ML17305B376
| ML17305B376 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 02/21/1991 |
| From: | ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | |
| Shared Package | |
| ML17305B374 | List: |
| References | |
| NUDOCS 9102280044 | |
| Download: ML17305B376 (147) | |
Text
0 ATTACHMENT 5 RELOAD ANALYSIS REPORT 910221 j
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Cl RELOAD ANALYSIS REPORT FOR PALO VERDE. NUCLEAR 'GENERATING STATION. UNIT 3 CYCL'E 3 .
TABL'E OF CONTENTS PAGE
- 1. INTRODUCTION AND
SUMMARY
- 2. OPERATING HISTORY OF THE REFERENCE CYCLE 2-1
- 3. GENERAL DESCRIPTION 3-1
- 4. FUEL SYSTEM DESIGN 4-1
- 5. NUCLEAR DESIGN
- 6. THERMAL-HYDRAULIC DESIGN 6-1
- 7. TRANS I ENT 'NALYSIS 7-1
- 8. ECCS ANALYSIS 8-1
- 9. REACTOR PROTECTION AND MONITORING SYSTEM, 9-1
- 10. TECHNICAL SPECIFICATIONS 10-1
- 11. STARTUP TESTING ll-l
- 12. REFERENCES 12-1,
0 4
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INTRODUCTION AND
SUMMARY
This report provides an evaluation of the design and performance of Palo Verde Nuclear Generating Station Unit 3 (PVNGS-3) during its third cycle of operation at 1001 rated core power of 3800 .HWt and NSSS power of 3822 MWt. 'Operating conditions for Cycle 3 have been assumed, to be consistent, with .those. of,.the previous .,cycle..and are...
summarized as full power operation under base load conditions. The core will consist of irradiated Batch B, C and D assemblies, al'ong with fresh Batch E assembl.ies. The Cycle 2 termination burnup has been assumed to,be between 384 and 436 EFPD (Effective Full Power Days)..
The second cycle of operation will hereafter be referred to in .this report as the "Reference Cycle." Reference 1-2 presented analyses for the Reference Cycle.
The safety criteria (margins of safety, dose limits, etc.)
applicable for the plant were established in Reference l-l. A review of all postulated accidents and anticipated operational occurrences has shown that the Cycle 3 core design meets these safety criteria.
The Cycle 3 reload core characteristics have been evaluated with respect to the Reference Cycle. Specific differences in core fuel loadings have been accounted for .in the present. analysis. The status of the postulated accidents and anticipated operational occurrences for Cycle 3 can be summarized as follows:
- 1. Transient data are less severe than those of the Reference Cycle analysis; therefore, no reanalysis is necessary, or
- 2. Transient data are not bounded by those of the Reference Cycle analysis, therefore, reanalysis is required.
1-1
E
.For those transients. requiring reanalysis (Type '2), analyses.:are
,presented, in Sections -7 and 8 showing results that meet the established safety criteria.
The Technical Specification changes needed for Cycle 3 are summarized in Section 10 and .described .in greater, detail ..in separate license amendment. applications.
1-,2
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2.0...'OPERATING HISTORY OF THE REFERENCE CYCLE The Reference Cycle began with initial criticality on December 26, 1989. Power Ascension began on December 29, 1989 and on 'December 30, 1989, the unit reached full power.
It is presently estimated that=.Cycle 2 will terminate. on or:.about March 16, 1991. The Cycle 2 termination point can vary between 384 and 436 EFPD to accommodate the plant schedule and still be within the assumptions of the Cycle 3 analyses.
2-1
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3.0 GENERAL DESCRIPTION The Cycle 3 core will consist of those assembly types and numbers listed in Table 3-1. Seventy-three Batch B assemblies and forty-eight Batch C will be removed from the Cycle 2 core to make way for eighty-eight fresh,, Batch E assemblies. One hundred and.
four Batch. D and sixteen, Batch C. assemblies now in the core'il.l. be retained. In addition, thirty three Batch B assemblies originally discharged at EOCl will be reinserted from the spent fuel storage.
Figure 3-1 shows the poison shim and zoning configuration for the discarded assemblies.
The reload batch will consist of 4 type EO assemblies, 24 type El assemblies with 8 burnable poison shims per assembly, 4 type E2 assemblies with 16 burnable poison shims per assembly, 40 type E3 assemblies with 12 burnable poison shims per assembly, 16 type E4 assemblies with 8 burnable poison shims per assembly. These sub-batch types are fuel zone-enriched and their configurations are shown in Figure 3-2.
