ML17303A722

Reload Analysis Rept for Palo Verde Nuclear Generating Station Unit 2,Cycle 2

ML17303A722
Person / Time
Site:	Palo Verde
Issue date:	12/03/1987
From:	ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
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NUDOCS 8712080335
Download: ML17303A722 (192)

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ATTACHMENT 1 PVNGS UNIT 2 CYCLE 2 87)P0Qo33 05000529 871203 DR gDOCV, 0 PDR P

0 J.

i

RELOAD ANALYSIS REPORT FOR PALO VERDE NUCLEAR GENERATING STATION UNIT-2 CYCLE 2 TABLE OF CONTENTS PAGE 1.

INTRODUCTION AND

SUMMARY

2.

OPERATING HISTORY OF THE REFERENCE CYCLE 2-1 3.

GENERAL DESCRIPTION 3-1 4.

FUEL SYSTEM DESIGN 5.

NUCLEAR DESIGN 5-1 6.

THERMAL-HYDRAULIC DESIGN 6-1 7.

TRAiNSIENT ANALYSIS 7-1 8.

ECCS ANALYSIS 8-1 9.

REACTOR PROTECTION AND iMiONITORING SYSTEM 9-1 10.

TECHNICAL SPECIFICATIONS 10-1 11.

STARTUP TESTING 11-1

t i

INTRODUCTION AND

SUMMARY

This report provides an evaluation of the design and performance of Palo Verde Nuclear Generating Station Unit 2 (PVNGS-2) during its second cycle of operation at 100%%d rated core power of 3800 NWt and NSSS power of 3822 MWt.

Operating conditions for Cycle 2 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 and C assemblies, along with fresh Batch 0 assemblies.

The Cycle 1 termination burnup has been assumed to be between 15,744 and 16,512 MWD/T.

The first cycle of operation will hereafter be referred 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 l-l. A review of all postulated accidents and anticipated operational occurrences has shown that the Cycle 2 core design meets these safety criteria.

The Cycle 2 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'he status of the postulated accidents and anticipated operational occurrences for Cycle 2 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.

0 Ol

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 2 are summarized in Section 10 and described in greater detail in separate license amendment applications.

Modifications to the Core Protection, Calculator (CPC)

System are being made to improve performance and reflect the Cycle 2 core configuration.

Algorithm changes are a result of the CPC Improvement Program (CIP) and are summarized in Section 9.

The concurrent data base changes are a result of plant-specific application of CIP to PYNGS-2 Cycle 2.

2.0 OPERATING HISTORY OF THE REFERENCE CYCLE The plant is currently in its first fuel cycle which began with initial critical.ity on April 18, 1986.

Power Ascension began on April 22, 1986 and on September 19, 1986 the unit was declared in commercial operation.

It is presently estimated that Cycle 1 will terminate on or about February 21, 1988.

The Cycle 1 termination point can vary between 15,744 NWD/T and 16,512 NWD/T to accommodate the plant schedule and still be within the assumptions of the Cycle 2 analyses.

<1

3.0 GENERAL DESCRIPTION The Cycle 2 core will consist of those assembly types and numbers listed in Table 3-1.

Sixty-nine Batch A assemblies and thirty nine Batch B will be removed from the Cycle 1 core to make way for 108

fresh, Batch D assemblies.

Sixty-nine Batch B and all Batch C

assemblies now in the core will be retained.

The reload batch will consist of 32 type DO assemblies, 20 type Dl assemblies with 16 burnable poison shims per assembly, 8 type D2 assemblies with 16 burnable poison shims per assembly, 16 type D3 assemblies with 16 burnable poison shims per assembly, 4 type D4 assemblies with 12 burnable poison shims per assembly, and 28 type D5 assemblies with 12 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 2, showing fuel type and location, is displayed in Figure 3-3, Figure 3-4 displays the beginning of Cycle 2 assembly average burnup distribution.

The burnup distribution is based on a Cycle 1 length of 16,512 MHD/T.

Control element assembly patterns and in-core instrument locations will remain unchanged from Cycle 1 and are shown in Figure 3-5 and-Figure 3-6.

i

TABLE 3-1 PALO VERDE NUCLEAR GENERATING STATION UNIT 2 Cycle 2 Core Loading Assembly Desig-nation B-Lo B-Hi C

C-Lo DO Dl D2 D3 D4 D5 TOTAL Number of Assemblies 24 40 24 32 20 16 28 241 Fuel Rods per Assembly 208 12 208 12 224 12 208 12 184 52 168 52 168 52 168 52 172 52 172 52 Initial Enrichment (w/o U-235) 2.78 1.92 2.78 1.92 3.30 2.78 3.30 2.78 4.02 3.57 4.02 3.57 4.02 3.57 3.57 3.09 3

~ 57 3.09 3.57 3.09 Number Shims/

Assembly 16 16 16 16 16 16 12 12 Initial Shim Loading (gm B10/in)

.018

.025

.012

.022

.020

.022

.008

.020 Total Number of Fuel Shim Rods Rods 4992 384 288 9360 720.

540 8960 480 4992 384 288 5888 1664 3360 320 1040 1344 128 416 2688 256 832 688 48 208 4816 336 1456 54300 2576

ASSEMBLY TYPE NUMBER OF ASSEMBLIES 69 FUEL ENRICHMENT W/T X u235 1.92 No. OF FUEL No. OF SHIM RODS PER RODS /

ASSEMBLY ASSEMBLY

'236

, gm B"0/IN.

BLO Cp CLO 1.92 2.78 1.92 2.78 2.78 3.30 2.78 3.30 12 208 12 208 12 224 12 208 16 16 16 0.01842 0.02532 0.01151 BLp BHI AND CLp Cp WATERHOLE 8

LOWER ENRICHED FUEI. PIN Q

HIGHER ENRICHED FUEL PIN g

SHIM PIN ARl20NA Palo Verde Nuclear Generating Station FiRST CYCLE ASSEMBLY FUEL LOADINGS WATERHOLE AND SHIM PLACEMENT Figure 3-1

~

i 0,

BS-NTCH 00- 2 RSSEI%LIES SN-NTCH D3-10 RSSBB IEB Q

4.o2 w/0 u-aa H

3.6V W/0 U-236 Q

o.sv w/0 u-236 H

. o.m w/0 u-236 84C-RL203 MIH PIN

.022 Qtn 8-fo/IN Ra-anCH 01-2O RSSae IES RB-fBTCH 04-4 ASSEI'KIES Q

4.02 w/0 u-236 8

o.sv w/0 u-~

B4C-~ SHIH PIN

.022 gm B-lo/IN Q

o.sv w/0 u-236 0

o.ofa W/0 U-K%

B4C~3 SHIH PIN 008 pm B-fo/IN Ra-BATCH 02-8 RSSE%LiES RS-KCH 05-28 RSSBB IES Q

4.02 w/0 u-236 H

3.67 W/0 U-236 B4C-~ SHIH PIN

.O20 ~ B-fO/IN Q

o.sv w/0 u-236 H

o.fxr w/0 u-236 84C-RL203 SHIH PIN

.020 pe 8-fO/IN AAIEONA Palo Verde Illoclear Generating Station SECOND CYCLE ASSEMBLY FUEL LOADINGS WATERHOLE AND SHIM PLACEMENT l9Ur8 3-2

~

Bx 00 Bx C

DD C

8 Bx 00 00 00 02 01 Bx Bx 00 OD C

Bx 01 01 C

8 03 05 C/

C/

05 01 Bx Bx 02 03 03 C/

05 00 05 c/

8 05 Bx 8

OD Bx 00 01 Bx C/

01 05 Bx C/

05 05 Bx Bx 04 04 Bx ARIZDNA Palo Verde ucfeer Generating Station CYCLE 2 FUE L MANAGEMENT Figure 3~3

0'

I 16757 2

0 3

4 17431 0

5 18731 6

0 7

9944 8

9 10 9940 18889 0

11 12 18220 0

)3 13505 14 15 16 12796 0

17 18832 18 19 18731 0

20 17262 21" 0

22 18377 23 0

24 25 13874 0

26 0

27 28 13505 0

29 13622 30 0

31 32 17094 0

33 18853 34 35 36

)6757 9944 0

37 18377 38 0

39 11293 40 0

4) 42 17087 0

43 0

44 45 46 9940 12796 0

17094 48 0

49 50 18772 0

51 18899 52 53 54 17431 18889 0

55 13874 56 0

57 17087 58 0

59 60 18901 0

61 0

62 0

63 64 I8832 0

65 18853 66 0

67 68 18904 0

69 18850 F I SLIRE 3-4 RSSEMBI Y RVERRGE BURNUP ( MWQPj RT BQC2

0 0

0

5 LEAD REGULATINGBANK 4

SECOND REGULATING BANK 3

THIRD REGULATINGBANK

'2 FOURTH REGULATING BANK 1, LASI REGULATINGBANK 8

SHUTDOWN BANK8 A

-SHUTDOWN BANKA P2 PLR GROUP 2 Pg PLR GROUP 1 S

SPACE CEA LOCATIONS

'10 I'I 14 15 ld 17 8

22 22 P2 24 8

2$

25 17 P2 41 2

8 21 4

4d 47 SO 51 Pq 5d Sd 58 P) 50

$0 8

57 4

70 71 72 74 75 8

77 7$

80

$2 52

$4 85 85

$ 7 00 92 94 9$

94 97 P2 98 100 101 102 102 104 105106 107 PZ 108 109 110 8

112 112 114 11$

115 117 115 119

'I20 11'I 122 112 Pg 124 125 115 117

'I28 129 120 121 112 124 115124 127 128129 140 141 tll t42 8

144 145 lid 147 14$

ld5 A

181 140 P2 1$ 4 1$ 2 150 151 IS1 152 IdT td$

189 170 1$2 t$4

'I$5 184 154 171 157 155 tdd 157 171 172 74 1$ $

180 190 158 175 4

191 159 150 1dt Pp

'I 75 ITT 17$

102 192 194 Pg 161 152 S

179 180 195 194 107 198 190 200 8

201 202 8

02 204 205 8

205 107 208100 210 2'II 212 212 214 115 Pg 215 217 21$

119 P2 220 221 222 124

~

22$

125 117 229 220 221 222 224 225 225 117 22$

3 219 140 ARIZDNA Palo Verde Nucfeer Generating Stetlnh CEA BANKIDENTIFICATION Figure 3-5A

0 0

P R

P R

P R

P R

P R

P R

ARIzo iliA Palo Var'da Nuclear Gencratiny Station CEA PATTERN Figure 34B

t

~

i

BOX INSTR.

9 10 11 3

12 24 6

4 5

13 14 4

25 28 15 7

1 17 18 31 9

32 33 10 47 48 50 51 13 53 14 39 40 41 12 55 56 15 58 16 59 60 51 17 18 70 71 72 73 74 75 75 19 78 80 81 20 83 84 21 91 24 25 26 101 102 27 104 105 106 108 109 110 111 112 113 114 29 115 117 118 119 120 30 121 31 122 123 124 125 32 128 127 33 131 132 134 135 36 13$

141 142 146 14S 147 148149 151 162 153 154 155 156 167 159 160 1$ 1 41 182 163 1$6 42 170 171 172 44 173 174 175 176 178 179 46 1$0 1$ 1 47 184 185 186 48 189 49 191 50 192 193 194 196 197 198 51 199 200 201 207 52 53 210 211 213 64 214 215 55 216 58 217 218 56 219 231 59 220 57 221 222 223 234 60 235 61 236 238 239 240 241 ARIZONA Palo Verde Nuclear Generating Station INSTRUMENT LOCATIONS Figure 34

FUEL SYSTEM DESIGN MECHANICAL DESIGN

'With the exception nf'he design features listed below the mechanical design of the Batch D reload fuel assemblies is identical to the design of the Core I Batch C fuel assemblies that were modified to increase the shoulder gap from 1.682 inches to 2.382 inches.

