Coils to 90%

Insertion Minimum DNBR CE-1 (SAFDL)

Macbeth (Fuel failure limit for post-trip SLB with LOAC-References 7-5 and 7-6)

I KM/ft KM/ft sec 13.5 21.0 4.0 l.24 1.30 13.5 21.0 4.0 1.24 1.30 7"11

Table 7.0-6 (continued)

Safet Parameters Moderator Temperature Coefficient Units 10 hp/

F Reference Cycle

'alue Figure 7.0-1 C cle 3 Value Figure 7.0-1 Shutdown Margin (Value Assumed in Limiting Hot Zero Power SLB)

-6.5 "6.5 7-12

INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Decrease in Feedwater Tem erature The results are bounded by the Reference Cycle.

0 Analysis of the Decrease in Feedwater Temperature with a Single Failure is discussed in Section 7.1.6.

Increase in Feedwater Flow The results are bounded by the Reference Cycle.

Analysis of the Increase in Feedwater Flow with a Single Failure is discussed in Section 7.1.6.

Increased Main Steam Flow The results are bounded by the Reference Cycle.

Analysis of the Increased Main Steam Flow with a Single Failure is discussed in Section 7.1.6.

Inadvertent 0 enin of a Steam Generator Safet Valve or Atmos heric The results are bounded by the Reference Cycle.

Analysis of this event with a Single Failure is discussed in Section 7.1.6.

7"13

7.1.5

~

~

Steam S stem Pioin Failures 7.1.5a Steam S~ stem Pioin Failures:

Inside and Outside Containment Pre-Tri Power Excur sions The results are bounded by the Reference Cycle.

7.1.5b Steam S stem Pioin Failures:

Post-Tri Return to Power

~ t The results are bounded by the Reference Cycle 7.1.6 Excess Load With a Sin le Failure 7.1.6. 1 Introduction SRP 15.1 requires an evaluation of the consequences of an excess load event in conjunction with a single active component. failure.

Previously, CESSANT used the inadvertent opening of an atmospheric

~

dump valve event with a turbine/generator trip,followed by a loss of off-site power before a reactor trip as the limiting excess load with a single failure.

I The loss of off-site power was assumed to have caused a coastdown of all four reactor coolant pumps at the worst possible time during the event.

At this ti'me the initiating. event had caused a degradation of the initial thermal margin to the point that the plant was on the verge of a CPC trip.

Based on this assumed sequence of events, eight percent of the. fuel, was reported to experience DNB for a short time.

The calculation of the amount of fuel which experiences ONB during this event uses a pin census of the number of fuel pins in each'f several radial peaking intervals.

Because this census is strongly affected by the core loading pattern, it has been necessary to re-evaluate the CESSAR event for each reload of the three P>JNCS units.

The cycle-to-cycle variability of this pin census could make-7"14

it necessary to increase conservatisms in LCOs and LSSSs to. avoid exceeding the docketed amount of fuel experiencing DNB.

ANPP has reviewed the CESSAR presentation of the inadvertent opening of an atmospheric dump value with loss of AC event and has concluded that the assumed sequence of events is overly conservative so that the consequences of the event are overstated.

No single active component failure could lead to loss of off-site power before reactor trip during.an excess load event.

Note that the loss of off-site power before reactor trip is not included in the list of single failures considered for the CESSAR safety analyses.

The ANPP review concluded that the specific failures which would be necessary to generate the current CESSAR sequence of events include all of the following:

1.

A Failure of the Turbine Control S stem

\\

A failure of the turbine control system could cause.

a turbine trip; Although there are a number of specific component failures which could cause a turbine trip, these fai1ures are infrequent so that the turbine would not be expected to trip before reactor trip in response to one of the excess lead events discussed in CESSAR section 15.1.

2.

A Failure of the Reactor Power Cutback S stem The CESSAR sequence of events assumes that the Reactor Power Cutback System (RPCS) is inoperable.

The RPCS is normally in service and would reduce core power to below 70K in response to a turbine trip.

This response would reduce the consequences of a turbine trip.

3.

Loss of Off-site Power A failure. could cause off-site-power to be lost following the 7-15

turbine/generator trip.