The loading pattern for Cycle 3, showing fuel type and location, is displayed in Figure 3-3.
Figure 3-4 displays the beginning of Cycle 3 assembly average burnup distribution. The burnup distribution is based on a Cycle 2 length of 436 EFPD.
Control element assembly patterns and in-core instrument locations will remain unchanged from the Reference Cycle and are shown in Figures 3-5 A 5 B and Figure 3-6, respectively.
3-1
TABLE 3-1 PALO VERDE .NUCLEAR GENERATING STATION UNIT 3 Cycle 3 Core Loading Initial Total Number Assembly Fuel Rods Ini ti al Number Shim of Desig- Number of per Enri chment Shims/ Loading Fuel'him nation Assemblies Assembly (w/o U-235), .Assembly (gm B10/in) Rods Rods 33 '208 2.78 .018 6864 528 12 1.92 396 16 .224 3.30 3584 12 2. 78 192 DO 32 184 .
3.90 5888 52 ,3.40 1664 D1. 40 172 3.40 '12 .026 6880 480 52 3.09 2080 168 ~ 3.90 16 .020 '4032 384 52 3.40 1248 r
172 3.90 12 .'024 1376 96 52 3.40 416 EO 184 3.96 736 0 52 3. 58. 208 El 176 3.96 .014 4224 192 52 3.13 1248 E2 168 '3.96 16 .020 672 64 52 3.13 208 E3 40 172 3.58 12 .026 6880 480 52 3.13 2080 E4 16 176 3.58 .028 2816 128 52 3.13 832 TOTAL 241 54524 2352 3-2
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FIGURE 3-1 ASSEMBLIES TO BE DISCHARGED FUEL AND BURNABLE POISON ROD PLACEMENT B-Lo, B-Hi, C-Lo CO FUEL Fuel No. of Enrich Enrich No. o Type Assemblies w/o w/o BPRs 8-j 0/in Q~"
B-Lo 2.78 1.92 16 .01 8 B-Hi 45 2.78 1.92 16 .025 CO 24 3.30 2.78 0 C-Lo 24 3.30 2.78 .012
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FIGURE 3-2 FRESH FEED FUEL ASSEMBI IES FUEL AND BURNABLE POISON ROD PLACEMENT EO FUEL E1 FUEL E2 FUEL E3 FUEL E4 FUEL Fuel No. of Enrich Enrich No. of gm Type Assemblies w/o w/o BPRs B-10/in 0 Q~ 5 EO 3.96 3.58 0 E1 24 3.96 3.13 0.014 E2 3.96 3.13 0.020 40 3.58 3.13 12 0.026 E4 16 3.58 3.13 0.028
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FIGURE 3-3 PVNGS-3 CYCLE 3 FUEL MANAGEME T Cycle 2 - 410'FPD Cycle 3 - 390 EFPD BX 02 EO C El DO El Dl DO 00 El 02 E3 C E2 B 00 E4 Dl DO 02 E3 00 C El 01 E3 01 E4 Dl f3 B El 02 , 00 01 BX Dl E3 BX BX DO E3 02 E4 Dl E4 D3 E3 02 El C E3 Dl E3 D3 E3 01 EO 01 E2 00 E3 'X E3 Dl BX 8 OF PIN ENRICH 8 ZONING SHIM LOADING 8 OF AVG. ASSY.
~GN 810 IN ASS Y ENRICHMENT EO 0 236 184 3.96 52 3.58 0.000 4 3.876 El 8 228 176 3.96 52 3.13 0.014 24 3.771 E2 16 220 168 3.96 52 3.13 0.020 4 3.764 E3 12 224 172 3.58 52 3.13 0.026 40 3.476 E4 8 228 176 3.58 52 3.13 0.028 16 3.477 88 3.588 3-5
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5 LEAD REGULATING BANK 4 SECOND 'R EG ULATING BANK 3 THIRD REGULATING BANK 2 ,FOURTHREGULATING BANK LAST REGULATING BANK B SHUTDOWN BANK B A SHUTDOWN BANK A P2 PLR GROUP 2 P1 PLR GROUP 1 S SPACE CEA LOCATIONS 6
S 10 12 13 'l4 15 16 17 18
- 1. 1 Al 19 20 1 22 23 24 25 26 27 28 29 30 31 4 2 P2, 3: P2: 2', 4 47 33 48 34 B
50 36 B.