Reference 4-1.

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

The design features incorporated into Batch D were made to improve fuel handling, to improve the fabricabi lity and auality of the fuel and to improve the burnuo caoability of the poison rods.

The specific chan~es are discussed in the following paragraphs.

1.

The design of the upper end fitting hold down plate was modified to eliminate the potential interference between the spent fuel handling tool and the hold down plate under worse case tolerance conditior s, Reference 4-2.

I. shnuld be noted that all of the fuel bundles utilizing the previous design have been handled at the Palo Verde site without revealing any interfe~ence problems.

i Qsoection envel ope =,r hc, fuji bunrll e a -s amb ly has beon changed from a square of 8.23 inches per side for all bundle with exception of the vicinitv around the upper most grid where a square o" 8.250 inches per s.'de is permissible to a sauare of 8.290 inches per side for the enti re length o

the

~'uel bundle.

Th's ch:rge will not a

~ct sa==.."

ual assembly oow, handlinn or other considerations.

Furthermore, a

y t

I t

reduction in the handling of a fuel bundle assembly for the purpose of placing it within the inspection envelope reduces the possibility of introducing unexpected mechanical damage to the various components and/or connections in the bundle.

3.

The poison rod assembly design was modified by replacing the solid Zircaloy-4 spacers with hollow Zircaloy-4 tubes.

This provides greater internal void volume which enables higher burnups with poison rods with higher 8-10 loadings while reducing end of life internal pressure.

A dr aft copy of an EPRI-sponsored report dealing with the phenomena of interpellet gap formation and clad collapse in modern PWR fuel rods was submitted to the NRC as Attachment 5 of Reference 4-3.

The final version of this report was subsequently issued as Reference 4-4.

A synopsis of the report focusing on C-E manufactured modern fuel was also submitted to the

NRC, as Attachment 4 of Reference 4-3.

The conclusion and recommendation of this synopsis was that clad collapse analyses are not necessary for modern C-E manufactured fuel because of the absence of gaps between pellets'he NRC

'concurred with this approach in Reference 4-5, provided:

a)

No new data has been developed relative to Reference 4-4 which would invalidate the bases for asserting that clad collapse analyses need not be performed and that augmentation factors are negligible.

b)

The fuel rod manufacturing process is either the same as that used to demonstrate no interpellet gaps or, if changed, not changed in a way that would adversely affect the clad collapse and augmentation factor analysis results.

~,

C-E has performed a review to address these items and has concluded that there is no new data that invalidates the bases of Reference 4-4 and that the fuel types'to be inserted in Cycle 2 were manufactured using the same process as was used on the fuel that demonstrated no interpellet gap formation.

Since the provisions of the NRC's concurrence have been satisfied, no cycle specific clad collapse analysis was performed for Cycle 2.

Since clad collapse has been removed as an issue for modern C-E fuel, discussion of clad collapse will not be included in the Reload Analysis Report in subsequent fuel cycles.

4.2 GUIDE TUBE NEAR Twenty of the fuel assemblies that had CEA's located in them during Cycle 1 at Palo Verde Unit 1 will be inspected for guide tube wear.

That inspection is part of the required licensing procedures required by the NRC for all plants after the first cycle of operation (References 4-10 and 4-11).

A similar program will also be performed on Unit 2 during the first refueling outage (Reference 4-12).

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

4.3 THERMAL DESIGN k

The thermal per formance of composite fuel pins that envelope the pins of fuel batches B,

C and D present in Cycle 2 have been evaluated using the FATES3A version of the C-E fuel evaluation model (References 4-6 and 4-7) as approved by the NRC (Reference 4-8).

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 2.

0

~

j

4.4 CHEMICAL DESIGN The metallurgical requirements of the fuel cladding and the fuel assembly 'structural members for the Batch 0 fuel are identical to those of the fuel batches included in Cycle 1.

Thus, the chemical or metallurgical performance of the Batch 0 fuel will remain unchanged from the performance of the Cycle 1 fuel (Reference 4-7).

4.5 SHOULDER GAP ADEQUACY Conservative estimates based on measured shoulder gap data acquired from other plants operating with 16 x 16 fuel supplied by C-E indicate that the fuel has adequate shoulder gap for Cycle 2

operation.

It is expected that the conservatism of the estimates in shoulder gap adequacy will be confirmed after completion of the shoulder gap measurements of the Cycle 1 fuel from Palo Verde Unit 1.

Those shoulder gap measurements are reouired by the NRC as part of the licensing procedures.

I

NUCLEAR DESIGN PHYSICS CHARACTERISTICS The Cycle 2 core makes use of a low-leakage fuel management

scheme, in which previously burned Batch B assemblies are placed on the core periphery.

Most of the fresh Batch D 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 1 endpoint at 16,512 MWD/T, the Cycle 2 reactivity lifetime for full power operation is expected to be 16,945 MWD/T.

Explicit evaluations have been performed to assure applicability of all analyses to a Cycle 1 termination burnup of between 15,744 and 16,512 MWD/T and for a Cycle 2 length up to 17,945 MWD/T.

Characteristic physics parameters for Cycle 2 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 2 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.

0

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 2

and the Reference Cycle.

5. 1.2 Power Distribution Figures 5-3 through 5-5 illustrate the calculated All Rods Out (ARO) planar radial power distributions during Cycle 2.

The one-pin planar radial power peaks presented in these figures represent the mid-plane of the core.

Time points at the beginning, middle, and end of cycle were chosen to display the variation in maximum planar radial peak as a function of burnup.

Radial power distributions 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 Bank 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 rance from 1. 14 at BOC2 to 1. 11 at EOC2.

0

5.2 SAFETY RELATED DATA 5.2.1 Au mentation Factors As indicated in reference 5-1, the increased power peaking associated with the small interpellet gaps found in modern fuel rods (non-densifying fuel in pre-pressurized tubes) is insignificant compared to the uncertainties in the safety analyses.

The report concluded that augmentation factors can be eliminated from the reload analyses os any reactor loaded exclusively with this type of fuel.

This discussion of the elimination of the augmentation fac.ors was used by BGSE in Reference 5-2 and accepted by the NRC in Reference 5-3.

Augmentation factors have been eliminated for Cycle 2.

5.3 PHYSICS ANALYSIS METHODS 5.3.1 Anal tical 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 5-4.

ROCS-DIT with the MC module wil'> be used.

ROCS-DIT and the MC module have been approved for this application in Reference 5-5.

5.3.2 Uncertainties in Measured Power Distributions The planar radial power distribution measurement uncertainty of 5.3

, based on Reference 5-4, will be applied to the Cycle 2

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 1.

r

~

5.3.3 Nuclear Desi n Methodolo The Cycle 2 nuclear design was performed with two and three dimensional core models using the ROCS and PDg computer codes employing DIT calculated cross sections.

The ROCS-DIT was described in Reference 5-5.

TABLE 5"1 PVNGS"2 CYCLE 2 NOMINAL PHYSICS CHARACTERISTICS Dissolved Boron Units Reference

~Cele

~Cele 2

Dissolved Boron Concentration for Criticality, CEAs Withdrawn, Hot Full Power Equilibrium Xenon, BOC PPM 657 1116 Boron Worth Hot Full Power, BOC Hot Full Power, EOC PPM/%%d PPM/%%d 91 83 120 95 Moderator Tem erature Coefficients

.Hot Full Power, Equilibrium Xenon Beginning of Cycle 10-4ap/

F End of Cycle 10-46,p/

F

-1.0

-2.5

<<0.5

-2.0 Do ler Coefficient Hot Zero Power, BOC Hot Full Power, BOC Hot Full Power, EOC 10-5hp/

F 10-5hp/

F 10-5hp/

F

-1.5 1 ~ 3

-1.5

-1'. 8

-].4

-1 7

Total Dela ed Neutron Fraction eff BOC EOC 0.0073 0.0053

.0063

.0052 Prom t Neutron Generation Time 1*

BOC EOC 10-6 sec 10-6 sec 28.1 31.3 21.7 27.7

TABLE 5-2 PVNGS-2 CYCLE, 2 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT FULL POWER STEAM LINE BREAK,

%%d, END"OF-CYCLE (EOC)

Reference

~Cele

~Cele 2

Worth of all CEAs Inserted "15.0

-16.0 Stuck CEA Allowan'ce

+4.9

+4.3 Worth of all CEAs LessHighest Worth CEA Stuck Out

-10.1 1 1 ~ 7 Full Power Dependent Insertion Limit CEA Bite

+0.2

+0.2 Calculated Scram Worth

-9.9

-11.5 Physics Uncertainty

+1.0

+1.2 Other Allowances (losses due to voiding)

+0.4

+0.1 Net Available Scram Worth

-8.2

-10.2 Scram Worth Used in Safety Analysis

-8.2

-10.0

TABLE 5"3 PVNGS-2 CYCLE 2 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER,

%hp Be innin of C cle End of C cle Regulating CEAs Reference Cycle Cycl e 2

Reference Cycl e Cycl e 2

Group 5

~ 32

-.25

~ 32

-.29 Group 4

".47

-.39

".47

-.46 Group 3

-.79

-.67 99

.74 Note:

Values shown assume sequential group insertion.

