ML17300A956

Reload Analysis Rept for Palo Verde Nuclear Generating Station Unit 1 Cycle 2.

ML17300A956
Person / Time
Site:	Palo Verde
Issue date:	06/29/1987
From:	ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
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ML17300A955	List: ML17300A954 ML17300A956
References
TAC-65691, NUDOCS 8707060343
Download: ML17300A956 (224)

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

l. INTRODUCTIOM AND

SUMMARY

2. OPERATING HISTORY OF THE REFERENCE CYCLE 2-1

3. GENERAL DESCRIPTIOM 3-1

4. FUEL SYSTEM DESIGN 4-1

5. NUCLEAR DESIGN 5-1

6. THERMAL-HYDRAULIC DESIGN 6-1

7. TRANSIENT ANALYSIS 7-1

8. ECCS ANALYSIS 8"1

9. REACTOR PROTECTION AND MONITORING SYSTEYi 9" 1

10. TECHNICAL SPECIFICATIOMS 10-1

11. STARTUP TESTING

12. REFERENCES 12-1

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INTRODUCTION AND

SUMMARY

This report provides an evaluation of the design and performance of Palo Verde Nuclear Generating Station Unit 1 (PVNGS-1) during its second cycle of operation at 100~ rated core power of 3800 MWt and NSSS power of 3817 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 D assemblies. The Cycle 1 termination burnup has been assumed to be between 16,512 and 17,280 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 References 1-} and 1-2.

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. The 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.

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'odifications 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 summ'arized in Section 9. The concurrent data base changes are a result of plant-specific application of CIP to PVNGS-1 Cycle 2.

E 2.0 OPERATING HISTORY OF TH" REFERENCE CYCLE The plant is currently in its first fuel cycle which began with initial criticality on May 25, 1985. Power Ascension began on June 1, 1985 and on January 28, 1986 the unit was declared in commercial operation.

It is presently estimated that Cycle 1 will terminate on or about September 12, 1987. The Cycle 1 termination point can vary between 16,512 MWD/T and 17,280 MWD/T to accommodate the plant schedule and still be within the assumptions of the Cycle 2 analyses.

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 eleven Batch B wi 11 be removed from the Cycle 1 core and be replaced with 80 fresh, Batch D assemblies. Ninety-seven Batch B and all Batch C assemblies now in the core will be retained.

The reload batch wi 11 consist of 36 type DO assemblies, 28 type D*

assemblies with 8 burnable poison shims per assembly, 12 type DX assemblies with 8 burnable poison shims per assembly and 4 type D/

assemblies with 8 burnable poison shims per assembly. These sub-batch types are zone-enriched and their configurations are shown as well as those of the Batch B and Batch C assemblies in Figures 3-1 and 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 17,280 MMD/T.

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

TABLE 3-1 PALO VERDE NUCLEAR GENERATING STATION UNIT 1 Cycle 2 Core Loading Initial Total Number Assembly Fuel Rods Initial Number Shim of Desig- Number of per Enrichment Shims/ Loading Fuel Shim nation Assemblies Assembly (w/o U-235} Assembly (gm B10/in) Rods Rods B-Lo 33 208 2.78 16 .01842 6864 528 12 1.92 396 B-Hi 64 208 2.78 16 .02532 13312 1024 12 1.92 768 C 40 224 3.30 8960 0

12 2.78 480 208 3.30 16 .01151 4992 384 12 2.78 288 36 184 4.05 6624 52 3.36 1872 p* 28 176 3.36 .008 4928 224 52 2.78 1456 DX 12 216 3.36 .008 2592 ,

96 12 2.78 144 216 3.36 .020 864 32 12 2.78 48 241 54588 2288

FUEL No. OF FUEL No, OF SHIM ASSEMBLY NUMBER OF ENRICHMENT RODS PER RODS / gm B10/IN.

TYPE ASSEMBLIES W/T % U235 ASSEMBLY ASSEMBLY 69 1.92 236 1.92 12 16 0.01842 BLO 2.78 208 64 1.92 12 BHI 16 0.02532 2.78 208 Cp 40 2.78 12 3.30 224 24 2,78 12 16 0.01151 CLO 3.30 208 BLp BHI AND CLp Cp WATERHOLE 8 LOWER ENRICHED FUEL PIN Q HIGHER ENRICHED FUEL PIN g . SHIM P,IN RIZONA Figure lo Verde FIRST CYCLE ASSEMBLY FUEL LOADINGS Nuclear Generating WATERHOLE AND SHIM PLACEMENT 3-1 Station

SUB-BATCH D - 36 ASSEMBLIES 0 4.05 w/o U-235 IHI 3.36 w/o U-235 SUB-BATCH D" - 28 ASSEMBLIES 0 3.36 w/o U-235 S 2.78 w/o U-235

~ B4C - AL203 SHIM PIN .008 gm B.10/IN SUB-BATCH DX - 12 ASSEMBLIES 0 3.36 w/o U-235 8 2.78 w/o U-235 t B4C - AL203 SHIM PIN .008 gm B-10/IN SUB-BATCH D/ - 4 ASSEMBLI ES 0 3,36 w/o U-235

.8 2,78.w/o U.235 0 B4C - AL203 SHIM PIN .020 gm B-10/IN IZONA Figure lo Verde SECOND CYCLE ASSEMBLY FUEL LOADINGS Nuclear Generating WATERHOLE AND SHIM PLACEMENT 3.2 I

Station

B.Hi 8-Lo D D 8-Hi D D . CLo C-00 8-Hi 8-Hi D C-Lo 8-Hi po C-Lo p%

8-Hi D C-00 8-Hi D" 8-Hi C-00 8-Hi D C-Lo 8-Hi D/ 8-Lo D 8-Lo C-00 8-Hi D 8-Hi p% 8-Lo C40 C-00 8-Hi DX 8-Lo C-Lo pt 8-Hi D" C-00 8-Lo DX 8-Hi D C-00 C-Lo C-00 8-Lo 8-Hi DX 8-Lo C-00 3 D 8-Hi p+ 8-Hi C DX 8-Hi C-00 8-Lo II ~

iZoaa Figure lo Verde 3.3 CYCLE 2 FUEL MANAGEMENT uciear Generating Station I e

B.HI 2 B.LO .

1 4 p I 95'og ler eo 5, BHI D 7 D 8 CLO 9 C.OO 10 B-HI lqeoq IMHso lo3u( (Socol 11 B HI 12 D 13 C.L 14 B.HI 15 D 16 C.I.O 17 'D l78( 3 l9(o( 2 18 B-HI 19 D 20 C40 21 BHI 22 D 23 B.HI 24 C.OO 25 B.HI I'IC o9 I 3323 (75 oc (97(o lo353 l 1GQo 26 D 27 C.LO 28 B.HII 29 30. B-LO 31 Do 32 B-LO 33 C-00 0 l7S(Z l 75oo /$ 2.((o (9c 8'c(

34 B.HI 35 D 36 B.HI 37 D 38 B-LO 39 C-00 40 C00 41 B-HI 42 D

((ro8 Il( (2. (82(r I (735 i@or 9 l 9v9~

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(ZONA Figure lo Verde BOC ASSEMBLY AVERAGE BURNUP (MWD/T) 3-4 Nuclear Generating Station

5 LEAD REGULATING BANK 4 SECOND REGULATING BANK 3 THIRD REGULATING BANK 2 FOURTH REGULATING BANK LAST REGULATING BANK 8 SHUTDOWN BANK B A -SHUTDOWN BANK A P2 PLR GROUP 2 Pl PL ROUP 1 S SPACE CEA LOCATIONS 2

S 3 10 12 13 14 15 16 17 '1 8

1 A 20 22 23 24 . 25 26 27 29 31 P2 P2 32 33 34 5 '36 38 40 41 45 46 B

47 50 51 53 55 56 57 58 59 60 61 P1 P1 2 62 63 65 66 67 68 69 70 71 72 73 74 75 76 77 78 A B 4 'A A a 79 80 82 83 84 86 87 89 90 91 3 94 S P2 P2 96 97 99 100 101 102 103 104 105 106 107 108 109 1 10 '1 12 113 114 1'l5 1'l6 117 118 119 120 121 '1 22 123 124 125 126 127 128 129 3 P1 130 131 133 134 135 136 137 '138 139 140 141 142 143 145 146 A 1 II 147 148 149 150 151 152 153 154 '155 156 157 158 159 160 161 162 163

,S P2 S P2 164 165 166 167 168 169 170 171 172 173 74 175 176 177 178 179 180 A

181 182 183 '184 185 186 187 188 189 190 '191 192 193 194 195 P1 5 P1 196 197 198 199 200 201 202 03 204 205 207 208 209 210 211 212 213 214 215 2'l6 217 218 219 220 221 222 223 P2 P2 ~ 2 224 225 226 227 228 229 230 231 233 234 A (

A 235 236 237 238 239 240 3 S IZOIIIA Figure o Verde CEA BANK IDENTIFICATION 3.5A Nuc ear Generating Station

i P R P R P R P R P R P R P R P R P R P R P R P R IZONA Figure lo Verde 3.5B CEA PATTERN Nuclear Gelierating

'tation

BOX INSTR.