Loss of off-site power following a turbine trip is an infrequent occurrence.

Because an excess load with loss of off-site power before a reactor trip cannot be caused by a single active component failure, this event need not be considered to satisfy SRP 15.1.

7..:2 ~1* 2 The. list of single failures given in CESSAR Table 15.0-6 was examined to determine the worst single failure which could occur during an excess load event.

The excess load events which were considered in this review included:

a ~

b.

c ~

d.

Decrease in feedwater temperature Increase in feedwater flow Increased main steam flow Inadvertent opening of a steam generator relief or safety.valve

. These excess load events are CPCS design basis events so that the CPCS will act to prevent violation of SAFDLs during these events.

Additional failures could result in more severe consequences so that a small amount of fuel would be predicted to experience ONB for a short time.

The single failures from CESSAR table 15.0-6 which

~ might result in a lower ONBR during the event are discussed below:

a.

Loss of Off-site Power After Turbine Tri Turbine trip occurs as a consequence of reactor trip.

The loss of off-site power could occur after turbine trip due to grid instability.

However, there will be at least a three second delay between the time of turbine trip and the time of loss of off-site power,'based on the discussion presented in Reference 7-12.

This delay would prevent a reactor coolant pump coastdown until after the time at which CEA insertion would terminate the approach to the DNB SAFDL.

For this 7-16

I

reason, the minimum DNBR during the event would be no lower than that for the case without loss of off-site power.

b.

Failure of One Breaker to achieve Fast Transfer to Backu Power

~Su i~1 Turbine trip occurs as a consequence of a reactor trip.

The reactor coolant pump electrical busses are fast transferred to off-site power after turbine trip.

The failure of one bus breaker to fast transfer would result in a coast-down of two reactor coolant pumps.

However, there will be a delay of a least three seconds between the time of the turbine trip and the time of the attempt to fast transfer.

This delay would prevent a reactor coolant pump coastdown unti 1 after the time at which CEA insertion would terminate the approach to the DNB SAFDL.

For this reason, the minimum DNBR during the event would be no lower than the excess load case without a,failure

.'o'ast transfer.

c.

Failure of S ra Control l/gives to Close The pressurizer spray control valves might open in response to a reactor trip. If the spray valves failed to close, the rate of pressure decrease at the time of trip would be faster than if no failure occurred.

The maximum rate of pressure decrease would be less than 3.5 psia/sec.

The additional pressure drop

'uring the event could result in a slightly lower minimum DNBR.

d.

Failure of Backu Heaters to Turn On Pressuriz'er pressure and RCS pressure decrease during an excess load event.

Failure of the backup heaters to turn on could result in a faster pressure decrease at the time of trip than if no failure occurred.

The effect of a failure of the backup heaters to turn on would be less severe than the effect of failure of the spray control valves to close.

7-17

7.1.6.3 Resul ts

~

~

~

Analysis of the excess load with a single failure has shown that less than 0.7 percent of the fuel will experience DNB.

Additionally, it was found that none of the single failures affecting the ESF system would result in a return to power.

Therefore, for the single failure cases affecting the ESF system, no fuel would experience DNB after a reactor trip.

7.1.6.4 Conclusion In the event of an excess load in conjunction with a single failure, only a small amount of fuel would experience DNB.

10 CFR 100 dose limits are not exceeded.

7.2 DECREASE N HEA'T REMOVAL BY THE SECONDARY SYSTEM Loss of External Load The results are bounded by the Reference Cycle.

7.2.2 Turbine Tri The results are bounded by the Reference Cycle.

7.2.3 Loss 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"18

k l

7.2.5

~

~

Loss of Normal Feedwater The results are bounded by the Reference Cycle.

7.2.6 Feedwater S stem Pi e Breaks The results are bounded by the Reference Cycle.

7.3 DECREASE IN REACTOR COOLANT FLOWRATE 7.3.1 Loss of Forced Reactor Coolant The results are bounded by the Reference Cycle..

7.3.2 Sin le Reactor Coolant Pum Shaft Seizure/Sheared Shaft The results are bounded by the Reference Cycle.

'EACTIVITY.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.