~

1.00 0.90

.0.00

~ 4 inill

~-4n ln Al I

'O in

'AJ nin m

0.70 0,60

0. 50 tJ >O cn> l Q il

'I.'I.) ~

Vs

)v ~

IQ lJh ~

i fic 0.40 c

0.30 0.20 j3 O.lO CoL S

oU s

roF saRY/ae 3:IJ I

I I

I I

I 0.00 I

l bl I

l~ I~

l l20 90 (io lo 0

I I io IS I I'.tff I At ttSI ttT l20 9 0 6l

!Oll l.lN COQUE SaRY l~

n lait 0

9ll 60

)0 ll i i~i l5ll l 20 90 (io 1ll 0

Iso I20 90 60 30 0

Cf:h MlTttAHAMAt.- It(ClttS

l

Q.Oa-Q.aa 0.70 cf Q.GO 0.60 0.<a 0.3a l

0.20 0.10-MR R A th mtll O

UNACCEPTABLE OPERATION RESTRICTEO OPERATION m

lA m

Cl Ul CD C) m lA 0

IGa

)co laa I2O IIo caa oa OO Vo Ga So Co ao Za So 0

1 I'AIITI.ENGTIICEA I'OSITION, INCIIESWITIIOAAlVN

~,

Fxg

=

1. 47 r ~ eox 1 n 0.46 0.42 12 0.37 6

7 0.93 0.92 13 14 1.06 1.26 2

0.69 8

0.96 15 I.ll 3

4 0.51 0.78 9

10 0.83 1.24 l6 17 1.25 0.97 18 0.42 19 I.II 0.98 1.27 22 1.00 23 1.29 24 25 1.17 1.27 34 0.37 43 0.68 52 0.51 26

0. 93 35 0.92 44 0.96 53 0.83 27 I. 06 36 1.26 45 1.11 54 1.24 28 1.27 37 1.00 46 1.28 55 1.17 29 30 1.12 1.23 38 39 1.22 1.14 47 48 1.11 1.16 56 57 1.25 1.05 31 1.12 40 1.16 49 0.91 58 1.12 32 33 1.25 0.98 41 42 1.05 1.18 50 51 1.12 0.90 59 60 0.88 1.15 61 0.78 62 1.24 63 0.97 64 1.27 65 0.97 1.18 67 0.90 68 69 1.15 0.90 F I GURE 5-2 FISSEMBl Y RELFlTI VE POWER DENSITY UNROOOEO F1T HFP BOCA wP Eg XE

1 0.36 2

0.63 0.48 0.69 Fxg

=

1. 47 IN BcjX 64 5

0.38 0.81 7

0.83 8

0.87 0.78 10 1.08 0.42 12 0.95 13 14 0.95 1.24 15 1.05 16 1.26 17 0.95 18 0.38 19 20 0.95 0.91 21 22 1.29 I

~ 01 23 1.32 24 1.17 25 1.36 26 0.80 27 28 0.95 I

~ 28 29

1. 12 30 1.34 31 1.15 1.33 33 1.03 34 0.36 35 0.83 36 37 1.24 1.01 38 39 1.34 1.19 1.32 41 1.12 42 1.30 43 0.63 44 0.87 45 46 1,05 1.31 1.15 48 1.32 49 I

~ 01 50 1.26 51 0.99 52 0.48 53 0.78 54 1.26 55 56 57 1.17 1.33 1.12 58 1.26 59 60 0.98 1.28 61 0.69 62 1.08 63 64 65 66 0.95 1.36 1.03

).30 67 0.99 68 69 1.28 0.99 FIGURE 5-4 FlSSEMBLY REl RT T. VE POWER OENS ITY UNROODEO 1=IT I-IFP MOC2 wi Eg XE

Fxg

=

1

. 48 IN 80X 21 0.42 0.84 0.41 7

0.86 8

9 0.90 0.81 10 1.09 2

3 0.68 0.53 0.73 11 0.46 0.98 13 0.98 14 1.30 15 1.05 16 1.31 17 0.97 18 0.42 26 0.84 0.98 27 0.98 20 0.94 28 1.35 21 1.35 29 I. 13 1.02 30 1.34 23 24 25 1.29 1.13 1.36 31 32 33 1.10 1.27 0.99 34 0.41 35 36 0.86 1.30 37 1.02 38 1.34 39 1.14 40 1.30 41 1.05 42 1.22 43 0.68 44 0

F 90 45 1.05 1.29 47 1.10 48 1.30 49 50 0'7 1.19 51 0.92 52 0.53 53 54 0.81 1.31 55 1.13 56 1.27 57 1.05 58 59 60 1.19 0.90 1.13 61 0.73 62 63 1.09 0.97 64 1.36 65 0.99 66 1.22 67 0.92 68 1.13 69 0.87 FIGURE 5-5 RSSEMBLY RELRT I VE POWER OENS ITY UNROOOEO RT HPP EOC2 wZ Eg XE

0

Fxy =

1. 45 IN BOX 62 0.41 0.95 0.36 0.90 8

9 0.92 0.81 10 1.29 2

3 4

0.68 0.50 0.80 11 12 13 0.46 1.15 1.06 14 1.23 15 16 17 0.97 1.24 0.97 18 0.41 19 1.15 20 21 0.97 1.25 22 0.95 23 24 25 I.27 I.18 I.32 26 0.95 27 28 29 30 1.06 1.24 0.98 1.21 31 1.12 32 33 1.31 1.00 34 0.36 35 0.89 36 37 38 1.23 0.95 1.20 39 1.14 40 1.20 41 1.09 42 1.26 43 0.68 44 0.92 45 46 0.97 1.27 48 1.20 49 50 51 0.93 1.19 0.94 52 0.50 53 0.81 54 55 56 1.23 1.18 1.31 57 1.09 1.19 59 0.92 60 1.21 61 0.80 62 1.29 63 64 0.97 1.32 1.00 66 1.26 0'4 68 1.21 69 0.84 FIGURE 5-6 FlSSEMBLY RELRT IVE POWER GENS ITY RT HFP BOCA WITH PLCER '

INSERTED wP Eg XE

~

1 2

0.40 0.74 3

0.53 0.83 Fxy =

1. 589 j:N BOX 12 5

6 0.46 1.05 0.99 8

I. 02 9

0.85 10 1.33 0.51 12 13 14 1.27 1.17 1.37 15 1.15 1.26 17 0.95 0.46 26 1.04 19 1.27 27 1.17 20 21 1.08 1.39 28 29 1.39 I.18 22 23 1.03 1.30 30 31 1.25 1.06 1.09 32 1.08 25 1.13 33 0.55 34 35 0.40 0.99 36 37 38 39 1.36 1.02 1.25 1.11 40 1.09 41 0.90 42 0.97 43 0.73 44 I.OI 45 46 1.15 1.30 47 48 49 1.05 1.09 0.81 50 I.OI 51 0.79 52 53 0.52 0.85 54 1.26 55 56 57 58 1.09 1.08 0.90 1.01 59 0.79 60 1.06 61 62 0.83 1.33 63 64 65 66 67 0.95 1.13 0.55 0.97 0.79 68 1.06 69 0.81 FIGURE 5-7 RSSEMBLY REl RT IVE POWER OENS ITY RT HPP BQC2 WITH BRNK 5 INSERTEO wl Eq XE

~

Fxg

=

1. 50 IN EIOX 12 1

0.39 5

6 7

0.47 1.05 0.98 2

0.73 8

0.98 0.84

'I 0 1.34 3

4 0.53 0.84 11 12 13 14 0.52 1.29 1.17 1.32 15 1.0'I 16 1.23 17 0.95 18 0.46 19 1.29 20 21 1.07 1.36 22 1.00 23 1.27 24 25 1.15 26 1.05 27 28 29 30 I.17 1.35 1.04 1.23 31 1.07 32 33 1.12 0.57 34 0.39 35 0.97 36 1.32 37 38 I

~ 00 1.22 39 1.13 40 1.14 41 0.96 42 I.05 43 0.73 44 0.98 45 46 47 48 1.00 1.27

'I.07 1.14 49 0.87 50 51 1.09 0.85 52 0.53 53 0.84 1.23 55 56 57 1.09 1.12 0.96 58 1.09 59 0.85 60 1.12 61 0.84 62 I ~ 33 63 64 65 66 0.95 1.15 0.57 1.04 67 0.85 68 69 1.12 0.78 F I GLlRE 5-8 RSSEMBL Y RELRT IVE POWER OENS I TY RT HFP BOC2 W ITH BRNK 5 RND PLCER '

INSERTED wP Eg XE

0',

Fxg

=

1. 5 1 6 IN EMBOX 54 5

0.40 6

0.82 0.38 7

0.82 2

3 0.65 0.49 8

9 0'4 0.77 0.72 10 I. IO 11 0.43 12 0.96 13 0.95 1.26 15 0.90 16 1.30 17 0.96 18 0.40 0.96 20 0.90 21 1.32 22 0.98 23 24 1.29 1.14 25 I.44 26 27 0.82 0.95 28 1.32 29 0.99 30 1.35 31 1.12 32 1.36 33 1.03 0.38 43 0.65 52 0.49 61 0.72 5:

0.77 54 1.30 62 63 1.10 0.96 35 36 0.82 1.26 44 45 0.84 0.91 37 0.98 46 1.29 55 1.14 64 1.44 38 1.35 1.12 56 1.36 65 1.03 39 1.19 48 1.40 57 1.12 1.35 40 41 1.40 1.12 49 50 1.03 1.32 58 59 1.32 0.95 67 68 0.98 1.21 42 1.35 51 0.98 60 1.21 69 0.81 F j:GLIRE 5 9 RSSEMBLY RELRTIVE POWER OENSITY RT HPP EQC2 WITH PLCER '

INSERTEC3 wl Eq XE

Fxg

=

1. 602 IN BOX 2 1

0.42 2

0.71 5

6 7

8 0.44 0.92 0.92 0.93 3

0.52 0.81 0.75 10 1.12 ll 12 13 0.49 1.08 1.07 14 1.42 15 1.08 16 1.33 17 0.93 18 0.44 19 1.08 20 21 22 1.02 1.50 1.06 1.32 24 1.05 25 1.21 26 0.92 34 35 0.42 0.92 27

, 28 29 30 31 1.07 1.50 1.21 1.42 1.06 36 37 38 39 40 1.42 1.06 1.42 1.16 1.27 32 41 0.92 33 0.54 42 1.03 43 0.71 44 0.93 45 1.08 46 47 48 49 1.32 1.06 1.27 0.90 50 1.12 51 0.82 52 53 0.52 0.81 54 55 1.33 1.05 57 58 0.92 I.IZ 59 0.82 60 1.07 61 62 0.75 1.12 63 64 65 66 67 0.93 1.21 0.54 1.03 0.82 68 1.07 69 0.80 F j:SLIRE 5 10 RSSEMBL Y RELRT IVE POWER DENS ITY RT HFP EQC2 WITH BRNK 5 iNSERTED wY Eg XE

~

Fxy =

l. 537 IN BOX 2 l 0.42 5

6 7

0.45 0.92 0.90 2

0. 71 0.89 3

0.53 9

0.80 4

0.76 10 1.13 18 0.45 11 12 0.49 1.09 19 20 1.09 I

~ 00 13 14 1.05 1.36 21 22 1.44 1.03 15 0.94 23 I.'29 16 1.29 24 1.05 17 0.94 25 1.23 26 0.92 27 28 29 30 1.05 1.44 1.05 1.39 1.08 32 1.16 33 0.57 34 0.42 35 0.90 36 37 38 1.36 1.03 1.39 39 1.19 40 1.34 41 O.QQ 42 1.12 43 0.71 44 0.90 0.94 46 47 1.29 1.08 48 1.34 49 0.98 50 1.22 51 0.90 52 0.53 53 0.80 54 1.2Q 55 56 57 1.05 1.16 1.00 58 1.22 59 0.90 60 1.14 61 0.76 1.13 63 0.94 64 1.23 65 66 0.57 1.12 67 0.90 '.14 69 0.77 F IGLIRE 5-1 1

FISSEMBLY RELFITIVE POWER DENS lTY FIT HFP EOC2 W X TH BRNK 5 FIND PLCER '

INSERTED wi Eg XE

f 0,

6.0 THERMAL-HYDRAULICDESIGN 6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 2 at the rated power level of 3800 MMT have been performed using the TORC computer code described in Reference 6-1, the CE-1 critical heat flux correlation described in Reference 6-2, 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-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 by the Core Protection Calculators (CPC) and Core Operating Limit Supervisory System (COLSS) which, when used with the Cycle 2

DNBR limit of 1.24*, provide assurance at the 95/95 confidence/probability level that the hot rod will not experience DNB.

This Cycle 2

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 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 2

MSCU analysis.

Calculated using the methodology of Reference 6-5, as done for the reference cycle.

i

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%

MDNBR, is valid for bundle burnups up to 30,000 MWD/NTU.

This penalty is included in the 1.24 DNBR limit.

For assemblies 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 all assembly burnups expected for Cycle 2.