8 10 tl 12 13 14 15 16 17 18 2 3 4 20 21 22. 24 25 26 27 28 29 31 5 6 7 8 9 32 33 35 36 37 39 40 41 42 43 46 10 11 12 47 48 49 50 51 62 63 SS SS 58 69 60 61 13 14 15. 16 62 63 65 66 67 68 70 71 72 73 74 76 77 78 17 18 19 80 81 82 83 84 85 86 87 88 89 91 95 20 21 22 23 24 25 96 97 88 99 100 101 102 103 104 105 106 107 108 109 110 112 26 27 28 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 129 29 30 31 32 33 34 130 131 132 133 135 136 139 140 141 142 144 145 146 35 36 37 147 '148 149 150 151 152 153 154 155 . 156 157 159 160 161 162 163 39 41 166 167 169 170 171 172 173 174 175 176 177 178 179 180 42 43 44 45 46 181 182 183 184 185 186 187 188 189 190 191 192 193 194 47 48 49 50 196 197 198 199 200 201 202 203 204 205 206 207 -

208 209 210 51 52 53 211 212 213 214 215 216 21? 218 219 220 221 222 223 54 55 56 57 22S 226 227 228 229 230 231 234 58 59 60 235 236 238 239 240 241 61 RIZONA Figure Io Verde 3-6 INSTR UMENT LOCATIONS Nuclear Generating Station

4.0 FUEL SYSTEM DESIGN MECHANICAL DESIGN 4.1.1 Fuel Desian Mith the exception of the 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 1. No changes in mechanical design 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 quality of the fuel and to improve the burnup capability of the poison rods. The specific changes 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 worst case tolerance conditions, Reference 4-2. It should be noted that all of the fuel bundles utilizing the previous design have been handled at PVNGS without revealing any interference problems.

2. The inspection envelope for the fuel bundle assembly has been changed from a square of 8.230 inches per side for all of the bundle with exception of the vicinity around the upper most grid where a square of 8.250 inches per side is permissible to a square of 8.290 inches per side for the entire length of the fuel bundle. This change will not affect any of the mechanical design considerations from the standpoint of structural

'U behavior and integrity, fuel assembly bow, handling, interface or other considerations. Furthermore, a 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 B-1Q loadings while reducing end of life internal pressure.

A draft copy of an EPR1-sponsored report dealing with the phenomena of interpellet gap formation and clad collapse in modern PMR 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 thi's synopsis was tha.

clad collapse analyses are not necessary for modern C-E manufactured fuel because of the absence of gaps between pellets. The 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 no interpellet gap formation. Since the provisions of 'emonstrated the NRC's concurrence have been satisfied, no cycle specific clad collapse analysis was performed for Cycle 2.

4.2 GUIDE TUBE WEAR Twenty of the fuel assemblies that had CEA's located in them during Cycle I will be inspected for guide tube wear. This inspection is part of the required licensing procedures required by the NRC for

'll plants after the first cycle of operation (References 4-10 and 4-Ij.).

4.3 THERMAL DESIGN The thermal performance 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 rang'e analyzed is in excess of that expected at the end of Cycle 2.

Results of these burnup dependent fuel performance calculations were used in the Transient Analysis presented in Section 7 and in the ECCS Analysis presented in Section 8.

4 CHEMICAL DESIGN

, The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch D fuel are identical to those of the fuel batches included in Cycle I. Thus, the chemical or metallurgical performance of the Batch D fuel will remain unchanged from the performance of the Cycle 1 fuel (Reference 4-7).

4,5 SHOULDER GAP ADE UACY Ten fuel bundle assemblies will be inspected visually and ten fuel bundles will, in addition, be measured for shoulder gap and guide tube length. These inspections are part of the licensing procedures required by the NRC for all plants after the first cycle of operation (References 4-12 and 4-13). Conservative estimates show that the fuel is acceptable for Cycle 2 operation. Additionally, oxide measurements on the clad will be taken on four fuel assemblies.

i 5.0 NUCLEAR DESIGN PHYSICS CHARACTERISTICS 5.1.1 Fuel Mana ement 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 17,280 MMD/T, the Cycle 2 reactivity lifetime for full power operation is expected to be 13,056 MMD/T. Explicit evaluations have been performed to ass'ure applicability of all analyses to a Cycle 1 termination burnup of between 16,512 and 19,085 MWD/T and for a Cycle 2 length up to 13,098 MUD/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-l.

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.

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 condition's 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 range from 1.18 at BOC2 to 1.08 at EOC2.

5.2 SAFETY RELATED DATA 5,2.1 Auomentation 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 of any reactor loaded exclusively with this type of fuel. This discussion of the elimination of the augmentation factors 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, Analytical Input to In-Core Measurements In-core detector measurement cons ants to be used in evalua.ing the reload cycle power distributions wi 11 be calculated in accordance with Reference 5-4. ROCS-DIT with the MC module will 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 Cycl e 1.

i 5.3.3 Nuclear Design Methodology 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 methodology

'was described in Reference 5-5.

TABLE 5-1 PVNGS-) CYCLE 2 NOMINAL PHYSICS CHARACTERISTICS Reference Dissolved Boron Uni ts Cvcl e Cvcle 2 Dissolved Boron Concentration for Criticality, CEAs hiithdrawn, Hot Full Power PPM 657 1088 Equi librium Xenon, BOC Boron Worth Ho. =Full Power, BOC PPM/% 91 112 Hot Full Power, EOC PPM/K 83 90 Moderator Temperature Coefficients Hot Full Power, Equilibrium Xenon Beginning of Cycle 10 4 BP/OF -).0 -0.5 End of Cycle 10-4 ~p/o F -2.5 -2.4 Doppler Coefficient Hot Zero Power, BOC 10-5 ~p/OF -1 ~ 5 Hot Full Power, BOC 10-5 ~)/OF "1.3 Hot Full Power, EOC 10-5 hp/ r "1.5 Total Dela ed Neutron Fraction. ~ ef BOC 0.0073 0.0060 EOC 0.0053 0.005)

Prompt Neutron Generation Time. 1*

BOC 10-6 sec 28.1 24.0 EOC 10-6 sec 31.3 29.9

TABLE 5-2 PVNGS-1 CYCLE 2 LIMI'TING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT FULL POWER STEAM LINE BREAK,  % ~P,END-QF"CYCLE (EOC)

Reference Cvel e Cvcle 2 Wor th of al l CEAs Inserted -15.0 -16.0 Stuck CEA Allowance +4.9 +4.5 Worth of all CEAs Less Highest Worth CEA Stuck Out -10.1 -11.5 Full Power Dependent Insertion Li'mit CEA Bite +0.2 +0.2 Calculated Scram Worth -11.3 Physics Uncertainty +1 0 + e2 Other Allowances (losses due to voiding) +0.4 +0.1 Net Available Scram Worth -8.2 -10.0

i TABLE 5-3 PVNGS-1 CYCLE 2 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, X Beoinninc of C cle End of C cl e Regulating Refer ence Reference CEAs Cycl e Cycle 2 Cycle Cycle 2 Group 5 -.32 -.24 ~ 32 .27 Group 4 -,47 -.37 -.42 Group 3 -.79 -.90 -.99 -.92 Note:

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1. NRC imposed'0.01 DNBR penalty for HID-I 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-I 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'he reference cycle ~ i 6.2 EFFECTS OF FUEL ROD BOWING OM 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, l.75% NDNBR, is valid for bundle burnups up to 30 GWD/NTU. This penalty is included in the 1.24 DNBR limi't; 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. TABLE 6-1 PVNGS-1 Cycle 2 Thermal Hydraulic Parameters at Full Power Reference General Characteristics Units Cvcle Cvcle 2 Total Heat Output (Core only) Mh 3800 3800 10 Btu/hr 12,970 12,970 Fraction of Heat Generated in 0.975 0.975 Fuel Rod Primary System Pressure psia 2250 2250 Nominal Inlet Temperature (Nominal) 565.0 565.0 Total Reactor Coolant Flow 445,600 423,300 gpss (Minimum Steady State) 10 lb/hr 164.0 155.8 Coolant Flow Through Core (Minimum) 10 lb/hr 159.0 151.1 Hydraulic Diameter (Nominal Channel) ft 0.039 0.039 Average Mass Velocity 10 1 b/hr- f . 2.61 2.49 Pressure Drop Across Core (Minimum ps1 15.9 14.5 steady state flow irreversible ~ P over enti re fuel assembly) Total Pressure Drop Across Vessel psl 56.7 51 ' (Based on nominal dimensions and minimum steady state flow) Core Average Heat Flux (Accounts BTU/hr-ft 184,400" 185,700~ for fraction of heat generated in fuel rod and axial densifica-tion factor) Total Heat Transfer Area (Accounts 68,600* 68,100** for axial densification factor) Film Coefficient at Average BTU/hr-ft OF 6300 6100 Conditions Average Film Temperature Difference 'F 30 Average Linear Heat Rate of Unden- kw/ft 5.4 5.4 sified Fuel Rod (Accounts for fraction of heat generated in fuel rod) TABLE 6-1 (continued) Reference General Characteristics Units ~Cele 2 Average Core Enthalpy. Rise BTU/lb 81.6 - 85.9 Maximum Clad Surface Temperature oF 656 - 656 . Engineering Heat Flux Factor ] 03*** 1.03+ Engineering Factor on Hot Channel 02'k** 1.03+ Heat Input Rod Pitch, Bowing and Clad Diameter 05*%* 1.05+ "Factor Fuel Densification Factor (Axial) 1.002 1.002 NOTES:

• Based on 1920 Poison Rods.

• • Based on 2288 poison rods..

• "* These factors were combined statistically with other uncertainty factors to define a new design limit on CE-1 minimum DNBR at the 95/955 confidence/probability level when iterating on power as discussed in Refer ence 6-5.

+ These factors" have been combined statistically with other uncertain.y factors as described in References 6-4 and 6-5 to define the DNBR limit and overall uncertainty penalty factors to be applied in the DNBR calculations in COLSS and CPC. These overall uncertainty penalty factors, when used in conjunction with the Cycle 2 DNBR limit provide assurance a. the 95/95 confidence/probability level that the hot rod will no experience DNB. ++ Tech. Spec. minimum flow rate. Al .1/2..2 7.0 Non-LOCA Safet Anal sis 7.0. 1 Introduction This section presents the results. of the Palo Verde Nuclear Generating Station Unit 1 (PVNGS-1), Cycle 2 Non-LOCA safety analyses at 3800 MMt. The Design Basis Events (DBEs) considered in the safety analy'ses are listed in Table 7.0-1. These events are categorized into three groups: Moderate 'requency, Infrequent, and Limiting Fault events. For the purpose of this report, the Moderate Frequency and Infrequent Events will be termed Antici-pated Oper'ational 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-1 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-1 docket, the CESSAR docket, or on other .dockets is used. i 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 (Reference .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'he RCP Shaft Seizure and Sheared Shaft event. The TORC code is described in References 7-10 and 7-11. The number of fuel pins predicted to experience clad fai lure 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 for which- the statis.ical 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. The HERMITE computer code (Reference 7-9) was used to simulate the r'eactor 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-.ime 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 PVNGS-3,. 7.0.4 Input Parameters and Anal sis Assumptions Table 7.0-6 summarizes the core parameters assumed 'in the Cycle 2 transient analysis and compares them to the values used in the Refer ence Cycle. Specific initial conditions for each event are tabulated in the section of the i report summarizing that event. 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'hose, 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 Specifications. 7.0.5 Conclusion All DBEs are evaluated for PVNGS-1, Cycle 2 to determine whether their results 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. Table 7.0-1 PVNGS Unit 1 Design'Basis Events Considered in the Cycle 2 Safety Analysis 7.1 Increase in Heat Removal by the Secondary System 7,.1.1 Decrease in Feedwater Temperature 7.1.2 Increase in Feedwaier Flow 7.1.3 Increased Main Steam Flow 7.1.4 Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve 7.1.5*, Steam System Piping Failures 7.2 Decrease in Heat Removal by. the Secondary System 7.2.1 Loss of External Load 7.2.2 Turbine Trip 7.2.3 Loss of Condenser Vacuum 7.2.4 Loss of Normal AC Power 7.2.5 Loss of Normal Feedwater 7 '.6* Feedwater System Pipe Breaks 7.3 Decrease in Reactor Coolant Flowrate 7.3. 1 Total Loss of Forced Reactor Coolant Flow 7.3.-2~ Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft 7.4 Reactivity and Power Distribution Anomalies 7.4.1 Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition 7.4.2 Uncontrolled CEA Withdrawal at Power 7.4.3 CEA Misoperation Events 7.4.4 CVCS Malfunction ( Inadvertent Boron .Dilution) 7 '.5 'tartup of an Inactive Reactor 'Coolant System Pump 7.4.6* Control Element Asse'mbly Ejection I 7.5 Increase in Reactor Coolant System Inventory 7.5.1 CVCS Malfunction 7.5 ' Inadvertent Operation of the ECCS During Power Operation

• Categori.zed as Limiting Fault Events,

Table 7.0-1 (continued) 0 Decrease 7.F 1 in Reactor Coolant System Inventory Pressurizer Pressure Decrease Events 7.6.2* Small Primary Line Break Outside Containment 7.6.3" Steam Generator Tube Rupture 7.7 Miscellaneous 7.7. 1 Asymmetric Steam Generator Events Categorized as Limiting Fau1t Events Table 7.0-2 DBEs Evaluated with Respect tc Offsite Dose Criterion Section Event Resul ts A) Anticipated Operational Occurrences 7.1.4 1) 1nadvertent Opening of a Steam Bounded by Generator Safety Valve or Reference Cycle Atmospheric Dump Valve .7.2.4 2) Loss of Normal AC Power Bounded by Reference Cycle B) Limiting Fault Events

1) Steam System Piping Failures: Bounded by Reference Cycle 7.1.5a a) Pre-Trip Power Excursions 7.1.5b b) Post-Trip Return-to-Power 7.2.6 2) Feedwater System Pipe Breaks Bounded by Reference Cycle 7.3.2 3) Single Reactor Coolant Pump Bounded by Shaft Seizure/Sheared Shaft Reference Cycle 7,4.6

~ ~ 4) Control Element Assembly Ejection Bounded by Reference Cycle 7.6.2 ') Small Primary Line Break Outside Containment Bounded by Reference Cycle 7.6.3 6) Steam Generator Tube Rupture Bounded by Reference Cycle ~ e Wli. ~ "/0 Table 7.0-3 DBEs Evaluated with Respect to RCS Pressure Criterion Section Event Results A) Anticipated Operational Occurrences 7.2.1 1) Loss of External Load Bounded by Reference Cycle 7.2.2 2) Turbine Trip Bounded by Reference Cycle 7.2.3 3) Loss of Condenser Vacuum Bounded by Reference Cycle 7.2.4 4) Loss of Normal AC Power Bounded by Reference -Cycle 7.2.5 5) Loss of Normal Feedwater Bounded by Reference Cycle

6) Uncontrolled CEA h~ithdrawal from Bounded by Subcritical or Low Power Condition Reference Cycle 7.4.2 7) Uncontrolled CEA Mithdrawal at Power Bounded by.