7.4.2 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"19

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

7.4.6 Control Element Assembl E'ection The results 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 0 eration

'The results are bounded by the Reference Cycle.

DECREASE IN REACTOR COOLANT 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 Sreak 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.

7-20

4

MISCELLANEOUS 7.7.1 As etric Steam Generator Events The results are bounded by the Reference Cycle.

7=21

-1,0

-1,5 ALLONBLE

-2.0

-2.5

-3,0 0

20 lIO 60 80 100 CORE POHER LEVEL, X OF RATED THEfNAL POMER talo Verde Nuekat Generating Statloa AI.LOHMLKHTC HODES 1 AIID 2 f'Igure 7.0-l

I

8.0 ECCS ANALYSIS 8.1 8.1.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT Int'roduction And Summar An ECCS performance analysis of the limiting break size was.

performed for PVNGS-1 Cycle 3 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled reactors (Reference 8-1).

The analysis 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 Method Of Anal sis The ECCS performance analysis for PVNGS-1 Cycle 3 consisted of an evaluation of the differences between Cycle 3 and PVNGS-1 Cycle 2, hereafter referred to as the reference

cycle, Acceptable ECCS performance was demonstrated

.for the refer'ence 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 NRC approved C-E large break ECCS performance evaluation model which is described in Reference 8-4 including the use of a more conservative axial power shape.

The blowdown hydraulic calculations, refill/reflood hydraulics calcu) ations, and steam cooling heat transft,r coefficients of the reference cycle apply to PVNGS-1 Cycle 3 since there have been no significant changes to RCS hardware characteristics.

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 NRC approved STRIKIN-II (Reference 8-5) code was used for this purpose.

Burnup dependent calculations were performed. with STRIKIN-II to determine the limiting conditions for the'ECCS performance analysi's; The fuel performance data was generated with the FATES-3A fuel evaluation model (References 8-6 and 8-7) with the NRC grain size restriction (Reference 8-8). It was demonstrated that the burnup

"':.".".."with the highest initial fuel stored energy was limiting.

This occurred at a hot rod burnup of 969 MMD/NTU.

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.

Results Significant core and system parameters for the reference cycle and PVNGS-1 Cycle 3 are shown in Table 8.1-1.

Table 8.1-2 presents the analysis results for the 1.0 DEG/PD break which produces the highest peak clad temperature.

This l.imiting case results in a peak clad temperature of 1944 F> which is well below the acceptance limit of 2200'F.

The maximum local and core wide zirconium oxidation, as shown in Table 8.1-2, remain well below the acceptance limit values of 17% and 1%, respectively.

These results remain applicable for up to 400 tubes plugged per steam generator and a reduction in system flow rate to 155.8X10 ibm/hr and a reduction in core flow rate to 151. lx10 1 bm/hr.

6 The reduction in delivered low pressure safety injection flow (see Reference 8-11) does not impact the reflooding of the reactor vessel following a large break loss-of-coolant accident as long as there is sufficient flow from the safety injection pumps to maintain a full downcomer annulus following discharge of the safety injection tanks.

With the revised low pressure safety injection flow, there is sufficient flow to maintain a full downcomer.

III i

II l

Conclusion The'ECCS performance evaluation for PVNGS-1 Cycle 3 results in a

peak clad temperature of 1944'F, a peak local clad oxidation percentage of 5.4% and a peak core wide clad oxidation percentage of less than 0.80% compared to the acceptance criteria of 2200 F, 17%

and 1%, respectively.

Therefore, operation of PVNGS-1 Cycle 3 at a

core power level of 3876 MWt (102% of 3800 MWt) and a

PLHGR of 13.5 kw/ft is in conformance with 10CFR50.46.

8.2 SHALL 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-1 Cycle 1 bounds PVNGS-1 Cycle 3.

Therefore, acceptable small break LOCA ECCS performance iq demonstrated at a peak linear heat generation rate of 13.5 kw/ft and a reactor power level of 3876 MWT (102% of 3800 MWT)., This acceptable performance has been confirmed with up to 400 plugged tubes per steam generator.