0'

TABLE 6-1 PVNGS-2 Cycle 2

Thermal H draulic Parameters at Full Power General Characteristics'otal Heat Output (Core only)

Fraction of Heat Generated in Fuel Rod Primary System Pressure Nominal 1nl et 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 aP 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)

Ceits MW$

10 Btu/hr psia 0F

-gpss 10 lb/hr 10 lb/hr ft 10 1 b/hr-ft PSl Psl BTU/hr-ft Reference

~Cele 3800 12,970 0.975 2250 565.0 445,600 164.0 159.0 0.039 2.61 15.9 56.7 184,400*

68,600*

~Cele 2

3800 12,970 0.975 2250 565.0 423%300 155.8 151.1 0.039 2.49 14.5 51.3 186,600*+

67,700**

Film Coefficient at Average Conditions BTU/hr-ft F

6300 6100 Average Film Temperature Difference F

29 31 Average Linear Heat Rate of Unden-sified Fuel Rod (Accounts for fraction of heat generated in fuel rod) kw/ft 5.4 5.5

TABLE 6-1 (continued)

General Characteristics Average Core Enthalpy Rise Maximum Clad Surface Temperature Engineering Heat Flux Factor Engineering Factor on Hot Channel Heat Input Unlts BTU/lb F

Refer ence

~Cc1 e 81.6 656 03*%'*

] 02*%*

~Cele 2

85.9

'56 1.03+

1

~ 03+

Rod Pitch, Bowing and Clad Diameter Factor 05*%%'.05+

Fuel Densification Factor (Axial) 1.002 1.002 NOTES:

Based on 1920 poison rods.

Based on 2576 poison rods.

• • • These factors were combined statistically with other uncertainty factors to define a new design limit on CE-1 minimum DNBR at the 95/95 confidence/probability level when iterating on power as discussed in Reference 6-5.

+

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 Cycle 2

DNBR limit provide assurance at the 95/95 confidence/probability level that the hot rod will not experience DNB.

Tech'pec

~ minimum flow rate.

0 0

7.0 NON-LOCA SAFETY ANALYSIS 7.0. 1 Introduction This section presents the results of the Palo Verde Nuclear Generating Station Unit 2 (PVNGS-2), Cycle 2 Non-LOCA safety analyses at 3800 MMt.

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 Limitina 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 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 2.

The DBEs chosen for analysis for each criterion are the limiting events with respect to that criterion.

The write-ups for those events presented consist of discussions of the reasons for the reanalyses, discussions of the causes of the

events, descriptions of the analyses performed, results, and conclusions.

7.0.2 Methods of Anal sis The analytical methodology used for PVNGS-2 Cycle 2 is the same as the Cycle 1 (Reference Cycle) methodology (References 7-1 and 7-2) unless otherwise stated in the event presentations.

Only methodology that has previously been reviewed and approved on the PVNGS dockets, the CESSAR docket, or on other dockets is used.

7.0.3 Mathematical Models The mathematical models and computer codes used in the Cycle 2

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

Plant response.for Non-LOCA Events was simulated using the CESEC III computer code (Peference 7-4).

Simulation of the fluid conditions within the hot channel of the reactor core and calculation of DNBR were performed using the CETOP-D computer code described in Reference 7-7.

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 Reference 7-10 and 7-11.

The number of fuel pins predicted to experience clad failure is taken as the number of pins which have a CE-1 DNBR value below 1.24.

The only exceptions are the Shaft Seizure and Sheared Shaft events

'or which the statistical convolution method, described in Reference 7-12, was used.

Reference 7-12 has been approved by the NRC and has been used for CESSAR and the PVNGS FSAR.

'l The HERMITE computer code (Reference 7-9) was used to simulate the reactor core for analyses which required more spatial detail than is provided by a point kinetics model.

Reference 7-9 has been approved by the NRC and has been used for CESSAR and the PVNGS FSAR.

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.

This method was approved for SONGS Units 2 and 3

(Reference 7-14).

Because of the similarity in the NSSS design, this methodology is applicable to PYNGS-2.

0

Input Parameters and Anal sis Assum tions Table 7.0-6 summarizes the core parameters assumed in the Cycle 2

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 (e.g.

Azimuthal Tilt, NTC, PDIL) 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 2 values.

Such assumptions resulted in more adverse consequences.

Those events presented which relied on VOPT protection credited the CPC software VOPT.

Events which have credited CPC trip protection have assumed instrument channel response times which are conservative relative to the Cycle 2 Technical Specificatic~s.

Conclusion All DBEs are evaluated for PVNGS-2, Cycle 2 to determine whether their resuli.s are bounded by the Reference Cycle.

Those events whose results were not bounded by the Reference Cycle and those events for which analysis methodology differs from the Reference Cycle methodology are presented herein.

All DBEs have results within NRC acceptance criteria.

F 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 Reference Cycle.

7. 1.4 Inadvertent 0 enin of a Steam Generator Safet Valve or Atmos heric

~Dum Valve The results are bounded by the Reference Cycle.

7.1.5 Steam S stem Pi inq 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.

This event credits the CPC software Variable Overpower Trip (VOPT).

The NRC accepted the CPC VOPT for SONGS Units 2 and 3 (Reference 7-14).

7. 1.5b Steam S stem Pi in Failures:

Post-Tri Return to Power The Steam Line Break event at zero power initial conditions was re-evaluated because the Cycle 2 Moderator cooldown reactivity insertion curve is more adverse than the Referece Cycle 1 curve.

~

Figure 7.1.5-1 compares the two curves.

In addition, a sweep-out volume of 119 ft before Safety Injection reaches the RCS was assumed for Cycle 2, which is more conservative than the 34.7 ft3 assumed for the Reference Cycle.

The effect of the more adverse reactivity insertion was accorrmodated for Cycle 2 by increasing the Shutdown Margin required by the Technical Specifications at zero power from 6%ap to 6.5@p.

The results of the Reference Cycle Steam Line Break event initiated at full power conditions bound Cycle 2 results.

7.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 7.2. 1 Loss of External Load The results are bounded by the Reference Cycle.

7.7.7 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.

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

~

i

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 ',e Reactor Coolant Pum Shaft Seizure/Sheared Shaft The results are bounded by the Reference Cycle.

7.4 REACTIVITY AND POWER DISTRIBL'TION ANOMALIES Uncontrolled CEA Withdrawal from a Subcri tical 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 single full-length and part-length CEA drop events are analyzed to determine the initial thermal margins that must be maintained by the Limiting Conditions for Operation (LCOs) such that the DNBR and Fuel Centerline-to-Melt (CTM) specified acceptable fuel design limits (SAFDLs) will not be violated.

The CEA position-related penalty factors for downward single CEA deviations of four-fingered CEAs have been set eoual to one (no penalty) in the Control Element Assembly Calculators (CEACs),

as described in Reference 7-8.

This applies to both full-length and part-length CEA downward deviations.

w t

Sufficient thermal margin will be maintained by the LCOs to compensate for the removal of CEA position related penalty factors for downward CEA deviations of 4-fingered CEAs.

The CEA position-related penalty factors for downward deviations of 12-fingered CEAs have been calculated such that the CPCs provide a

trip when necessary.

A part-length POIL has been added which restricts the part-length CEA insertion to less than 25~ for power levels greater than 50%.

From these initial conditions, the part-length single or subgroup drop inserts only negative reactivity (similar to a full-length single or subgroup drop event).

The method used to analyze the single CEA drop event is described in Reference 7-13.

For CEA subgroup

drops, the CEA position-related penalty factors for downward deviations are used by the Core Protection Calculators as in the Reference Cycle to provide a trip when necessary.

7.4.3. 1 Identification of Causes The CEA Hisoperation Events are defined as the inadvertent release of a single CEA or CEA subgroup causing it to drop into the core.

The occurrence of an electrical or mechanical failure in a CEA drive mechanism could result in a

CEA drop.

7.4.3.2 Anal sis of Effects and Consequences The single full length CEA drop is analyzed because this event requires the maximum initial margin to be maintained by LCOs.

This analysis considered several cases over the parameter space given in Table 7.0-6.

The case chosen for presentation is typical of no-trip cases.

Table 7.4.3-1 presents the initial conditions assumed in the analysis.

Additional conservative assumptions include:

a)

The turbine load is not reduced,'but is assumed to remain the same as prior to the CEA drop.

This results in a power mismatch between the primary and secondary

systems, which leads to a cooldown of the RCS.

b)

The most negative moderator and fuel temperature coefficients of reactivity are used because these coefficients produce the minimum PCS coolant temperature decrease upon return to 100 percent power and thus minimize DNBR.

c)

Charging pumps and pressurizer heaters are assumed to be inoperable during the transient.

This maximizes the pressure drop during the event and minimizes DNBR.

d)

All other systems are assumed to be in the manual mode of operation and have no impact on this event.

The event is initiated by dropping a full length CEA over a period of 1.0 second.

A value of 8.5 percent is used for the initial radial peaking factor increase.

The axial power shape in the hot channel remains unchanged

and, hence the increase in the 3-0 peak for the maximum power is directly proportional to the maximum increase in radial peaking factor.

Since there is no trip assumed, the peaks will stabilize at these asympototic values after a few minutes as the secondary side continues to demand 100 percent power.

7.4.3.3 Results Table 7.4.3-2 presents an illustrative sequence of events for the full-length CEA drop event initiated at the conditions described in Table 7.4.3-1.

0

A minimum CE-1 DNBR of greater than 1.24 is obtained at 900 seconds as determined from the initial radial power peaking increase following CEA drop plus 15 minutes of xenon redistribution at the final coolant conditions.

Before this time, at 10 minutes after the drop the operator will take action to reduce power in accordance with Figure 3. 1-1A of the Technical Specifications, if the misaligned CEA has not been realigned.

A maximum allowable initial linear heat generation rate of 18.0 KW/ft could exist as an initial condition without exceeding the Acceptable Fuel Centerline Melt Limit of 21.0 KW/ft during this transient.

This amount of margin is assured because the linear heat rate LCO is based on the more limiting allowable linear heat rate for LOCA (13.5 KW/ft, see Table 7.0-6).

The results for the CEA subgroup drops are bounded by the Reference Cycle.

7.4.3.4 Concl usi ons The full length CEA drop event initiated from the Technical Specification LCOs does not violate the DNBR and CTM SAFDLs.

7.4.4 CVCS Malfunction Inadvertent Boron Dilution The results are bounded by the Reference Cycle

7.4.5 Startu of an Inactive Reactor Coolant Pum Event 7.4.6 7.5 7.5.1 The results are bounded by the Reference Cycle.

Control Element Assembl Ejection The results are bounded by the Reference Cycle.

INCREASE IN REACTOR COOLANT SYSTEM INVENTORY CVCS Malfunction 7.5.2 7.6 The results are bounded by the Reference Cycle.

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

OECREASE IN PEACTOR COOLANT SYSTEM INVENTORY 7.6.1 Pressurizer Pressure Oecrease Events 7.6.2 7.6.3 The results are bounded by the Reference Cycle.

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

Steam Generator Tube Rupture The results are bounded by tie Reference Cycle.

7.7 7.7.1 MISCELLANEOUS As mmetric Steam Generator Events 7.7.1.1 The ASGT event is presented because of changes being made as part of the CPC Improvement Program (Reference 7-8), which affect CPC respons'e during the event.