37 52 38 53 B'0 39 54
'B 41 2 B1 43 45 46 51 55 56 57 59 60 61 2 P1 5 P1.
63 64 65 67 68 69 70 71 72 73 4'874 75 76 77 78 A B 4 A ., A B A .'
79 80 1 82 83 84 85 86 87 88 89 90 91 93 94 95 S P2: P.2, S:,
96 97 98 99 100 101 102 103 104 105 106 '07 108 109 110 112 1 B 3. 3 B 113 114 115 'I 16 117 118 1 19 'l20 '21 122 'l23 124 125 126 127 128 129 P1. 5 3 3 130 131 32 133 134 'l35 '36 137 138 139 140 141 142 143 144 145 146 1 B A 3 3 B 1 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 S P2 P.2 164 165 166 167 169 170 171 172 173 174 175 176 177 178 179 180 A B 4 A A 4 B A 181 182 2
183 184 P1 I85 186 187 188 5
189 190 191 192 P1 193 194 2'95 196 I, 197 198 199 200 201 202 i 203 204 205 206 207 208 209 210 B B B B 211 212 213 2'l4 215 216 217 218 219 220 221 222 223 4 2 P2 3 P2 2. .4 224 ~ 225 226 227 228 229 230 231 232 233 234 A 1 1 A 235 236 237 238 239 240 241 3 S ARIZONA Figure Palo Verde Nuclear Generating CEA BANK IDENTIFICATION Station 3-5A 3-7
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S ARIZONA Palo Verde Nuclear Generating CEA PATTERN'igure 3-58 Station
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1, 8 10 11 '12 13 14 15 17 18 2" 4 32 19 33 20 34 5 22 36 5I 23 37 24 38 6
25 26 40 27 7
28 29 8.'1 9 39 42 43 45 46 10 11; 12 47 48 50 51 52 53 54 55 56 57 58 59 60 61 13 14, 15 16 62 63 64 65 67 68 '9 70 71 72 73 74 75 76 l
77 78 17; 18 19 79 80 1, 82 83 84 85 86 87 88 89 90 91 92 93 94 95 20 21- 22 23 24l 25'6 97 98 99 100 101 102 103 104 105 106 107 108 109 110 11 'I 112 26 27 28 113 I '14 1'15 116 1'17 118 'I 19 120 121 122 123 124 125 126 127 128 129 29; 30 31 32< 33 34..
130 131 32 133 134 'I 35 136 137 138 139 140 141 142 143 35 36. 37 38'44 145 146 147 148 '149 150 'I 52 '153 154 155 156 157 39'51 158 40
'I 59 160 16'I 162 163 41 164 165 166 167 '169 '170 171 172 173 174 175 176 177 178 179 180 42 43 44 45 46 181 182 183 184 185 186 187 188 189 191 192 193 194 47/ 48 49'90 50 195 19G. 197 198 199 200 201 202 203 204 205 206 207 208 209 210 51 52 53 211 212 213 2'14 215 216 217 218 219 220 221 222 223
- 54. 55 56 57 224 225 22G 227 228 229 230 231 232 233 234 58 59 60 235 236 237 238 239 240 241 61 AR IZONA Figure Palo Verde Nuclear Generating INSTRUMENT LOCATIQNSI Station 3-6 3-9
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4.0 FUEL SYSTEM DESIGN 4.1 MECHANICAL DESIGN 4.l.l The mechanical design: of,the'Batch,E reload. fuel'ssemblies i,s identical'o the design of the Reference Cycle Batch D reload fuel assemblies except for a modification to the poison rod assembly design, lower end fitting, and center guide tube design. No changes in mechanical design bases have occurred since the original fuel design.
A design feature was incorporated into Batch E to'-improve the-burnup capability of the poison rods. The poison rod assembly design was modified by increasing the overall length from 160.918 inches to 161.168 inches. This provides greater internal void" volume which enables higher burnups with poison rods with higher B-10 loadings while reducing end .of life internal pressure. In addition, this change makes the fuel and poison rods equal in length.
The lower end fitting design was changed from a two piece assembly to a single piece casting with a recess for the center guide tube to fit within the flow plate. This change will enhance the structural integrity of the fuel bundle assembly for all loading conditions.
The length of the center guide tube was increased from 163.715 inches to 163.965 inches in order to fit within the new lower end fitting.
4-1
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GUIDE TUBE WEAR Twenty of the fuel assemblies that had CEA's located in them during Cycle 1 at Palo Verde Unit 1 were inspected for guide tube wear.