The reduction in delivered low pressure safety injection flow (see Reference 8-11) 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

II I

TABLE 8.1-1 PVNGS-1 Cycle 3

Core and S stem Parameters Parameter Units Reactor Power 9 102% of Nominal(MWt)

Average Linear Heat Rate 9 102% of Nominal (kw/ft)

Peak Linear Heat Generation Rate (kw/ft)

Core Inlet Temperature

( F)

System Flow Rate (ibm/hr)

Core Flow Rate (ibm/hr)

Gap Conductance

) (Btu/hr ft F)

Fuel Center>ine Tetttperature -

( F).

Fuel Average Temperature

( F)

Hot Rod Gas Pressure (psia)

Hot Rod Burnup (mwd/mtu)

Number of Steam Generator Tubes Plugged per Steam Generator Augmentation Factor Minimum Initial Containment Pressure (psia)

Containment Free Volume (ft )

Axial Peaking Factor Low Pressure Safety Injection Runout Flow (gpm)

Reference

~Ccl e 3876 5.68 13.5 565.0 164xlO 159xlO 1652 3493.7 2124.0

• 1126.3 1000 400 1.0 13.4 3.0(10

)

1.52 4214

~Cele 3

3876 5.70 13.5 565.0 164x10 159xlO 1533 3316.7 2111.8 1149.8 1000 400 1.0 13.4 3.0(10

)

1.52 3744

(

) Initial values at the limiting hot rod burnup as calculated by STRIKIN-II at the peak linear heat generation rate.

TABLE 8. 1-2 PVNGS-1 Cycle 3

Limitin Break Size 1.0 DEG/PD Peak Linear Heat Generation Rate (kw/ft)

Peak Clad Temperature

(

F)

Time of Peak Clad Temperature (Seconds)

Time of Clad Rupture (Seconds)

Peak Local Clad Oxidation

(%)

Total Core-Mide Clad Oxidation

(%)

~Cele 2

13.5 1925 267.8 98.9 4.6

< 0.80

~Cele 3

13.5 1944 268.8 96.5 5.4

< 0.80

9.0 REACTOR PROTECTION AND MONITOR NG SYSTEM

'I

9. 1 INTRODUCTION The Core Protection Calculator System (CPCS) is designed to provide the low DNBR and high Local Power Density (LPD) trips to (I) ensure that the specified 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 design basis

events, provided that at the initiation of these 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 will be evaluated for applicability consistent with the cycle design, performance and safety analyses.

Any necessary change to the RDB constants will be installed in accordance with Reference 0-4.

9-1

l

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 annealing matrix, boundary point power correlation coefficient's, and adjustments for planar radial peaking factors),

(2) uncertainty factors to account for processing and measurement uncertainties in DNBR and 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 factors and other 'addressable constants will be determined for this cycle consistent with the software and methodology established in the CPC Improvement Program and the cycle

design, performance'nd safety analyses.

As for the Reference Cycle; uncertainty fhctors will be determined using the modified statistical combination of uncertainties method (Reference 9-5).

9.4 DIGITAL MONITORING SYST M

COLSS The 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 will be updated, as required, to reflect the reload core design.

9-2

10.0 TECHNICAL SPECIFICATIONS This section provides a summary of the proposed changes to the PVNGS Unit 1 Technical Specifications for Cycle 3.

A description of each change as well as a

summary of the justification for the change are provided.

The proposed changes are arranged in the order shown below.

Detailed change request. packages are presented in a separate submittal.

SECTION

~LC0 s SUBJECT 10.1 3.1.1.2 3.1.2.7 Shutdown Margin 10.2 3.1.3.6 Regulating CEA insertion Limits 10.3 3.2.3 Azimuthal Power Tilt 10A 3.2.4 DNBR Margin 10-1

t h

10.1 SECTION OF TECHNICAL SPECIFICATIONS AFFECTED Limiting Conditions for Operation 3.1.1.2 (Shutdown Margin) and 3.1.2.7 (Boron Dilution Alarms)

DESCRIPTION OF THE PROPOSED AMENDMENT This proposed change involves the revision of Figure 3.1-1A and Tables 3.1-2, 3.1-3 and 3.1-5.