The transients resulting from the malfunction of one steam generator are analyzed to determine the initial margins that must be maintained by the LCOs such that, in conjunction with the RPS (CPC high differential cold leg temperature trip), the DNBR and Fuel Centerline-to-Melt (CTM) SAFDLs are not violated.

Identification of Causes The four events which affect a single steam generator are identified below:

a)

Loss of Load to One Steam Generator (LL/1SG) b)

Excess Load to One Steam Generator (EL/1SG) c)

Loss of Feedwater to One Steam Generator (LF/1SG) d)

Excess Feedwater to One Steam Generator (EF/1SG)

Of the four events described above, it has been determined that the Loss of Load to One Steam Generator (LL/1SG) event is the limiting asymmetric event.

Hence, only.he results of this transient are reported.

The event is initiated by the inadvertent closure of both Main Steam Isolation Valves (MSIVs), which results in a loss of load to the affected steam generator.

Upon the loss of load to a single steam generator, its temperature increases, its pressure increases to the

opening pressure of the secondary safety valves, and its water level decreases.

The core inlet temperature of the loop with the affected steam generator increases resulting in a temperature tilt across the core.

In the presence of a negative moderator temperature coefficient, the radial peaking increases in the cold side of the core, resulting in a condition which potentially could cause an approach to DNBR and CTM SAFDLs.

The CPC high differential cold leg temperature trip serves as the primary means of mitigating this transient.

Additional protection is -provided by the steam generator low level trip.

7.7.1.2 Anal sis of Effects and Consequences The most negative value of the moderator temperature coefficient is assumed to maximize the calculated severity of the associated power peaking.

The LL/12SG is initiated at the initial conditions presented in Table 7.7. 1-1 and is analyzed parametrically on axial shape index to determine the maximum initial margin needed to ensure the SAFDLs are not violated.

The NSSS response is calculated using the CESEC III code.

The resulting core parameters (core flow, RCS inlet temperature, RCS

pressure, and reactor, trip time) are then input into a 2-D simulation of the core using the HERMITE code.

HERMITE is used to model both the effects of the temperature tilt on radial power distribution and the space-time impact of the scram.

The thermal margin changes are evaluated with the CETOP code.

7.7.1.3 Results A reactor trip is aenerated by the CPCs at 6.0 seconds based on hiqh differential cold leg temperature associated with the steam generators.

Table 7.7.1-2 presents the sequence of events for the loss of load to one steam generator.

Figures 7.7. 1-1 to 7.7. 1-5 show the NSSS response for core power, core heat flux, RCS temperatures, RCS

pressure, and steam generator pressure.

The minimum CE-1 DNBR calculated for the LL/1SG event is greater than 1.24.

A maximum allowable initial linear heat generation rate of 17.0 KM/ft could exist'as an initial condition without exceeding the Acceptable Fuel to Centerline Melt Limit of 21.0 KW/ft during this transient.

This amount of margin is assured because the linear heat rate LCO is based on the more limiting allowable linear heat rate for LOCA (13.5 KM/ft, see Table 7.0-6).

7.7. 1. 4 Conclusions The loss of load to one steam generator event, initiated from the Technical Specifications

LCOs, does not violate the DNBR and CTN SAFDLS.

Table 7.0-1 PVNGS Unit 2 Design Basis Events Consi ere in t e c

e a et Anal sis 7.1 Increase in Heat Removal by the Secondary System 7.1.1 7.1.2 7.1.3 7.1.4 7.1.5*

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 7.2 Decrease in Heat Removal by the Secondary System 7.2.1 7.2.2 7.2.3 7.2.4 7.2.5 7.2.6*

Loss of External Load Turbine Trip Loss of Condenser Vacuum Loss of Normal AC Power Loss of Normal Feedwater Feedwater System Pipe Breaks 7.3 Decrease in Reactor Coolant Flowrate 7.3.1 7.3.2*

Total Loss of Forced Reactor Coolant Flow Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft 7.4 Reac ivity 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

Table 7.0-1 (continued) 7.6 7.7 Decrease in Reactor Coolant System Inventory 7.6. 1 Pressurizer Pressure Decrease Events 7.6.2*

Small Primary Line Break Outside Containment 7.6.3*

Steam Generator Tube Rupture Miscellaneous 7.7. 1 Asymmetric Steam Generator Events

• Categorized as Limiting Fault Events

Table 7.0-2 DBEs Evaluated with Res ect to Offsite Dose Criterion Section 7.1.4 7.2.4 7.1.5a 7.1.5b 7.2.6 7.3.2 7.4.6 7.6.2 7.6.3 Event A)

Anticipated Operational Occurrences 1)

Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve 2)

Loss of Normal AC Power B)

Limiting Fault Events 1)

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 Shaft Seizure/Sheared Shaft 4)

Control Element Assembly Ejection 5)

Small Primary L'ine Break Outside Containment 6)

Steam Generator Tube Rupture Results Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle

Table 7.0-3 DBEs Evaluated with Res ect to RCS Pressure Criterion Section 7.2.2 7.2.3 7.2.4 7.2.5 7.4.2 7.5.2 7.2.6 7.4.6 Event A)

Anticipated Operational Occurrences 1)

Loss of External Load 2)

Turbine Trip 3)

Loss of Condenser Vacuum 4)

Loss of Normal AC Power 5)

Loss of Normal Feedwater 6)

Uncontrolled CEA Withdrawal from Subcritical or Low Power Condition 7)

Uncontrolled CEA Withdrawal at Power 8)

CYCS Malfunction 9)

Inadvertent Operation of the ECCS During Power Operation B)

Limiting Faul t Events 1)

Feedwater System Pipe Breaks 2)

Control Element Assembly Ejection Results Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Peference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle

Table 7.0-4 DBEs Evaluated with Res ect to Fuel Performance Section 7.1.1 7.1.2 7.1.3 7.3.1 7.4.1 7.4.2 7.4.3 7.6.1 7.1.5a 7.'1.5b 7.3.2 7.4.6 Event A)

Anticipated Operational Occurrences 1)

Decrease in Feedwater Temperature 2)

Increase in Feedwater flow 3)

Increased main Steam Flow 5)

Total Loss of Forced Reactor Coolant Flow 6)

Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition 7)

Uncontrolled CEA Withdrawal at Power 8)

CEA Misoperation Events 9)

Pressurizer Pressure Decrease Events 10)

Asymmetric Steam Generator Events B)

Limiting Fault Events 1)

Steam System Piping Failures:

a)

Pre-Trip Power Excursions b)

Post-Trip Return to Power 2)

Single Reactor Coolant Pump Shaft Seizur'e/Sheared Shaft 3)

Control Element Assembly Ejection Results Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Presented*

Bounded by Reference Cycle Presented Bounded by Reference Cycle Bounded by Peference Cycle Bounded by Reference Cycle

• The results of this event remain bounded by the Reference Cycle.

The event is presented due to a change in analytical methodology.

Table 7.0-5 DBEs Evaluated with Res ect to Shutdown Mar in Criterion Section 7.1.4 7.4.4 7.4.5 7.1.5b Event A)

Anticipated Operational Occurrences 1)

Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve 2)

CVCS Malfunction (Inadvertent Boron Dilution) 3)

Startup of an Inactive Reactor Coolant System Pump B)

Limiting Fault Events 1)

Steam System Pipina Failures:

a)

Post-Trip Return-to-Power Results Bounded by Reference Cycle Bounded by Reference Cycle Bounded by Reference Cycle Presented

Table 7.0-6 PVNGS Unit 2, C cle 2

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

+ Pump Heat)

Units MWt Reference Cycle Value 3893 C cle 2 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 LCO Band Assumed for All Powers psia gpm ASI Units 1785 - 2400 423,320

-0.3 to +0.3 2000 - 2325 423,320

-0.3 to +0.3 Maximum CEA Insertion at Full Power I Insertion 28 of Lead Bank

'A 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 Holdina Coils to 90%

Insertion Minimum DNBR CE-1 (SAFDL)

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

KW/ft KW/ft sec 14.0 21.0 4.0 1.25 1.30 13.5 21.0 4.0 1.24 1.30

Table 7.0-6 (continued)

Safet Parameters Moderator Temperature Coefficient Shutdown Margin (Value Assumed in Limiting Hot 2ero Power SLB)

Units 10

~sip/

F Reference Cycle Value Figure 7.0-1

-6.0 C cle 2 Value Figure 7.0-2

-6.5

Table 7.4.3-1 Ke Parameters Assumed for the Full Length CEA Dro Event Parameter Core Thermal Power Initial Core Coolant Inlet Temperature Initial Pressurizer Pressure Initial RCS Vessel Flow Rate moderator Temperature Coefficient Doppler Coefficient Multiplier Dropped CEA Morth CEA Drop Radial Peaking Distor tion Factor Time for Dropped CEA to be Fully Inserted Units Mwt 0F psia gpm 10 46p/

F sec Reference Cycle Value 3876 580 2067 423,320

-3.5 1.15

-0.06 1.07*

2.0 Cycle 2

Value 3876 570 2000 423,320

-3.5 1.15

-0.06 1.114**

1.0

~Typical asymptotic distortion factor that would not cause CPC trip when CEAC penalty applied.

• • 5'aximum distortion factor including 10 minutes of xenon redistribution for which no CEAC penalty is applied.

Table 7.4.3-2 Se uence of Events for Full Len th CEA Drop Time (sec) 0.0 1.0 1.1 20.8 34.3 900.

Event CEA Begins to Drop into Core CEA Reaches Fully Inserted Position Core Power level Reaches Minimum and Begins to increase due to Reactivity Feedbacks Minimum Pressurizer Pressure Core Power Returns to Maximum Value Minimum DNBR is Peached Set oint or Value 100'4 Inserted 92.6/ of Initial 1989 psia 1005 of Initial

>1.24 900.

Operator Action - Core Power Reduced if Dropped CEA not Realigned

~

Table 7.7.1-1 Ke Parameters Assumed for the Loss o

Load to ne team ienerator Event Parameter Total RCS Power (Core Thermal Power

+

Pump Heat)

Initial Core Inlet Temperature Initial Pressurizer Pressure Units MWt psia Cycle 2

Value 3898 565 2250 Moderator Temperature Coefficient Doppler Coef icient Multiplier Rafial Distortion Factor for 18 F Core Inlet Temperature Asymmetry 10 6p/

F

-3.5 0.85

1. 13

Table 7.7.1-2 Se uence of Events for the Loss of Loa to ne team enerator vent

~Time (sec 0.0 0.1 4.9 Event Initiate Closure of a Single Main Steam Isolation Valve (MSIV)

MSIV on Affected Steam Generator is Fully Closed Safety Valves Open on Isolated Steam Generator Setpoint or Value 1277 psia 6.0 6.75 7.09 7.7 8.7 CPC Delta-T Setpoint Reached (Differential Cold Leg Temperature Analysis Setpoint)

Trip Breakers Open CEAs Begin to Drop Minimum CE-1 DNBR Maximum Steam Generator Pressure 18 F

a1.24 1279 psia

o I-z O

~Q>>

~Q Lu Ql M ct:4 D gQ uj K0 tL Q

t0.22

-1.0

-2.0

-3.0 t0.22x 10 hp/ F ALLOWABLE MTC C

H TAVG 5960F O.yhp/

F (594 F, -3.0x10 hp/

F) 4.0 480 sooo sso'VERAGE MODERATOR TEMPERATURE, F

6oo'alo Verde Nuclaar Cenarating Station CYCLE 1

ALLOWABLE HTC NODES 1

AND 2 Figure 7.0-1

~

U O

I CD 0

-0. 5 lil i5 CD I

LJJ Cl CD

-1.0

-1.5

-2.0

-2.5

-3.0 Al LOHABLE

-3. 5 20 llP 60 80 100 CORE PONER LEVELS X OF RATED THERNAL POWER Palo Vercle Nuclear Generating Station CYCLE 2 ALLOllhBLE HTC MODES 1

AND 2 Figure 7.0-2

C)

I CC 8

~CYCLE 2 6

(D 2

C) 200 300 000 500 EQO 700 NODERATOR TBlPERATURE, F

Palo Verde.'nuclear Cenerabng

~~choo STEAM LENE BREAK MODERATOR COOLDOWN REACTIV STY t s(S ERT a0.