Reference Cycle 7.5.1 8) CVCS Malfunction Bounded by Reference Cycle 7.5.2 9) Inadvertent Operation of the Bounded by ECCS Ouring Power Operation Reference Cycle B) Limiting Fault Events 7 '.6 1) Feedwater System Pipe Breaks Bounded by Reference Cycle 7.4.6 2) Control Element Assembly Ejection Bounded by Reference Cycle Table 7.0-4 DBEs Evaluated with Respect to Fuel Performance Settion Even'. Results A) Anticipated Operational Occurrences 7.1.1 1) Decrease in Feedwater Temperature Bounded by Reference Cycle 7.1.2 2) Increase in Feedwater Flow Bounded by Reference Cycle 7.1.3 3) Increased Main Steam Flow Bounded by Reference Cycle 7.3.1 5) Total Loss of Forced Reactor Presented Coolant Flow 7.4.1 6) Uncontrolled CEA Withdrawal from a Bounded by Subcritical or Low Power Condition Reference Cycle 7.4.2 7) Uncontrolled CEA Withdrawal Bounded by at Power Reference Cycle 7.4.3 r 8) CEA Misoperation Events Presented* 7.6.1 9) Pressurizer Pressure Decrease Bounded by Events Reference Cycle 7.7.1 10) Asymmetric Steam Generator Events Presented B) Limiting Fault Events

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

. 7.3.2 2) Single Reactor Coolant Pump Bounded by Shaf t Sei zure/Sheared Shaf t Reference Cycle 7.4.6 3) Control Element Assembly Ejection 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. Il Table 7.0"5 DB"-s "valua ed' th Respec. o Shutdown Narcin Crite.ion Even= Resul ts A) Anticipated Operational Occurrences

1) Inadverten. Opening o a Steam Bounded by Genera-or Safety Valve or Reference Cvcle Atmospheric Dump Valve

2) CVCS Malfunction (Inadver en- Bounded by Bor on Dilution) Reference Cycle

3) Startup oi an Inactive Reactor Bounded by Coolant System Pump Refer ence Cycj,e B) Limiting Faul. vents

) Steam System Pipinc Failures: Presentec a) Post-Trip Returr.-to-Power Table 7.0-6 PVNGS Unit 1 Cycle 2 Core Parameters Input to Safety Analyses Reference Cycle Safety Parameters Units Value Cvcl e 2 Value Total RCS Power MWt 3893 3898 (Core Thermal Power + Pump Heat) Core Inlet Steady State 'F 560 to 570 560 to 570 Temperature (90% power'and (90% power and above) above) 550 to 572 550 to 572 (below 90% power) (below 90% power) Steady State psia 1785 - 2400 2000 - 2325 RCS Pressure Min'imum Guaranteed gpm 423,320 423,320 Delivered Volumetric Flow Rate Axial Shape Index LCO ASI -0 ' to +0.3 -0.3 to +0 3 ~ Band Assumed for Units All Powers Maximum CEA Insertion  % Insertion 28 28 at Full Power of Lead Bank / Insertion 25 25 of Part-Length Maximum Initial Linear KW/f t 14.0 13.5 Heat Rate Steady State Linear KW/f t 21.0 21.0 Heat Rate for Fuel Center Line Melt CEA Drop Time from sec 4.0 4.0 Removal of Power to Holding Coils to 90% Insertion Minimum DNBR CE-1 (SAFDL) 1.25 l. 24 Macbeth (Fuel failure 1.30 1.30 limit for post-trip SLB with LOAC-References 7-5 and 7-6) AR.1/i2 Table 7.0-6 (continued) Reference Cycle Safet Parameters Units Value Cvcle 2 Value Noderator'emperature 10 ~p /oF Figure 7.0-1 Figure 7.0-2 Coefficient Shutdown Nargin (Value %hp -6.0 "6s5 Assumed in Limiting Hot 2ero Power SLB) 7.) Increase in Heat Removal bv the Secondar System Decrease in Feedwater Temperature 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 Ooenino of a Steam Generator Safety Valve or Atmospheric Dump Valve The results are bounded by the Reference Cycle. 7.1.5 ~ 1 Steam S stem Pi ino Failures 7.1,5a Steam S stem Pi ina Failures: Inside and Outside Containmen. Pr e-Trio 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 System Piping Failures: Post-Trio 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 Cycle 1 curve. Figure 7. 1.5-1 compares the two curves. In addition, a sweep-out volume of 119 ft3 before Safety Injection reaches the RCS was assumed for Cycle 2, which is more 'onservative than the 34.7 ft3 a'ssumed for the Reference Cycle. The effect of the more adverse reactivity insertion was accommodated for Cycle 2 by increasing the Shutdown Margin required by the Technical Speci ications at zero power from 6/~p to 6.5% hp . E 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 bv the Secondary S stem 7.2.1 Loss of External Load I The results are bounded by the Reference Cycle. 7.2.2 Turbine Trio 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. 7.2.6 Feedwater S stem Pipe Breaks The results are bounded by the Reference Cycle. Decrease in Reactor Coolant Flowrate Loss of Forced Reactor Coolant The Loss of Coolant Flow '(LOF) Event is analyzed to determine the minimum initial margin that must be maintained by the Limiting Conditions for Operation (LCOs) such that, in conjunction with the Reactor Protection System (RPS), the DNBR SAFDL will not be exceeded. This event is presented in order to show the effect of two changes relative to the reference analysis (Reference 7-2). First, a change is being made to the CPC constants to include a trip on low reactor coolant pump (RCP) speed which replaced the previous trip on low flow-projected DNBR. The NRC accepted the pump speed trip in the CPCs for SONGS Units 2 and 3 (Reference 7-14). Second, the current analysis used a more rapid coastdown than the reference analysis. 7.3. 1. 1 Identification of Causes A Loss of Coolant Flow may result from a loss of electrical power to one or more of the four reactor coolant pumps. Reactor trip on Loss of Coolant Flow is initiated by the CPCs on low RCP speed. For a Loss of Flow at any power operating condition, a trip will be initiated when any RCP shaft speed drops to 95.0 percent of nominal speed. Because partial Loss of Flow events are less limiting, only the 4-pump total Loss of Flow even. is reanalyzed. 7.3. 1.2 Analysis of Effects and Consequences The 4-pump Loss of Flow transient is characterized by the flow coastdown curve given in Figure 7.3. 1-1. This flow coastdown bounds the coastdowns observed during startup testing at PVNGS-1 and 2. The 1-D HERMITE space-time code was employed to determine the reactor core response. The core inlet coolant 'temperature and the primary system pressure were held constant at their . initial values to evaluate the short term DNBR transient as explained in CESSAR Appendix 15A. The present analysis considered several cases over the parameter space given in Table 7.0-.6. The cases chosen for this analysis were determined to be limiting based on parametric studies performed in Cycle l. All. 1/16 7.3. 1. 3 Results The single representative case presented here used .initial conditions shown in Table 7.3. 1-1 and an axial shape index of zero. This set of initial conditions is one of a very large number of combinations within the parameter space (Table 7.0-6), which would provide the minimum initial margin required by the COLSS power operating limit. The consequences following a total Loss of Flow initiated from any one of these combinations of conditions would be no more adverse than those presented here. Table 7.3. 1-2 presents the sequence of events. The CPC low RCP speed trip setpoint is reached at 0.61 seconds. The CEAs start to drop into the core at 1.25 seconds. A minimum CE-1 DNBR above 1.24 is reached at 2.2 seconds. Figures 7.3. 1-2 and 7.3. 1-3 present the core power and the heat flux as a function of time. The DNBR transient is shown in Figure 7.3. 1-4. 7 3 ~ ~ 1.4 Conclusions ~ e 4-pump Loss of Flow event initiated from the Technical Specification LCOs in conjunction with the CPC trip on low RCP shaft speed does, not exceed the DNBR SAFDL. The initial margin required as a result of this analysis is preserved by the DNBR Margin LCO's via an adjustment to COLSS constants. ~ AW 1/17 Table 7.3.1-1 Ke Parameters Assumed for the Total Loss of Forced Reactor Coolant Flow Event Reference Cycle Cycl e 2 Parameter Units Value Value Core Thermal Power MWt 3876 3876 Initial Core Coolant oF '565 565 Inlet Temperature Initial Core Flow Rate 10 ibm/hr 157.4 159.1 Initial Reactor Coolant psia 2250 2250 System Pressure Moderator Temperature Coefficient 10 P~p/oF 0.0 0.0 Doppler Coefficient Multiplier 0.85 0.85 Radial Power Peaking Factor 1.62 1. 70. RPS Trip Setpoint Projected 95.0/ of nominal DNBR limit shaft speed Scram Worth at Trip "10.0 -10.0 Table 7.3.1-2 Sequence of Events for Total Loss of Forced Reactor Coolant Flow Even Time sec Event Setooint or Value 0.0 Loss of Power to all Four Reactor Coolant Pumps Low Reactor Coolant Pump Shaft Speed 95.05 of nominal Trip Condition shaft speed Trip Breakers Open 1.25 CEAs Begin to Drop 2.20 Minimum CE-1.DMBR Sinole Reactor Coolant Pump Shaft Seizure/Sheared Shaft The results are bounded by the Reference Cycle. Reactivit and Power Distribution Anomalies 7.4.) Uncontrolled.- CEA Withdrawal from a Subcritical or Low Power Condition The results are bounded by the Reference Cycle. 7.4.2 Uncontrolled CEA Withdrawal at Power The results are bounded by the Reference Cycle. 7.4.3 CEA Misooeration Even. The single full-length and part-length CEA drop events are analyzed .to determine the initial thermal margins tha. must be maintained by the Limiting, Conditions for Operation (LCOs) such that the DNBR and Fuel Centerline-to-Yiel (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 equal to one (no penalty) in the Control Element Assembly Calculators (CEACs), as described in Refer ence .7-8. This . applies to both full-length and part-length CEA downward deviations. Sufficient thermal margin wi 11 be maintained by the LCOs to compensate for the removal of CEA position-related penalty factors for downward CEA deviations of l 4-fingered CEAs. The CEA position-related penalty factors for downward deviations of l2-fingered CEAs have been calculated such that the CPCs provide 4 a trip when necessary. A part-length PDIL 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 th e Reference Cycle to provide a trip when necessary. 7.4.3. 1 Identification o Causes The CEA Misoperation 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 Conseouences The single full length CEA drop is analyze'd P because this event requires, the maximum initial margin t'o 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 tne 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 RCS coolant tempera.ure decrease upon return to 100 percen 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 even and minimizes DNBR. d) All other systems are assumed to be in the manual mode of operation and have no impact on thi's 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, the increase in the 3-D 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 asymptotic 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. A minimum CE-1 DNBR of'reater than 1.24 is obtained at 900 seconds as determined from the initi'al radial power peaking increase following GEA drop 'plus 15 minutes of xenon redistribution at the final coolant conditions. At this time 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 Yiel t Limit of 21.0 KW/ft dur ing this transient. 0 ~ 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 Conc)usions The full length CEA drop event initiated from the Technical Specification LCOs does not violate the DNBR and CTM SAFDLs. AW. /20 Table 7.4.3-1 0 Kev Parameters Assumed for the Full Length CEA Dro Event Reference Cycle .Cycle 2 Parameter Units Value Value Core Thermal Power MWt 3876 3876 Initial Core Coolant oF 580 570 Inlet Temperature Initial Pressurizer psia 2067 2000 Pressure Initial RCS Vessel gpm 423,320 423,320 Flow Rate Moderator Temperature 10 Dp/ F "3.o "3.5 Coefficient Doppler Coefficient 1.15 1.15 Multiplier Dropped CEA t Worth ehp "0.06 "0.06 CEA Drop Radial 1.07* Peaking Distorti'on Factor Time for Dropped CEA to be sec 2.0 1.0 Fully Inserted