Figure 3.1-1 A provides shutdown margin requirements versus RCS cold leg temperature for the case where any full-length CEA is withdrawn.

Tables 3.1-2, 3.1-3 and 3.1-5 provide required boron monitoring frequencies in the event that one or both startup channel high neutron flux alarms are inoperable.

The proposed revisions are. required to reflect cycle specific requirements.

The revisions result in more restrictive operating limits.

JUSTIFICATION FOR THE PROPOSED AMENDMENT The proposed changes to the shutdown margin and boron monit'oring frequency requirements are necessary to ensure that the Technical Specifications are consistent with the safety analyses performed for the Cycle 3

'ore design.

More restrictive operating limits are necessary to remain bounded by reference analysis results....

10-2

E TION OF TECHNICAL SPECIFICATIONS AFFECTED Limiting Condition for Operation 3.1.3.6 (Regulating CEA Insertion Limits)

DE CRIPTION OF THE PROP SED AMENDMENT This proposed amendment will revise Technical Specification Figures 3.1-3 and 3.1-4.

These figures provide regulating group Control Element Assembly (CEA) insertion limits.

Figure 3.1-3 provides CEA insertion limits when the Core Operating Limit Supervisory System (COLSS) is in service and Figure 3.1-4 provides the insertion limits when COLSS is out of service.

The following two changes are proposed:

i) The revised Figure 3.1-3 (COLSS in service) will not permit insertion of regulating group 3 CEAs above 20% of rated thermal power. This is more restrictive than the current specification which does allow for regulating group 3 insertion above 20% of rated thermal power.

ii) The revised Figure 3.1-4 (COLSS out of service) will permit slightly increased insertion of regulating group 3 CEAs between 15% and

~.20% of rated thermal power.

TIFICATION FOR THE PROPOSED AMENDMENT The proposed changes to the transient insertion limits are necessary to ensure that the Technical Specifications are consistent with the safety analyses performed for the Cycle 3 core design.

A more restrictive transient insertion limit line is required when COLSS is in service and a slightly less restrictive limit is allowed when COLSS is out of service.

10-3

l

10.3 E TION F TE HNICAL PE IFI ATIONS AFFECTED Limiting Condition for Operation 3.2.3 (Azimuthal Power Tilt)

DES RIPTI N OF THE PROPO ED AMENDMENT This proposed amendment will revise Technical Specification Figure 3.2-1,A.

This figure provides azimuthal power tilt limits versus core power for COLSS in service.

The azimuthal power tilt limit will be increased for operation below 40% of rated thermal power.

USTIFICATION FOR THE PROPO ED AMENDMENT Relaxation of the azimuthal power tilt limits with COLSS in service will allow the operators to better mitigate the consequences of xenon transients occurring below 40% of rated thermal power.

This change has been previously approved for PVNGS Unit 2 (reference Amendment 25 to Facility Operating License No. NPF-51).

In the previous request to modify this specification for Unit 2 (refer to letter 161-01059 dated May 27, 1988), ANPP stated that this change was unit and cycle specific and that supporting analysis would. need to be conducted for each unit.

'Additionally, ANPP committed to submit the change requests for Units 1 and 3 as part of the next reload Technical Specification submittals.

For Unit 1, the supporting analyses have been completed.

This amendment request satisfies our previous commitment to modify the Unit 1 Technical Specifications.

10-4

10.4 E TION OF TECHNICAL SPECIFICATION AFFECTED Limiting Condition for Operation 3.2.4 (DNBR Margin)

~

~

~

~

~

~

~

DE RIPTION OF THE PROPOSED AMENDMENT This proposed amendment will revise Technical Specification Figures 3.2-2 and 3.2-2A.

These figures provide DNBR margin limits for various configurations of COLSS and CEACs inoperable.

The changes are necessary to ensure that the Technical Specifications are consistent with the safety analyses performed for the Cycle 3 core design.

TIFICATION FOR THE PROPOSED AMENDMENT The proposed changes to the DNBR limit curves are necessary to ensure that the Technical Specifications are consistent with the safety analyses performed forthe Cycle 3 core design.