VS MODERATOR TEMPERATURE Figure T. l. 5-1

0 0

120 100

'0 60 40 20 12 20 TIMEi SECONDS Palo Verde.'nuclear Genetating Station ASYMMETRIC STc~

GENERATOR EVENTS CORE POWER VS TINE Figure 7.7.1-1

120 100 80 CD 60 00 I

UJ UJ CC CD 20 12 20 TINEi SECONDS Palo Verde Nuclear Generating Station ASYMMETRlC STEAM GENERATOR EVENTS CORE HEAT FLUX VS TIME Figure 7.7.1-2

700 650 O

600 i5 550 l

CD CD 5

500 CORE OUTLET CORE AVERAGE CORE INLET 450 400 12 20 TINEi SECONDS Pa)o Verde Nuclear Cenaating Stati~

ASYMMETRIC STEAM GENERATOR EVENTS REACTOR COOLANT TEMPERATURES VS TIME Figure 7.7.1-3

2300 2200 2100 2000 1900 1800 1700 12 20 TINEA'ECONDS Palo Verde Nuclear GenerahAg Stabcxl ASYMMETRIC STEAM GENERATOR EVENTS REACTOR COOLANT SYSTBI PRESSURE VS TIME Figure 7.7.1-4

1300 1200 AFFECTED SG 1100 1000 900 UNAFFECTED SG 800 700 12 16 20 TIMEi SECONDS Palo Verde Nuclear Cenetating Station ASYMMETRIC STEAM GENERATOR EVENTS STEAM GENERATOR PRESSURES VS TIME Figure 7.7. 1-5

8.0 ECCS ANALYSIS 8.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT

8. 1. 1 Introduction And Summar An ECCS performance analysis of the limiting break size was performed for PVNGS-2 Cycle 2 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 with nuclear flux augmentation factors of unity.

This PLHGR is a reduction of 0.5 kw/ft from the Cycle 1 limit for PVNGS-2.

The method of analysis and detailed results which support this value are presented herein.

8.1. 2 Method Of Anal s i s The ECCS performance analysis for PVNGS-2 Cycle 2 consisted of an evaluation of the differences between Cycle 2 and PVNGS-2 Cycle 1, hereafter referred to as the reference cycle.

Acceptable ECCS performance was demonstrated for the reference cycle in Reference 8-2 and approved by the NPC 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 calculations, and steam cooling heat transfer coefficients of the reference cycle apply to PVNGS-2 Cycle 2 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 2.

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

0

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) 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 1000NMD/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 2.

8. 1. 3 Results Significant core and system parameters for the reference cycle and PVNGS-2 Cycle 2 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 limiting case results in a peak clad temperature of 1960'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 rat'e to 6

151. lx10 lbm/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.

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

8.1.4 Conclusion The ECCS performance evaluation for PVNGS-2 Cycle 2 results in a peak clad temperature of 1960'F, a peak local clad oxidation percentage of 5:7% and a peak core wide clad oxidation percentage of less than 0.80% compared to the acceptance criteria of 2200 F, 17%

and l%%d, respectively.

Therefore, operation of PVNGS-2 Cycle 2 at a core power level of 3876 MWt (102%%d of 3800 MWt) and a

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

8.2 SMALL BREAK LOSS"OF-COOLANT ACCIDENT A review of Cycle 2 fuel and core data confirmed that the reported smal'I break loss-of-coolant accident results (Reference 8-9) for PVNGS-2 Cycle 1 bounds PVNGS-2 Cycle 2.

These results have been approved by the NRC in Reference 8-10.

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% 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.

TABLE 8.1-1 PVNGS-2 Cycle 2

Core and S stem Parameters Parameter Units Reactor Power 9

102%%u of Nominal(NWt)

Average Linear Heat Rate 8 102K 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)

(1)

Fuel Centerline Temperature

( 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)

Refer ence

~Cc1 e 3876 5.64

14. 0 565.0 164x10 159xl0 1514 3424.7 2175.0 1129.0 774 Function of Elevation 13.5 3.0(10

)

1.52 4214

~Cele 2

3876 5.71 13.5 565.0 164x10 159x10 1520 3315.9 2113.0 1127.0 1000 400 1.0 13.5 3.0(10

)

1.52 3744 Initial values at the limiting hot rod burnup as calculated by STRIKIN-II (1) at the peak linear heat generation rate.

~

(

TABLE 8. 1-2 PVNGS-2 Cycle 2

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

(%%d)

Total Core-Mide Clad Oxidation

(%%d)

~Cele 1

14.0 2091 278.5 88.2 9.0

< 0.80

~Cele 2

13.5 1960 269.6 96.1 5.7

< 0.80

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 Density (LPD) trips to (1) ensure that the specified acceptable fuel design limits on departure from nucleate boiling and centerline fuel meltino 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 sub-systems, components and parameters are maintained within operating limits and Limiting Conditions for Operation (LCOs).

9.2 CPCS SOFTWARE !MODIFICATIONS The CPC/CEAC software for PVNGS-2 is being modified for operation in Cycle 2.

This modification is being made to implement the CPC Improvement Program (CIP) including algorithm and plant-specific data base

changes, changes to the list of addressable constants and implementation of Reload Data Block (RDB).

The CPC/CEAC algorithms for PVNGS-2 Cycle 2 are the same as those implemented at PVNGS-I Cycle 2, and described in References 9-1 and 9-2.

The revised list of addressable constants are defined in Reference 9-3.

The Reload Data Block (Reference 9-5) is a group of constants that are located in protected memory of the CPC and the CEAC, separate from other non-addressable constants.

The RDB

constants are loaded from a separate RDB disk and can be changed without requiring a

CPC/CEAC software change.

The RDB has previously been implemented at PVNGS-1 for Cycle 2.

The modifications for PVNGS-2 Cycle 2 relative to the Cycle 1

software are identical to those made for PVNGS-1 Cycle 2 and are described in References 9-3, 9-4, 9-5 and 9-9.

The modifications described in References 9-3, 9-4, and 9-9 are incorporated in References 9-1 and 9-2.

The implementation of all changes will be done in accordance with the established software change procedures, References 9-6 and 9-7.

Cycle dependent values of the data base and RDB constants will be determined for PVNGS-2 Cycle 2 consistent with the cycle design, performance and safety analyses.

The RDB constants will be installed on the Cycle 2

RDB Disk for loading at the site as described in Reference 9-8.

9.3 ADDRESSABLE CONSTANTS Certain CPC constants are addressable so that they can be changed as required during operation.

Addressable constants include

( 1) constants that are measured during startup (e.g.,

shape annealing matrix, boundary point power correlation coefficients, and adjustments for CEA shadowing and 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 PVNGS-2 Cycle 2 consistent with the software and methodology established in the CIP (Reference 9-3, 9-4 and 9-5).

Uncertainty factors will be determined using a modified statistical combination of uncertainties method (Reference 9-10).

This method has been approved by the NRC in Reference 9-12.

This method has also previously been implemented at PVNGS-1 for Cycle 2.

0 0

9.4 DIGITAL MONITORING SYSTEM (COLSS)

The Core Operating Limit Supervisory System (COLSS), described in Reference 9-11, 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 Cycle 2

core design.

10.0 Technical S ecifications This section provides a

summary of recommended changes that should be made to the PVNGS-2 Technical Specifications in order to update the Technical Specifications for Cycle 2 operation.

A description of each change and the corresponding technical specification section are presented in the following pages.

ATTACHMENT NO.

1 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.1.

1.2 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment changes the Shutdown Margin versus Cold Leg Temperature curve as set forth in Technical Specification (T.S.) 3.1.1.2.

The change is to the Hot Zero Power endpoint.

The change is from 6.0X ~ P to 6.5X < p.

JUSTIFICATION FOR PROPOSED AMENDMENT Due to the design of Cycle 2,

the Cycle 2

moderator cooldown reactivity insertion curve is more adverse than the Cycle 1 curve.

Because of the more adverse cooldown reactivity insertion curve for Cycle 2,

the Shutdown Margin is required to be increased from 6X h p to 6.5%

h P at zero power.

The increase in margin is required to maintain the operation of Cycle 2 within the safety analysis.

10-2

ATTACHMENT NO.

2 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAI SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3 ~ 1 ~ 1 ~ 3 DESCRIPTION OF PROPOSED AMENDMENT The proposed amendment changes the Moderator Temperature Coefficient (MTC)

Figure 3.3-1 as

set, forth in Technical Specification (T.S.)

3.1.1.3.

The changes are two fold.

The operating bounds of the MTC are being broadened to accommodate the op'eration of Cycle 2 and the x axis is being changed to core power level instead of average moderator temperature.

JUSTIFICATION FOR PROPOSED AMENDMENT In preparation for future 18 months cycles, the Cycle 2 core physics is such that a change in the MTC operating band will occur.

To accommodate operation throughout Cycle 2, the MTC operating band has become more positive because of the increase in fuel enrichment which requires higher boron concentration at beginning of the cycle.

As operation into the cycle

proceeds, the MTC will become more negative.

In addition, the x axis is to be changed to core power level instead of average moderator temperature.

By changing the x axis to core power level, the method of calculating the bounding MTC for the most limiting case becomes simplified.

Making the MTC a dependent variable of core power only and not of inlet temperature and core

power, as the present curve represents, the calculation of the limiting MTC need only be performed once.

The present method of manipulating MTC requires performing the analyses several

times, at various average moderator temperatures, to be sure of obtaining the most limiting case

but, with the new

method, MTC can be calculated once and there is assurance that the most limiting case value is obtained.

Both graphs are the results of the same set of codes, only the method of manipulating the data is slightly different.

10-3

ATTACHMENT NO.

3 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

2.8 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment changes the operational pressure band of the pressurizer, as set forth in Technical Specification (T.S.) 3.2.8 to a tighter operational band.

The band is being changed from 1815 psia through 2370 psia to 2025 psia through 2300 psia.

JUSTIFICATION FOR PROPOSED AMENDMENT To support the Core Protection Calculator (CPC)

Improvement

Program, the operational pressure band of the pressurizer requires tightening.

Potential transients initiated at the extremes of the Cycle 1 pressure range were not analyzed for Cycle 2.

Because the calculations were not performed, the CPCs cannot support normal operation outside of the proposed pressurizer pressure band.

10-4

I Ol 0

ATTACHMENT NO.

4 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.1.3.1 and 3.1.

3.2 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment modifies the CEA Position Technical Specifications (T.S.)