• Typical asymptotic distortion factor that would not cause CPC trip when CEAC penalty applied'*Maximum distortion factor including 15 minutes of xenon redistribution for which no CEAC penalty is applied.

Table 7.4.3-2 Seauence of Events f'r Full Lenath CEA Dro Time sec Event Setpoint or Value 0.0 1.0 1.1 CEA CEA Begins to Drop Reaches Core Power Level Reaches Begins to Increase due to Reactivity Feedbacks '00% into Core Fully Inserted Position Minimum and 92.6% Inserted of Initial 20.8 Minimum Pr essurizer Pressure 1989 psia 34.3 Core Power Returns to Maximum Value 100% of Initial 900. Minimum DNBR is Reached > 1.24 900. Operator Action - Core Power Reduced if Dropped CEA not Realigned 7.A.4 CVCS Malfunction Inadvertent Boron Dilution The results are bounded by" the Reference Cycle. 7.4.5 Star uo of a'n inac:ive Reactor Coolan: ovmo Even . The results are bounded by the 'Reference Cycle. 7.t,.6 Control Elemen. Assembly Eiec.ion The results are bounded by the Reference Cycle. 7.5 increase in Reac:or Coolant System inven-orv 7 o ~ CVCS Mal unc.i on The results are bounded by I he Reference Cyc1e, 7,5,2 ~ ~ inadverten. Operation o the ECCS Durinc Power Operation The results are bounded by the Reference Cycle. 7.6 De rease in Rea ol Coolant Svs:em inventory ".6.I Pressurize. Pressure Decrease Events The results are bounded by the Reference Cycle. 7.6.2 Small Prima". v Line Pipe Break Outside Con ainment The results are bounaed by the Reference Cycle. 7.6.3 Steam Generator Tube Rupture The results are bounded by the Reference Cycle. AM.1/29.1 7.7 Miscellaneous 7.7. 1 As mmetric Steam Generator Events The ASGT event is presented because of changes being made as part of the CCIP, which affect CPC response 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 Center line-to-Melt (CTM) SAFDLs are not violated. 7.7. 1. 1 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 the 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, arid 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 tempera.ure 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 Analysis of Effects and Conseouences The most negative value of the moderator temperature coefficient is assumed to maximize the calculated severity of the associated power peaking. The LI /1SG 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 11I 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 HERNiTE code. HERYiiTE is used to model both the effects of the temperature ti lt 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 genera.ed by the CPCs at 6.0 seconds based on high 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 -Kh'/ft could exist as an initial condition without exceeding the Acceptable Fuel to Centerline Melt Limit of 21.0 KM/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 DN8R and CTM SAFDLs. Table 7.7.1-1 Kev Parameters Assumed for the Loss of Load to One Steam Generator Event Cycle 2 Parameter Uni ta Value Total RCS Power 3898 (Core Thermal Power + Pump Heat) Initial Core Inlet Temperature 565 Initial Pressurizer psl a 2250 Pressure Moderator Temperature Coefficient 1O ~p/'F -3.5 Doppler Coefficient Multiplier 0.85 Radial Distortion Factor for 18'F Core Inlet Temperature Asymmetry . Table 7. 7. 1-2 Se uence of Events for the Loss of. Load to.One Steam Generator Event Time sec Event Setooint or Value Q.Q Initiate Closure of a Single Main Steam Isolation Valve (MSIV) 0.1 MSIV on Affected Steam Generator is Fully Closed 0.1 ,Steam Flow from Unaffected Steam Generator Increases to Maintain Turbine Power Safety Valves Open on Isolated Steam 12?7 psia Generator CPC Delta-T Setpoint Reached 18 F (Differential Cold Leg Temperature Analysis Setpoint) 6.75 Trip Breakers Open 7,09 CEAs Begin to Drop 7.7 Minimum CE-1 DNBR 21.24 8.7 Maximum Steam Generator Pressure 1279 psia t J 1 LL C) +0. 5 +0.'22X10 5 P/ F (596 Fi 0.0 5P/ F) +0.22 .I 0 CD OC I LLI -1. 0 RLL'OMRBLE I'1TC LL LL LIJ LU . CD CQ -2.0 -3. 0 C+ II CD I TRVG LLI CD (596 F, -3.5x10 hP/'F) -4,0 480 500 550 COO RVERRGE f10DI:RRTOIl TI'.ViPEIlRTURE, F Palo Verde Nuclear Generating Station CYCLE 1 ALI 014ABLE NTC MODES 1 AND 2 FigIII e 7.0-1 0 -0. 5 -1.0 ALLOWABLE. -1,5 -2.0 -2.5 -3.0 -3.5 0 20 40 60 80 100 CORE POWER LEVEL % OF RATED TJIERflAL POWER Palo Verde Nuclear Generating Station CYCLE 2 ALLOWABLE tlTC HOOES 1 Attt) 2 f ~CYCLE 2 CYCLE j. 200 500 F00 500 E'.00 700 NGD"RATOR TENPERATURE, F Palo Verde Nuclear Generating Station STEAYi LINE BREAK MODERATOR COOLDOWN REACTIYITY It<SERT:O.'i YS MODERATOR TEMPERATURE