\\

10-5

11.0 STARTUP TESTING The planned startup test program associated wi.th core performance is outlined below.

The described tests verify that core performance is consistent with the engineering design and safety analysis.

The program conforms to ANSI/ANS-19.6. 1-1985, "Reload Startup Physics Tests for Pressurized Water Reactors" and supplements normal surveillance tests which are required by Technical Specifications (i.e.,

CEA drop time testing, RCS flow measurement, NTC verification, etc).

11.1.

LOW POWER PHYSICS TESTS 11.1.1 Initial Criticalit Initial criticality will be achieved by one of two methods.

By the fi.rst method, all CEA groups would be fully withdrawn with the exception of the lead r'egulating group which would be positioned at approximately mid-core.

The boron concentration of the reactor coolant would then be redu'ced until criticality 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 unrodded configuration and for a partially rodded configuration.

The measured CBC values will be verified to be within +1% b,k/k of the predicted values.

11-1

11.1.3 Tem erature Reactivit Coefficient

~

~

The isothermal temperature coeffscient (ITC) will be measured at the Essentially All Rods Out (EARO) configuration and at a partially rodded configuration.

The coolant temperature will be varied and the resulting reactivity change will be measured.

The measured values will be verifi'ed to be within +0.3 x 10 ak/k/

F of the

-4 predicted values.

11.1.4 CEA Reacti vit Worth 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.

All 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 one with me'dium 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 will be verified to be within +15 ppm/%%d ak/k of the predicted value.

11.2 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 monitored through use of an off-line 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

jr h

processed plant data against which they are normally calibrated.

If necessary, the power ascension will be suspended while necessary data reduction and equipment calibrations are.performed.

The following 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 will be. compared, to predicted distributions at two power plateaus.

There 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 within 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 (RMS) 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

/I 1

B.

The RMS of the 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), core average axial peak (F ),

and the 3-D power peak (Fq) will be verified to be within

+10% of their predicted values.

11.2.3 Sha e Annealin Matrix SAM and Boundar Point Power Correlation Coefficients BPPCC Verification The SAM and BPPCC values will be determined from a linear regression analysis of the measured excore detector readings and corresponding core power distribution determined from incore detector signals.

Since these values must be representative for a rodded and unrodded core throughout the cycle, it is desirable to use as; wide a range of axial shap'es 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 compiled 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 adial 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 COLSS constants will 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

I I

11.2.5 Tem erature Reactivit Coefficients at Power

~

~

The isothermal temperature coefficient (ITC) will be measured at approximately full power.

The ITC will be measured by changing coolant temperature, compensating with CEA motion, and maintaining power steady.

The ITC will be verified to be within +0.3 x 10 ak/k/'F of the predicted value.

11.2.6 Critical Boron Concentration The CBC will be determined for conditions of full power, equilibrium xenon.

The measured CBC will be verified 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 NET 0

J The resul'ts.of all tests will 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.

The results of this evaluation will be presented to the Plant Review Board.

Resolution will be required prior to subsequent power escalation.

If an unreviewed safety question is involved, the NRC will be notified.

11-5

Ij

REFERENCES SECTION

1.0 REFERENCES

(1-1)

"Palo Verde Nuclear Generating Station Unit No.

1, Final Safety Analysis Report," Arizona Public Service

Company, Docket No. 50-528.

~

S CT ON

.0 REFER NC S

None SECTION 3.0 REFERENC S

None S CTION 4. 0 R

F RENC S

161-00730-EEVB/LJM, "Final Surveillance Test Results for PVNGS-1, Cycle 1," January 8,. 1988.

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

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

(4-4)

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

July, 1981.

(4-5)

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

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

(4-6)

"Combustion Engineering Standard Safety Analysis Report (CESSAR)", Docket ¹STN-50-470F.

12-1

.)

(4-7)

"Palo Verde Nuclear Generating Station Unit No.

1, Final Safety Analysis Report," Arizona Public Service

Company, Docket No. 50-528, Section 4.2.4.

(4-8)

CESSAR

SSER2, Section 4.2.5, "Guide Tube Wear Surveillance".

12.5 SECTION 5,0 REFER NCES (5-1)

CENPD-153-P, Rev.