3.1.3.1 and 3.1.3.2 by removing direct references of the control of insertion of the Part-length Control Element Assemblies (PLCEA) and creates an additional T.S.

that addresses the length of time for insertion and the insertion limit of the PLCEA specifically.

JUSTIFICATION FOR PROPOSED AMENDMENT Creating a separate T.S. for addressing operation of the PLCEA would provide an improvement to the potential consequences of a PLCEA drop or slip initiated from an allowable inserted position.

It would also add a

more explicit Limiting Condition for Operation to clarify the allowable duration for the PLCEA to remain within the defined ranges of axial position.

10-5

e

ATTACHMENT NO.

5 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

3.1 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment changes the response time of the DNBR-Low Reactor Coolant Pump (RCP) shaft speed trip in Technical Specification (T.S.) 3.3.1, Table 3.3-2.

The change is due to redefining the events which take place before the Control Element Assemblies drop into the core.

During Cycle 1, the response time of

.75 seconds was measured from the time a trip condition

existed, such as a loss of power to the RCP motors, to the moment the Control Element Drive Mechanisms (CEDM) coil breakers opened.

During Cycle 2

operation, the response time of

.3 seconds will be defined from the time a

signal is sent down the RCP shaft speed sensor line to the CPCs to the moment the CEDM coil breakers open.

JUSTIFICATION FOR PROPOSED AMENDMENT During the Cycle 1 startup testing, it was found that the projected Reactor Coolant flow rate trip software housed in the Core Protection Calculators, which monitors the RCP shaft speed and projects what the Reactor Coolant System flow will be in the future, was too sensitive to small deviations in RCP shaft speeds and caused unnecessary trips to the Unit.

To correct this

problem, the software dealing with the projected flow rate was taken out.

In its place, trip software which trips the unit when the RCP shaft speed slows to 95X of its normal

speed, as did the projected flow rate

trip, was installed.

Because of this change, the response time as defined for the RCP shaft speed trip has been redefined for Cycle 2 to reflect the purpose of the new trip.

As a result of the redefinition of the response

time, the safety analysis for Cycle 2 has taken credit for the faster time and to ensure that the Unit is operated within the safety

analysis, Table 3.3-2 will have to reflect the credited response time as it was used in the safety analysis.

10-6

0 0

ATTACHMENT NO.

6 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.1.

3.6 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment revises the CEA Insertion Limits as set forth in Technical Specification (T.S.) 3.1.3.6.

Operation of the regulating Control Element Assemblies (CEAs) during Cycle 2 will be more limited than in Cycle 1.

The revisions to the curves wi11 maintain the margin of safety and ensure that there will be sufficient shutdown margin to handle the most limiting Anticipated Operational Occurrence (AOO) and limiting fault events.

JUSTIFICATION FOR PROPOSED AMENDMENT The proposed changes mad to the CEA Insertion Limits are due to the change in the Cycle 2 core physics.

Because of the change to the core, the worth of the CEAs has changed and as a result, the effects of the dropped and ejected CEA events change.

To ensure that there is sufficient margin to mitigate such

events, CEA insertion has to be restricted by the insertion limits set forth in the proposed T.S. 3.1.3.6.

10-7

~

ATTACHMENT NO.

7 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO. BE AMENDED 3.3.1 Table 3.3-2a DESCRIPTION OF PROPOSED AMENDMENT The existing PVNGS Unit 1 Technical Specifications provide an allowance for entering penalty factors into the Core Protection Calculators (CPCs) to compensate for Resistance Temperature Detector (RTD) response times greater than 8 seconds (but less than or equal to 13 seconds).

These CPC penalty factors are provided in Technical Specification Table 3.3-2a and are supported by the Cycle 1 safety analyses.

However, the Cycle 2 safety analyses will not support these CPC penalty factors.

Therefore, Table 3.3-2a must be deleted and Table 3.3-2 must be, revised to remove this CPC penalty factor allowance.

JUSTIFICATION FOR PROPOSED AMENDMENT This Technical Specification change is necessary in order to ensure that the Cycle 2 safety analyses assumptions are complied with during Unit 1, Cycle 2

operations.

The Cycle 2 safety analyses assume a maximum RTD response time of 8

seconds and do not include an allowance to enter CPC penalty factors to compensate for RTD response times greater than 8

seconds.

Therefore, there should not be any allowances in the Technical Specifications for using the CPC penalty factors.

For this reason, Technical Specification Table 3.3-2a should be deleted and Table 3.3-2 should be revised to remove the penalty factor allowances.

10-8

0

ATTACHMENT NO.

8 PROPOSED AMENDMENT TO PVNGS UNIT TWO - CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 2.1.1.1 and Basis DESCRIPTION OF PROPOSED AMENDMENT The proposed amendment changes references to the calculated Departure from Nucleate Boiling Ratio (DNBR) from 1.231 to 1.24 as set forth in Technical Specification (T.S;) 2.1.1.1, Table 2.2-1, Bases 2.1.1, and Bases 2.2.1.

The amendment also deletes references to the calculation of additional rod bow penalties if the rod bow penalty incorporated into the DNBR limit is not sufficient for any part'f the cycle.

The low pressurizer pressure floor is also changed from 1861 to 1860 because of the changed DNBR value.

JUSTIFICATION FOR PROPOSED AMENDMENT During Cycle 1 operation, the rod bow penalty factox was applied to the DNBR in increments.

This method provided a

means for not penalizing the operational margin unnecessarily during the cycle.

As the fuel assemblies approach higher burnup, the advantage of the Cycle 1 method no longer exists.

The application of a rod bow penalty factor large enough to provide protection throughout the cycle is now more advantageous.

This can be accomplished because the physics of the Cycle 2 core is such that, by applying a rod bow penalty factor of 1.75X Minimum DNBR (MDNBR) to the DNBR limit, there will be sufficient margin to compensate for the effects of rod bow caused by those bundles less than 30,000 MWD/MTU.

For those bundles with burnups of greater than 30 GWD/MTU, there is sufficient margin from other factors to offset the small increase in the rod bow penalty the greater than 30 GWD/MTU bundles would incur.

As a result of the DNBR change, a xeevaluation of the safety analysis was performed to determine if the low pressurizer pressure floor for the DNBR-low trip would change.

The low DNBR trip provides protection in the event of an increase in heat removal by the secondary system and subsequent cooldown of the reactor coolant.

The analysis has shown that a pressuxizer pressure of 1860 instead of 1861 will ensure that, if a reactor trip occurs on Low-DNBR, the plant will not reach the Specified Acceptable Fuel Design Limits (SAFDLs).

10-9

~

ATTACHMENT NO.

9 PROPOSED AMENDMENT TO PVNGS UNIT TWO - CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

2.5 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment changes the Reactor Coolant System (RCS) total flow rate as set forth in Technical Specification (T.S.) 3.2.5 from greater than or equal to 164.0 x 106 ibm/hr to greater than or equal to 155.8 x 10 ibm/hr.

6 JUSTIFICATION FOR PROPOSED AMENDMENT T.S.

3.2.5 is being changed to eliminate an ambiguity in where instrument uncertainty is to be included when comparing measured RCS flow rate against the RCS flow rate used in the safety analysis.

As currently worded, actual total RCS flow rate is to be compared against the 100X design flow value of 164.0 x

106 ibm/hr.

The term "actual" implies that the RCS flow rate determined by the Reactor Coolant Pump (RCP) delta-pressure method is to be corrected for pressure transmitter uncertainty.

The uncertainty amounts to a

maximum of 4% of flow for transmitters within their calibration period.

The corrected flow rate is then compared to 164.0 x 106 ibm/hr.

The RCS flow rate used in the safety analysis, however, is 95X of the design flow of 155.8 x 106 ibm/hr.

The 100X design flow rate of 164.0 x 106 ibm/hr conservatively accommodated the maximum instrument uncertainty of 4X, removing the need to correct for instrument uncertainty.

The T.S.

bases states that the specification is provided to ensure that the actual total RCS flow rate is maintained at or above the minimum value used in the safety analysis.

This T.S.

change will remove the ambiguity and permit any changes in instrument uncertainty to be handled procedurally rather than requiring additional T.S.

changes.

10-10

ATTACHMENT NO.

10 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

2.1 DESCRIPTION

OF PROPOSED AMENDMENT The 'proposed amendment changes the Linear Heat Rate (LHR) limit as defined in Technical Specification (T.S.)

3.2.1 from 14.0 kw/ft to 13.5 kw/ft.

The change also provides information for the appropriate methods of monitoring LHR and formats the T.S. with regard to human factors.

JUSTIFICATION FOR PROPOSED AMENDMENT In support of the Unit 1 reload, the reanalysis of the Safety Analysis resulted in a change in the Linear Heat Rate limit to ensure the peak fuel clad temperature is not exceeded.

The change in the LHR is partly due to the change in the method of performing the safety analysis.

As part of the

analysis, penalties are applied to compensate for increased power peaking due to small inter-pellet gaps caused by the densification of small inter-pellet gaps.

These penalties are called Augmentation Factors and were not used for the

. Cycle 2 analysis.

This method change has been approved by the NRC in "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

104 to Facility Operating License No.

DPR-53, Baltimore Gas and Electric

Company, Calvert Cliffs Nuclear Power Plant Unit No.

1, Docket No.

50-317".

Other factors contributing to the change in LHR are from increased fuel enrichment and the core loading pattern.

In addition to changing t'e references to LHR, the amendment also delineates how LHR is to be monitored.

.By providing more detail of the monitoring of LHR, assurance is provided that the LHR will be maintained below the specified limit.

The amendment also changes the format of the ACTION statement in such way as to facilitate assessment of the actions required if the limit should be exceeded.

10-11

ATTACHMENT NO. 11 PROPOSED AMENDMENT TO PVNGS UNIT TWO - CYCLE TWO

'TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.2.4, 3.3.1 and Bases DESCRIPTION OF PROPOSED AMENDMENT The proposed amendment will revise Technical Specifications 3.2.4, 3.3.1, Bases 3.1.3.1/3.1.3.2 and Bases 3.2.4.

The changes are as follows:

T.S.

3.2.4-(1)

Replaces the T.S.

with a

new format which addresses the specific conditions for monitoring DNBR with or without COLSS and/or the

CEACs, (2) delineates by a

new format what ACTIONS should be

taken, (3) xemoves reference to the DNBR Penalty Factor table used 'in T.S.

4.2.4.4 and (4) replaces the present graph figures 3.2-1 and 3.2-2 of the DNBR limits with graph figures 3.2-1, 3.2-2 and 3.2-2A addressing DNBR operating limits for the conditions mentioned in (1) above.

T.S. 3.3.1-(l)

Removes references to the operation of the reactor with both CEACs inoperable and with or without COLSS inservice and (2) deletes the

graph, of DNBR margin operating limit, figure 3.3-1, based on COLSS for both CEACs inoperable.

These changes are result of being incorporated into the proposed T.S. 3.2.4.

Bases 3.1.3.1/3.1.3.2-(l)

Removes references to Cycle 1 specific information and (2) modifies Bases due.to T.S. 3.2.4 changes.

Bases 3.2.4-Modifies Bases due to the T.S. 3.2.4 changes.

These changes are due in part to ensuring operation of Cycle 2 within the approved safety analysis and to improving the Technical Specifications from a human factors point of view.