"igure 7.1.5-1

1,0 0.9 0,8 0,7 o 0 6 0.5 0,4 0,5 0,2 0,1 , 0 0 2,0 4,0 6,0 8,0 10,0 TINE> SECONDS Palo Verde Fiuclear Generating Station TOTAL LOSS OF REACTOR COOLANT FLOW CORE FLOW FRACTION .VS Tlt1E Fioure 7.3.1-1' 110 100 90 80 70 C) 60 50 40 30 C) 20, 10 0 0 2,0 4,0 .6. 0 8.0 10,0 TIflEi SECONDS Palo Verde nuclear Generating Station rgih".. \ TOTAL LOSS OF REACTOR COOLANT FLOW CORE POWER VS TINE Figure 7.3. 1-2 Cl 110 100 ~l (~ 90 80 70 60 50 40 >C So ~ 20 Q 10 0 0 2,0 4,0 6.0 8,0 10,0 TINE, SECONDS Palo Verde Nuclear Generating Station TOTAL I.OSS OF REACTOR COOLANT FLOW CORE HEAT FLUX YS"TIYiE Figure 7.3. l-3 1,2 1,1 1,0 0 1,0 2,0 3,0 TIl'1Ei SECONDS TOTAL LOSS OF REACTOR COOLANT FLOW MINIMUM DNBR VS TIME Fulgur e 7.3.I 4 0 120 100 80 60 00 20 0 0 20 TIYiE, SECONDS Palo" Verde ~Nuclear Generating Station ASYMMETRIC STEAN GENERATOR EVENTS CORE POVER VS TIME Figure 7.7. 1-1 120 100 . 80 60 40 20 0 0 12 16 20 TINE, SECONDS Palo Verde Nuclear Generating Station ASYMMETRIC STEAM GENERATOR EYENTS CORE HEAT FLUX VS TIME = Figure 7.7. 1-2 700 650 CORE OUTLET CORE AVERAGE 600 CORE INLET ~e 550 CO CO 500 5 400 0 16 20 TIYiEi SECONDS Palo Verde huclear,Generating Station ASYMMETRIC STEAM GENERATOR EVENTS REACTOR COOLANT TEMPERATURES VS TIt~iE , Fi gur e 7.7. 1-3 2300 2200 2100 2000 1900 1800 1700 0 12 20 TINE, SECONDS Palo Yerde Nuclear Generating Station 1 p ASYMMETRIC STEAM GENERATOR EVENTS REACTOR COOLANT SYSTEM PRESSURE VS TIME Figure 7.7.1-4 1500 AFFECTED SG 1200 1100 1000 900 UNAFFECTED SG 800 700 0 12 20 TINEA'ECONDS r Palo Verde Fiuclear Generating Station ASYMMETRIC STEAN GENERATOR EVENTS STEAM GENERATOR PRESSURES VS TIME Figure 7.7.1-5 8.0 ECCS ANALYSIS S.l LARGE BREAK LOSS-OF-COOLANT ACCIDENT Introduction And Summar ~ An ECCS performance analysis of the limiting break size was performed for PVNGS-1 Cycle 2 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Mater-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-1. The method of analysis and detailed results which support this value are presented herein. Method Of. Anal sis The ECCS performance analysis for PVNGS-1 Cycle 2 consisted of an evaluation of the differences between Cycle 2 and PVNGS-1 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 NRC in Reference 8-3. As in the reference cycle, the calculations performed for this evaluation used the NRC approved C-E large break ECCS performance evaluation model which is described in Refer ence 8-4 including the use of a more conservative axial power shape. The blowdown hydraulic calculations, refill/reflood hydraulics calculations, and steam heat transfer coefficients of the reference cycle apply cooling to PVNGS-1 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. t 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 evalua.ion 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 1000MWD/MTU. 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 Resul ts Significant core and system parameters for the reference cycle and PVNGS-1 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 cl,ad temperature. This limiting case results in a peak clad 'temperature of 1925 F, which is well below the acceptance limit'f 2200 F. The maximum local and core wide zirconium oxidation, as shown in Table 8.1-2, remain wel,l below the acceptance limit values of 17% and 1%, respectively. These results remain applicable for up to 400 tubes plugged ~er steam generator and a reduction in system flow rate )o 155.8X10 ibm/hr and a reduction in core flow rate to 151. 1 X 10 .ibm/hr. Table 8. 1-3 presents a list of the significant parameters displayed graphically for the limiting 1.0 DEG/PD break. 8.1.4 Conclusion The ECCS performance evaluation for PVNGS-1 Cycle 2 results in a peak clad temperature of 1925'F, a peak local clad oxidation percentage of 4.6% and a peak core wide clad oxidation percentage of less than 0.80% compared to the acceptance criteria of 2200'F, 17% and 1%, respectively. Therefore, operation of PVNGS-1 Cycle 2 at a core power level of 3876 Mwt (102% of 3800 Mwt) and a PLHGR of 13.5 kw/ft is in conformance -with 10CFR50.46. 8.2 SMALL BREAK LOSS-OF-COOLANT ACCIDENT A review of Cycle 2 fuel and core data confirmed that the reported small break loss-of-coolant accident results (Reference 8-9) for PVNGS-1 Cycle 1 bounds PVNGS-1 Cycle 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.Ni'T (102/ of 3800 MWT). This acceptable performance has been confirmed with up to 400 plugged tubes per steam generator. TABLE 8.1-1 PVNGS-1 Cycle 2 Core and System Parameters Reference Parameter Units Cvcl e ~Cele 2 Reactor Power 9 )02% of Nominal(mwt) 3876 3876 Average Linear Heat Rate 8 102K of 5. 64 5.68 Nominal (kw/ft) Peak Linear Heat Generation Rate (kw/ft) 14.0 13.5 Core Inlet Temperature (OF) 565.0 565.0 System Flow 'Rate (ibm/hr) 164x10 164xlO Core Flow Rate (ibm/hr) .159xlO 159x10 Gap Conductance (1)'Btu/hr ft 'F) 1514 1652 Fuel Centerline Temperature ('F) 3424.7 3493.7 Fuel Average Temperature ('F) 2175.0 2124.0 Hot Rod Gas Pressure (psia) 1129.0 1126.3 Hot Rod Burnup (mwd/mtu) 774 1000 Number of Steam Generator Tubes Plugged per Steam Generator 0 400 Augmentation Factor Function of 1.0 Elevation Minimum Ini .ial Containmen. 13e4 Pressure (psia) Containmen't Free Volume (ft ) 3.0(10 ) 3.0()0 ) Axial Peaking Factor 1.52 1.52 (1) Initial values at the limiting hot rod burnup as calculated by STRIKIN-II at the peak linear heat generation rate. TABLE 8. 1.-2 PANGS-1 Cycle 2 Limitino Break Size . 1.0 DEG/PD Cvcle 1 Cvcle c Peak Linear Heat Generation )4.0 13.5 Rate (kw/ft) Peak Clad Temperature (~F) 2091 1925 Time of Peak Clad Temperature 278.5 267.8 (Seconds) Time o Clad Rupture (Seconds) 88.2 98.9 Peak Local Clad Oxidation (+) 9.0 4.6 Total Core-h'ide Clad Oxidation (%) < 0.80 <.0:80 TABLE 8. 1-3 PVNGS-1 Cycle 2 Variables Plotted as a Function of Time for the Limitino Lar e Break Figure Variable Desionation Peak Clad Temperature 8. 1-1 Hot Spot Gap Conductance 8.1-2 Peak Local Clad Oxidation 8.1-3 Clad Temperature, Centerline Fuel Temperature, 8.1"4 Average Fuel Temperature and Coolant Temperature for Hottest Node Hot Spot Hea. Transfer Coefficient 8.1-5 Hot Rod 1nternal Gas Pressure 8.1-6 22GG 2000 18GG 1600 / f / 1400 1200 I 1000 PEAK CLAD TEMPERATURE NODE PEAK CLAD OXIDATIOH HOD= 400 100 2GG 300 400 TIME, Sc.CONDS ARIZONA i.0 x DOUBLE ENDED GUILLOTINE BREAK Fjg TJyg PVNGS 1 IN PUMP DISCHARGE LEG 8.1-1 CYCLE 2 PEAK CLAD TEMPERATURE 0 iiGQ 1406 120G 1000 BGC "OO 200 200 300 4GG DGO TIN=. S.:CONOS ARIZONA 1.0 x DOUBLE ENDED GUILLOTINE BREAK Figure PYNGS 1 IN PUMP DISCHARGE LEG 8.1-2 CYCLE 2 HOT SPOT GAP CONDUCTANCE PEAK CLAD TEhlPEPATUiE NODE PEAK CLAD OXIDATION IIODE i0 5 wean i00 2GG . 300 40G QvU T T i~= "--=CON S A RIZONA 1.0 x DOUBLE ENDED GUILIOTINE BREAK FIaur~ PVNGS 1 IN PUMP DISCHARGE LEG 5.1-5 CYCLE 2 PEAK LOCAL CLAD OXIDATION 2 (GG, I I 21GG FUEL CENTERLINE 1800 AYERAGE f'UEL CLAD I- 12GG ="OG EGG COOLANT 3GG 2GG 300 SGG 6GG TII'!t. SECONOS ARIZONA 1.0 x DOUBLE ENDED GUILLOTINE BREAK- Figure PVNGS 1 IN PUMP DISCHARG" LEG S. i -4 CYCLE 2 CLAD EMP:RATURE, CENTERLINE F'UEL TEMPERATUR, AVERAGE .F UEL TEMPERATURE AND COOLANT TEMPERATURE FOR HOTTEST NODE 12G a~ I e4 = I I I IGG I CD L' . ZG JGG 2GG 3GG EGG "=r Gjrg ARIZONA 1.0 x DOUBLE ENDED GUILLOTINE BREAK PYNGS 1 IN PUMP DISCHARGE LEG 2 HOT SPOT HEAT TRANSFER COEFFICIENT 'YCLE 'I ~ Fwmu=1126.3 PSIA I RUPTUR =98.85 SEC 0.9 -'~ O.B q tL I 0.7-gC ~Q 0,6 33 NO N-fg O.S 6. J 0 , 0.2 0,$ I 7 I l I 20 Bp lpp 120 C ARIZONA j.0 x DOUBLE ENDED GUILLOTINE BREAK Figure PVNGS 1 IN PUMP DISCMARGE LEG 8.1-o YCLE 2 HOT ROD INTERNAL GAS PRESSURE 'EACTOR 9.0 PROTECTION AND MONITORING SYSTEM INTRODUCTION The Core Protection'alculator 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 melting are not exceeded during Anti ci pated Operational Occurrences (AOO's) 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 0" Limiting Conditions for Operation (LCO's). CPCS SOFTWARE MODIFICATIONS The CPC/CEAC software for PVNGS-I 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),. These changes have been designed to be applicable to all CPC plants. The CPC/CEAC algorithms for PVNGS-1 Cycle 2 are the same as those implemented at WSES-3 Cycle 2, at SONGS-2 and -3 Cycle 3, and at ANO-2 Cycle 6 and described in References 9-1 and 9-2. The revised list of addressable constants is defined in Reference 9-3. The I Reload Data Block (Reference 9-5) is'a group of constants that is located in protected memory of the CPC and the CEAC, separate from other non-addressable constants. The RDB constants are loaded from P a separate RDB disk and can be changed without requiring a CPC/CEAC software change. The RDB has previously been implemented at USES-3 for Cycle,2. The modifications for PVNGS-1 Cycle 2 relative to the Cycle 1 software are described in References 9-3, 9-4, 9-5 and 9-9. The described in References 9-3, 9-4, and 9-9 are 'odificat'ions incorporated in References 9-3. 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-1 Cycle 2 consistent with the cycle design, performance and safety analyses. The RBD'onstants will be installed on the Cycle 2 RDB Disk for loading at the site as described in Reference 9-8. 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 (BERRQ 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, calibra'tion constants, etc.). Trip setpoints, uncertainty factors and other addressable constants will be determined f'r PVNGS-1 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). I 9.4 OIG ITAL MONITORING SYSTEM COLSS The Core Operating Limit Supervisory System (COLSS), as described in Reference 9-11, is a monitoring system that initiates alarms if the LCO's on DNBR, peak linear heat rate, core power, or core azimuthal tilt are exceeded. The COLSS data base and uncertainties will be updated to reflect the Cycle 2 core design. i 10.0 Technical Specifications This section provides a summary of recommended changes that should be made to the PVNGS-1 Technical Specifications in order to update the Technical Specifications for Cycle 2 operation.' description of each change and the corresponding technical specification section . are presented in the following pages. 10-1 ATTACHMENT NO. 1 PROPOSED AMENDMENT TO PVNGS UNIT ONE CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 5.