1-P-A, "INCA/CECOR Power Peaking Uncertainty," Hay, 1980.

(5-2)

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

12;6 J

SECTION.

6.0 REFERENCES

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.

CEN-160(S)-P, Rev.

1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2.

and 3", September, 1981.

(6-4)

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

Hay, 1988.

Enclosure 1-P to LD-82-054, "Statistical Combination of System Parameter 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.

12-2

li

)l l

CESSAR SSER 2 Section 4.4.6, Statistical Combination of Uncertainties.

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

"Palo Verde Nuclear Generating Station Unit No. 1, Final Safety Analysis Report," Arizona Public Service

Company, Docket No. 50-528.

"CESSAR, Combustion Engineer ing Standard Safety Analysis Report," Docket No. 50-470.

(7-3)

. "CESEC, Di.gital. Simulation of a.Combustion Engineering Nuclear Steam Supply System,"

December, 1981, Enclosure 1-P to LD-82-001, January 6,

1982.

CEN-160(S)-P, Rev 1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3," September, 1981.

(7-5)

CENPD-188-A, "HERMITE Space-Time Knetics," July, 1975.

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

(7-7)

CENPD-206-P, "TORC Code Verification and Simplified Modeling Methods," January 1977.

CENPD-183, "Loss of Flow - C-E Methods for Loss of Flow Analysis," July 1975.

12-3

(7-9)'61-00321-JGH/LJM, "Palo Verde Nuclear Generating Station (PVNGS) Unit 1 Docket No.

STN 50-528 (License NPF-41)

Submittal of the Reload Analysi.s Report for Unit 1 Cycle 2," June 29, 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 Liscense 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. Ei.senhut from A. E. Scherer, Letter No.

LD-82-040, March 31, 1982:

S CT ON 8 0 R

NC S-ECCS L S (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) 161-00321-JGH/LJM, Dated June 29,

1987, "Submittal of the Reload Analysis Report for Unit 1 Cycle 2".

"Issuance of Amendment No.

24 to Facility Operating License No. NFP-41 for the Palo Verde Nuclear Generating Station Unit No.

1 (TAC Nos.

65460, 65461,

65462, and 65691 through 65706),"

E. A. Licitra, October 21, 1987.

12-4

a

(8-4)

CENPD-132-P, "Calculative Methods for the C-E Large Break LOCA Evaluation Hodel", August 1974.

CENPD-132-P, Supplement 1-P, "Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model", December 1974.

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

(8-5)

CENPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1974.

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

February 1975.

CENPD-135-P, Supplement.4P. "STRIKIN-II, A Cylindrical Geometry Fuel.Rod Heat Transfer Program",

August 1976.

(8-6)

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

(8-7)

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

July, 1981.

(8-8)

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

(BGSE), "Safety Evaluation of CEN-161 (FATES 3),"

March 31, 1983.

(8-9)

ANPP-33609-EEVB/KLH, dated September 30, 1985, "Limiting Small Break LOCA Analysis - Additional Information".

(8-10)

DELETE (8-11) 161-00890-EEVB/BJA, dated March 16,

1988, "Proposed Technical Specifications Change - LPSI Flow Requirements".

161-01155-EEVB/BJA, dated July 6,

1988, "LPSI Flow Requirements".

12-5

I

4 Letter from the NRC, dated October 17,

1988, "Amendment to PVNGS-1, 2 5 3 Operating Liscense for LPSI Flow".

12.9 SECTION

9.0 REFERENCES

(9-1)

CEN-304-P, Rev. 01-P, "Functional Design 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.

'C (9-3)

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

(9-4)

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.

CEN-312"P, Rev. Ol-P, "Overview Description of the Core Operating Limit Supervisory System (COLSS)",

November, 1986.

12.10 SECTION

10.0 REFERENCES

NONE 12.11 SECTION

11.0 REFERENCES

(11-1)

ANSI/ANS-19.6.1-1985, "Reload Startup Physics Tests for Pressu} ized Mater Reactors".

I (11-2)

CEN-319, "Control Rod Group Exchange Technique,"

November 1985.

12-6

IO

~

l t

,