JUSTIFICATION FOR PROPOSED AMENDMENT The proposed changes are due to (1) ensuring operation of the reactor within appxoved safety analysis for Cycle 2

by modifying the T.S.

graphs, (2) increasing operator reliability by placing DNBR operating limits in one place, and (3) eliminating superfluous information to reduce confusion and the possibility of misuse (i.e., eliminating the Table in T.S. 4.2.4.4).

10-12

0

ATTACHMENT NO.

12 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

2.3 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment change expands the operating limits of Azimuthal Tilt with COLSS in service as specified in Technical Specification 3.2.3.

The azimuthal tilt limits will be a step function of power with the upper limit of 0.20 at 20%

power and stepping down to 0.10 at 40%

power, where it remains steady state through to 100% power.

JUSTIFICATION FOR PROPOSED AMENDMENT During a reactor power cutback event in Unit 1 the plant was unable to go above 20%

power because the Azimuthal Tilt Limit would have been exceeded.

They were required to remain below 20%

power for approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> until xenon burned out.

This delay could have been prevented and the azimuthal tilt corrected if the plant had been allowed to increase power.

This would cause the xenon to burn out faster thus restoring the plant within the limits sooner.

By implementing the proposed change such delays could be avoided.

10-13

ATTACHMENT NO.

13 PROPOSED AMENDMENT TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

3.2 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment ensures the trip value of the Refueling Water Storage Tank for recirculation is maintained at the midpoint of the allowable operational values as set forth in Technical Specification (T.S.) 3.3.2 Table 3.3-4.

JUSTIFICATION FOR PROPOSED AMENDMENT The proposed change to T.S.

3.3.2 Table 3.3-4 will ensure optimal protection of the Refueling Water Storage Tank pumps by maintaining adequate margin for the trip value within the allowable operational values.

10-14

ATTACHMENT NO.

14 PROPOSED AMENDME%T TO PVNGS UNIT TWO CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED Various Bases DESCRIPTION OF PROPOSED AMENDMENT The proposed amendment is a

number of administrative changes for the following Technical Specifications (T.S.):

Bases 3/4.3.1 and 3/4.3.2 1)

Page 3-2 Remove Cycle 1 cycle specific information no longer needed for Cycle 2.

Bases 2.2.1 1)

Page 2-2 Remove reference to CESSAR for description of the method of calculation for the trip variables for DNBR-Low and Local Power Density High trips and replace with the correct CE Topicals.

2)

Page 2-3 Update the latest revision used for calculating the PVNGS trip setpoint values.

JUSTIFICATION FOR PROPOSED AMENDMENT The administrative changes are required to ensure clarity and conciseness.

The change to Bases 3/4.3.1 removes information which pertained to Cycle 1 and is no longer valid for Cycle 2.

The change to Bases 2.2.1 changes the source of the description of the method of calculation for the trip variables for DNBR-Low and Local Power Density High trips from the CESSAR to the correct CE Topicals and updates the T.S. to the latest revision of CEN 286 (V), Rev.

2.

10-15

0

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

The described test 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 Mater Reactors" and supplements normal surveillance tests which are required by Technical Specifications (i.e.,

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

11. 1.

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

By the first method, all CEA groups would be fully withdrawn with the exception of the lead regulating group which would be positioned at approximately mid-core.

The boron concentration of the reactor coolant would then be reduced until criticality is attained.

By the second

method, the shutdown CEA groups would be adjusted to the expected critical concentration, and 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% ak/k of the predicted values.

0

11. 1. 3 Tem er atur e React ivit Coeffici ent The isothermal temperature coefficient (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

-4 values will be verified to be within +0.3 x 10 ak/k 'F of 'the predicted values.

11.1.4 CEA Reactivit 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 wor ths.

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 medium worth) may be used.

The groups to be measured will be exchanged with the appropriate reference group.

Acceptance criteria wiT1 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 IBN value will be verified to be within +15 ppm/%%d hk/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

processed plant data against which they are normally calibrated.

If necessary, the power ascension wi'll be suspended while necessary data r eduction and equipment calibrations are performed.

The following measurements will be performed during the program.

11.2. 1 Flux S mmetr Verfication 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.

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 within the design limits while reactor power is increased to 100%%u, where the second comparisons 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%.

B.

The RNS of the difference between the measured and predicted core aver age 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 SAN and Boundar Point Power Correlation Coefficients BPPCC Verification The SAM and BPPCC values will be determined from a linear regression analysis of the measure 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 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 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 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 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 margin penalties are incorporated into the CPC and COLSS uncertainty constants.

11.2. 5 Reacti vit Coe ffici ents at Power The isothermal temperature coefficient (ITC) and the power coefficient (PC) 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 PC will be measured by changing

power, compensating with CEA motion, and maintaining coolant temperature steady.

The ITC and PC will be verified to be within +0.3 x 10 Ak/k'F and

+ 0.3 x 10 ak/k%

power, respectively, of predicted vales.

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 with +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 The results of all tests will be reviewed by the plant's reactor engineering group.

If the acceptance criteria of the star tup 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.

0

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.

(1-2)

"Combustion Engineering Standard Safety Analysis Report CESSAR", Docket PSTN-50-470F.

J.

G.

Haynes to G.

M. Knighton, "Technical Specification Amendment - Sections 1.0, 2.2, 3/4. 1, 3/4.3, 3/4.10" ANPP-39798, January 23, 1987.

( 1-4)

J.

G.

Haynes to USNRC "Submittal of the Reload Analysis Report for Unit 1 Cycle 2", 161-00321-JGH/LJH, June 29, 1987.

(1-5)

Submittal of changes to CPC Loss of Flow Constant.

Letter 161-00362-JGH/JRP, July 13, 1987.

SECTION

2.0 REFERENCES

None SECTION

3.0 REFERENCES

None SECTION

4.0 REFERENCES

"Palo Verde Unit 1 Fuel Design Report" NPSD-207-P, November 1982.

e'

(4 2)

V-CE-33635, "Palo Verde Nuclear Generating Station Fuel Handling Interference", April 10, 1986.

Letter, A. E. Lundvall, Jr. to J.

R. Miller (Chief Operati'ng Reactors Branch

$3), "Calvert Cliffs Nuclear Power Plant Unit Nos.

1 and 2, Docket Nos.

50-317 and 50-318,, Request for Amendment",

December 31, 1984.

EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding Volume 5:

Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PMR Fuel Rods," April, 1985.

(4 5)

"Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

104 to Facility Operating License No. DPR-53, Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant Unit No.

1 Docket No. 50-317", May, 1985.

(4-6)

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

(4-7)

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

July, 1981.

(4-8)

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

(BG8E), "Safety Evaluation of CEN-161 (FATES3)," March 31, 1983.

(4-9)

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

(4-10)

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

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

(4-11)

CESSAR

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

0

(4-12)

J.

G. Haynes (ANPP) to Document Control Desk (NRC), "Fuel Assembly Guide Tube Wear Program for PVNGS Unit 2,"

161-00453-JGH/SGB, August 20, 1987.

SECTION

5.0 REFERENCES

EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding, Volume 5:

Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PWR Fuel Rods," April, 1985.

(5-2)

A. E. Lundvall (BG8E) to J.

R. Miller (NRC), "Calvert Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317 Eighth cycle License Application," February 22, 1985.

(5-3)

"Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

104 to Facility Operating License No. DPR-53, Baltimore Gas and Electric Company, Calvert Cliffs Nuclear Power Plant Unit No.

1, Docket No. 50-317," May, 1985.

(5-4)

CENPD-153-P, Rev.

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

(5-5)

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

e

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)-,

Rev.

1-P, "CETOP Code Structure and Modeling Methods for San ONofre Nuclear Generating Station Units 2 and 3",- September, 1981.

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

July, 1987.

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.

CESSAR SSER 2 Section 4.4.6, Statistical Combination of Uncertainties.

(6 7)

CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983.

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 Engineering Standard Safety Analysis Report," Docket No. 50-470.

"Standard Review Plant,"

NUREG-0800, Rev. 2, 1981.

(7-4)

"CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System,"

December

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

1982.

I Ik 0

(7-5)

R.

V. Macbeth, "An Appraisal of Forced Convection Burnout Data," Proc. Instr.

Mech. Engrs.,

Vol. 180, Pt.

3C, PP 37-50, 1965"1966.

(7-6)

D. H. Lee, "An Experimental Investigation of Forced Convection Burnout in High Pressure Water - Part IV, Large Diameter Tubes at about 1600 psia," A.E.E.W. Report

R479, 1986.

(7-7)

CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs 1 and 2," December 1981.

(7-10)

CENPD-161-P, "TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core," July 1975.

(7-11)

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

(7-12)

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

'7-13)

CENPD-199-P-A, Rev.

1-P, "CE Setpoint Methodology,"

January, 1986

'ECTION

8.0 REFERENCES

-ECCS ANALYSIS 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)

Letter from C.

Ferguson (C-E) to E.

E.

Van Burnt, Jr.

(ANPP), V-CE-32964, "License Condition 21-Large Break LOCA," October 1,

1985.

0' 0

(NRC approved of the PVNGS-2 Cycle 1 Performance Results of (8-2).

To be supplied by Project/PVNGS).

(8-4)

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

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

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.

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

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.

Letter from C. Ferguson (C-E) to E.

E.

Van Brunt, Jr.

(ANPP), V-CE-32895, "Large and Small Break LOCA Re-Analysis for License Condition 21," September 13, 1985 (Assumes this is references by PVNGS for PV2 Cycle 2).

r 0

(NRC approval of License Condition 21 ECCS performance results of (8-9).

To be supplied by Project/PVNGS.

Letter from C. Ferguson (C-E) to P.

F. Crawley (ANPP),

V-CE-34932, "Tech.

Spec.

LPSI Flow Limit," July 7, 1987.

12.9 SECTION

9.0 REFERENCES

(9-1)

CEN-304-P, Rev. 01-P, "Functional Design Requirements 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-308-P-A, "CPC/CEAC Software Modifications for the CPC Improvement Program," April, 1986.

(9-4)

CEN-310-P-A, "CPC and Methodology Changes for the CPC Improvement Program," April, 1986.

(9-5)

CEN-330-PA, Rev.

OO-P, "CPC/CEAC Software Modifications for the CPC Improvement Program Reload Data Block,"* 1986.

CEN-39(A)-P, Rev. 3-P-A, "CPC Protection Algorithm Software Change Procedure,"

November 1986.

(9-7)

CEN-39(A)-P, Supplement 1-P, Rev. 3-P-A, "CPC Protection Algorithm Software Change Procedure Supplement 1,"

November, 1986.

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

(9-9)

CEN-281(S)-P, Rev. 01-P, "CPC/CEAC Software Modifications for San Onofre Nuclear Steam Generating Station Units No.

2 and 3," November, 1984.

"will be used prior to final RAR transmittal.

r 0

(9-10)

CEN-356(V)-P, Rev. 01-P, "Modified Statistical Combination of Uncertainties," July, 1987.

(9-11)

CEN-312-P, Rev. 01-P, "Overview Description of the Core Operating Limit Supervisory System (COLSS)",

November, 1986.

(9-12)

Letter from E. A. Licitra (NRC) to E.

E.

Van Brunt, Jr.

(ANPP) "Issuance of Amendment No.

24 io Faii'lity Operating License No.

NPF-41 for the Palo Verde Nuclear Generating Station Unit 2," October 21, 1987.

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 Pressurized Water Reactors".

(11-2)

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

November 1985.