3.1 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment increases the maximum enrichment from 4.0 to 4.05 weight percent U-235 as set forth in Technical Specification (T,.S- ) 5.3.1.

JUSTIFICATION FOR PROPOSED AMENDMENT To support the desired fuel management design of Cycle 2 and to allow future 18-month equilibrium cycles, the maximum peak pin enrichment, as stated in T.S. 5.3.1, will be increased to th'e value of 4.05 weight percent U-235. By increasing the enrichment to 4.05, PVNGS will be able to meet the long term goal of 18 month equilibrium cycles without adversely affecting safety margins.

10-2

ATTACHMENT NO. 2 PROPOSED AMENDMENT TO PVNGS UNIT ONE 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 6.0% 6 p to 6.5% c. 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 6% c p to 6.5% 4 p at zero power. The increase in margin is required to maintain the operation of Cycle 2 within the safety analysis.

10-3

ATTACHYZNT NO. 3 PROPOSED AYENDMENT TO PVNGS UNIT ONE CYCLE TWO TECHNICAL 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 operation 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 p'ositive 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.

10-4

ATTACHMENT NO. 4 PROPOSED AMENDMENT TO PVNGS UNIT ONE 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.

10-5

ATTACHMENT NO. 5 PROPOSED AYiENDMENT TO PVNGS UNIT ONE CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.1.3.1 and 3.1.

3.2 DESCRIPTION

OF PROPOSED AIKNDMENT 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 p'osition.

10-6

ATTACHMENT NO. 6 PROPOSED AMENDMENT TO PVNGS UNIT ONE CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

3.1 DESCRIPTION

OF PROPOSED AMENDYiENT 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 momen" the CEDM coil breakers open.

JUSTIFICATION FOR PROPOSED AMENDMENT H

During the Cycle 1 startup testing, it was found that the projected Reacto" 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 analysi:s, Table 3.3-2 will have to reflect the credited response time as it was used in the safety analysis.

10-7

ATTACHMENT NO. 7 PROPOSED AMENDMENT TO PUNGS UNIT ONE CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 2.l.l.l,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 of 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 factor 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's such that, by applying a rod bow penalty factor of 1.75K 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

'han 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 reevaluation 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'ooldown of the reactor coolant. The analysis has shown, that a pressu'rizer pressure of 1860 instead of 1861 will ensure that, if a reactor trip occurs'n Low-DNBR, the plant will not reach the Specified Acceptable Fuel Design Limits (SAFDLs).

10-8

ATTACHMENT NO. 8 PROPOSED AMENDMENT TO PVNGS UNIT ONE CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AYZNDED 3.1.

3.6 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment revises the CEA Insertion Limits as set fortn 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. Tne revisions to the curves will 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 made 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.

10-9

ATTACHMENT NO ~ 9 PROPOSED AMENDMENT TO PVNGS UNIT ONE 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 sh'ould 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'eleted and Table 3.3-2 should be revised to remove the penalty factor allowances.

10-10

ATTACHMENT NO. 10.

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

2.5 DESCRIPTION

OF PROPOSED AMENDMENT Tne 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 10 ibm/hr to greater than or equal to 155.8 x 10 ibm/hr.

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 100% 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 95% of the design flow of 155.8 x 106 ibm/hr. The 100% design flow rate of 164.0 x 106 ibm/hr conservatively accommodated the maximum instrument uncertainty of 4%, 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.

10-11

ATTACHMENT NO. 11 PROPOSED AMENDMENT TO PVNGS UNIT ONE 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 AYiENDMENT 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'HR are from increased fuel enrichment and the core loading pattern.

In addition to changing the 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.

10-12

ATTACHMENT NO. 12 PROPOSED AMENDMENT TO PVNGS UNIT ONE CYCLE THO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE'MENDED 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) removes 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-(1) 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-(1) 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 approved 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).

10-13

ATTACHMENT 'O. 13 PROPOSED AMENDMENT TO PVNGS UNIT ONE - CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED 3.

2.7 DESCRIPTION

OF PROPOSED AMENDMENT The proposed amendment provides clarification and consistency for monitoring the Axial Shape Index (ASI) as set forth in Technical Specifi.cation (T.S.)

3.2.7.

JUSTIFICATION FOR PROPOSED AMENDMENT The proposed change to T.S. 3.2.7 will eliminate the need to change this Technical Specification for each cycle. By deleting the COLSS operable numerical references, any small deviations in the cycle analysis to the COLSS ASI alarms limits will not require submitting an amendment request. The change also clarifies and provides a more consistent approach to monitoring ASI.-

10-14

ATTACHMENT NO. 14 PROPOSED AMENDMENT TO PVNGS UNIT ONE 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.ST 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.

10-15

ATTACHMENT NO. 15 PROPOSED AMENDMENT TO PVNGS UNIT ONE - CYCLE TWO TECHNICAL SPECIFICATIONS SECTION OF TECHNICAL SPECIFICATIONS TO BE AMENDED Various Bases DESCRIPTION OF PROPOSED AtiENDMENT 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 3-1 Update to the latest revision used for controlling the changing of the Core Protection Calculators software.

2) 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 administr'ative changes are required to ensure clarity and conciseness.

The change to Bases 3/4.3.1 updates the T.S. to the latest xevision of the Topical CEN-39(A)-P, "CPC Protection Algorithm Software Change Procedure" and 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 tiip 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.

10-16

i The planned startup test program associated with core performance is outlined below. The described tests verify that core performance is consistent with the engineering design and -safety analysis. The program conforms to ANSI/ANS-19.6.1-1985, "Reload Startup Physics Tests for Pressurized Water Reactors" and supplements normal surveillance tests which are required by Technical Specifications (i.e., CEA drop time testing, RCS flow measurement, MTC verification, etc.).

11.1 Low Power Ph sics Tests 11.1.1 Initial Criticalit Initial criticality will be achi.eved 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 withdrawn, the boron concentration of the reactor coolant 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 configurati'on and for a partially rodded configuration. The measured CBC values will be verified to be within +1% Delta-k/k of

,the predicted values.

'1.1.3 Tem erature Reactivit Coefficient 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 values will be verified to be within +0.3 x 10 4 Delta-k/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 worths. All full-length CEAs will be included in the measurement. Due to the large

~ differences in CEA group worths, two reference groups (one 11-1

i 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 will be as specified in Reference 11-2.

11.1.5 Inverse Boron Worth (IBW)

The IBW will be calculated using results from the CBC measurements, and the CEA group worth measurements. The calculated IBW value will be verified to be within +15 ppm/% delta-k/k of the predicted value.

11.2 Power Ascension Testing Following completion of the Low Power Physics Test sequence, reactor power will be increased in accordance with normal operating procedures. The power ascension will be, monitored through use of an off-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 will be suspended while necessary data reduction and equipment calibrations are performed. The follow'ing measurements will be performed during the program.

11.2.1 Flux S mmetr Verification Core power distribution, as determined from fixed incore detector data, will be examined prior to exceeding 30%

power to verify that no detectable fuel misloadings exist. Differences between measured powers in symmetric, instrumented assemblies will be verified to be within 10%

of the symmetric group average.

11.2.2 Core Power Distribution Core power distributions derived from the fixed incore neutron detectors will be compared to predicted distributions at two power plateaus. 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)

'ill provides reasonable assurance that the power distribution remain within the design limits'hile reactor power is increased to 100%, where the second comparison will be performed.

11-2

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 5X.

B. The RMS of the difference between the measured and predicted core average axial power distribution for each axial node will be verified to be less than or equal to 5X.

C. The measured values of planar radial peaking factor (Fxy), integrated radial peaking factor (Fr), core average axial peak (Fz), and the 3-D power peak (Fq)

'will be verified to be within +lOX of their predicted values.

11.2.3 Sha e Annealin Matrix (SAM) and Boundar Point Power Correlation Coefficients BPPCC Verification The SAM and BPPCC values will be determined from a linear regression analysis of the measured excore detection readings and c'orresponding 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 throughout 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. excare 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 will be modified based upon the calculated .'onstants 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.

11-3

11.2.5 Tem erature Shadowin Factor Verification The effect of cold leg temperature upon excore detector response will be determined by recording excore detector signals and reactor power while varying cold leg temperature. The results- will be used to verify the adequacy of CPC data base constants.

11.2.6 Reactivit Coefficients 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 " Delta-k/k 'F and +0.3 x 10 "

Delta-k/k/% power, respectively, of predicted values.

11.2.7 Critical Boron Concentration The CBC will be determined for conditions of full power, equilibrium xenon. The measured CBC will be verified to be within +50 ppm of the predicted value after adjustment for the bias observed between measured and predicted CBC values at zero power. 0 11.3 Procedure If Acce tance Criteria Are Not Met The results of all tests will be reviewed by the plant's reactor.

engineering group. If the acceptance criteria of the startup physics tests are not met, an evaluation will be performed with assistance from the fuel vendor as needed. The results of this evaluation will be presented to the Plant. Review Board.

Resolution will be required prior to subsequent power escalation.

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

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.

11-4

References Section 1.0 References f

"Palo Verde Nuclear Generating Station Unit No. 1, Final Safety Analysis Report," Arizona Public Service Company, Docket No. 50-528.

f (1-2} "Combustion Engineering Standard Safety Analysis Report CESSAR)", Docket ¹STN-50-470F.

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.

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

Letter, A. E. Lundvall, Jr. to J. R. Miller (Chief Operating Reactors Branch ¹3), "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 and 2, Docket Nos. 50-317 and 50-318, Request for Amendment", December 31, 1984.

(4-4) 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.

P' "Safety Evaluation by the Office of Nuclear Reactor Regul ati on Related to Amendement 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.

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

(4-7) CEN-161(B)-P, "Improvements to Fuel Evaluation Model,"

July, 1981.

R. A. Clark (NRC) to A. E. Lundval l, Jr. (BGE E), "Safety Evaluation of CEN-161 (FATES3)," March 31, 1983.

(4-9) "Combustion Engineering Standard Safety Analysis Report (CESSAR)", Docket ASTN-50-470F.

"Palo Verde Nuclear Generating Station Unit No. 1, Final Safety Analysis Report," Arizona Public Service Company, Docke. No. 50-528, Section 4.2.4.

(4-11) CESSAR SSER2, Section 4.2.5, "Guide Tube Wear Surveillance".

(4-12) CESSAR SSER8, Section 4.2.4, "Fuel Assembly Surveillance."

(4-13) CESSAR SSER5, Section 4.2.4, "Fuel Rod Growth".

Section 5.0 References (5-1) 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.

r A. E. Lundvall (BGSE) 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 Rev. 1-P-A, "INCA/CECOR Power Peaking

'5-4)

CENPD-153-P, Uncertainty, " May, 1980.

(5-5) CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design," April, 1983.

Section 6' References (6-1) CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April, 1986, CENPD-162-A, "Critical Hea. Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," September, 1976.

(6-3) CEN-160(S)-P, Rev. 1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3,"- September, 1981.

(6-4) CEN-356(V)-P, "Modified Statistical Combination of Uncertainties," May, 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, Statistica.l Combination of Uncertainties.

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 Engineein'g Standard Safety Analysis Report," Docket No. 50-470.

(7-3) "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.

R, V. Macbeth, "An Appraisal of Forced Convection Burnout Data," Proc. Instr. Mech. Engrs., Vol. 180, Pt. 3C, PP 37-50, 1965-1966.

D. H. Lee, "An Experimental Investigation o Forced Convection Burnout in High Pressure Water - Part IV, Large Diameter Tubes at about 1600 psia," A.E.E.W. Report R479, 19866.

(7-7) CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs 1 and 2," December 1981.

(7-8) CEN-308-P-A Revision QO-P, "CPC/CEAC Software Modifications for the CPC improvement Program," April 1986.

(7-9) CENPD-188-A, "HERMITE Space-Time Kinetics," July 1975.

(7-10) CENPD-161-P, "TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core;" July 1975.

(?-11) CENPD-206-P, "TORC Code. Verification and Simplified Modeling Methods," January 1977.

(7-12) CENPD-183-A, "Loss of Flow - C-E Methods for Loss of Flow Analysis," June 1984.

(7-13) CENPD-199-P-A, Rev. 1-P, "CE Setpoint Methodology,"

January, 1986.

(7-14) Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 47 to NPF-10 and Amendment No. 36 to NPF-15, Southern California Edison Company, et. al., San Onofre Nuclear Generating Station, Units 2 and 3 ~

Sec.ion 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.

Letter from C. Ferguson (C-E) to E. E. Van Brunt, Jr.

(ANPP), V-CE-32964, "License Condition 21-Large Break LOCA," October 1, 1985.

(8-3) (NRC approved of the PVNGS-1 Cycle 1 Performance Results of (8-2). To be supplied by 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'he C-E Large Break LOCA Evaluation Model", July 1975.

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

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

February 1975.

CENPD-135-P, Supplement 4P. "STRIKIN-II, A .Cylindrical Geometry Fuel Rod Heat Transfer Program", August 1976.

(8-6) CENPD-139-P-A, "C-E Fuel Evaluation Model" July, 1974.

(

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

July, 1981.

(8-8) Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr., I (BG8E), "Safety Evaluation of CEN-161 (FATES 3)," March 31, 1983.

(8-9) 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.

(8-10) (NRC approval of License Condition 21 ECCS performance results of (8-9). To be supplied by PVNGS.)

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

(9-6). 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. Ol-P, "CPC/CEAC Software Modifications for San Onofre Nuclear Steam Generating Station Units No.

2 and-3," November, 1984.

~P (9-10) CEN-356(V)-P, Rev. OO-P, "Modified Statistical Combination of Uncertainties," May, 1987.

(9-11) CEN-312-P, Rev. 0 1-P, "Overview Description of the Core Operating Limit Supe'rvisory System (COLSS)," November, 1986 '

'C a