That inspection was part of the required licensing procedures required by the NRC for all plants after the first cycle of operation (References 4-1, 4-7, and 4-8). .A similar program was also, performed on Unit 3"during"the first refueling. outage:: -'. *.,'-
(Reference 4-2 and 4-6). The number of assemblies inspected for guide tube wear was determined based on the results of the Unit 1 inspection. The inspections revealed that guide tube wear was minor and will not adversely affect the fuel assembly performance
,throughout'its expected life in the core. Thus no guide tube wear inspections are necessary.
4.3 THERMAL DESIGN The thermal performance of composite fuel pins that envelope the pins of fuel batches B; C, D and E present in Cycle 3 has been evaluated using the FATES3A version of the C-E fuel evaluation model (References 4-3 and 4-4). The analysis was performed using a power.
history that enveloped the power and burnup levels representative of the peak pin at each burnup interval,, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of Cycle 3. The rod internal pressure remains below the reactor coolant pressure throughout Cycle 3. The power to centerline melt limit has been determined to be in excess of 21 kw/ft.
CHEMICAL DESIGN The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch E fuel are the same as the Batch D fuel batches included in Cycle 2. Thus, the chemical or metallurgical performance of the Batch E fuel will. be similar to the Batch D fuel used in Cycle 2.
4-2
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SHOUL'DER GAP ADE UACY Heasured shoulder gap data (references 4-1 and 4-2) acquired. from post Cycle .1 inspection of fuel assembl'ies at 'PVNGS Units 1 and 2
.indicate that the .fuel had adequate..shoulder gap. for.-.Cycle .2 .
operation. The present, shoulder,gap:,is projected to be. adequate for
- Cycle 3 operation.
4-3
i 4 t II
NUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS
- 5. 1. 1 Fuel Hang ement
, The Cycle 3 core makes use ~of,.a,l,ow-.leakage,.fuel .management,.scheme,;
'in, which previously burned 'assemblies are placed on the'core periphery. Host of the fresh Batch E assemblies are located throughout the interior of the core where they are mixed with the previously burned fuel in a pattern that minimizes power peaking.
With this loading and a Cycle 2 endpoint at 418 EFPD, the Cycle 3 reactivity lifetime for full power operation is expected to be 390 EFPD. Explicit evaluations have been performed to assure applicability of all analyses to a Cycle 2 termination burnup of between 384 and 436 EFPD and for a Cycle 3 length up to 416 EFPD.
/
Characteristic physics parameters for Cycle 3 are compared to those of the Reference Cycle in Table 5-1. The values in this table are intended to represent nominal core parameters. Those values used in the safety analysis (see Sections 7 and 8) contain appropriate uncertainties, or incorporate values to bound future operating cycles, and in all cases are conservative with respect to the values reported in Table 5-1..
Table 5-2 presents a summary of CEA reactivity worths and allowances for the end of Cycle 3 full power steam line .break transient with a comparison to the Reference 'Cycle data. The full power steam line break was. chosen to illustrate differences in CEA reactivity worths for the two cycles.
5-1
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The CEA core locations and group identifications remain the same. as in the Reference Cycle. The power dependent insertion limit .(PDIL) for regulating groups and part length CEA groups is shown in Figures 5-1 and 5-2 respectively. Table 5-3 shows the reactivity worths of various CEA. groups calculated .at full power conditions...for:Cycle 3 and the Reference Cycle.
- 5. 1.2 Power Distribution Figures 5-3 through 5-5 illustrate the calculated All Rods Out (ARO) relative assembly power densities during Cycle 3. The one-pin planar radial, power peaks (Fxy) presented in these figures represent the maximum over the mid eighty percent of the core axially. Time points at the beginning, middle, and end of cycle were chosen to display the variation in assembly and maximum planar radial peaking .
as a function of burnup.
Relative assembly power densities for rodded configurations are given for BOC and EOC in Figures 5-6 through 5-11. The rodded configurations shown are those allowed by the PDIL at full power:
part length CEAs (PLCEAs), Bank 5, and 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 1.22 at BOC3 to 1.08 at EOC3.
5-2
'l PHYSICS ANALYSIS METHODS 5.2. 1 ,Anal tical In ut to In-Core Measurements In-core detector .measurement constants to,be used .in. evaluating,;the reload cycle power distributions will be calculated .in accordance with Reference 5-1.. The ROCS; and'C',codes,.empl.oying,DIT. cal.culated
..~ cross.,sections-,wi.l.l'.be. used. - ROCS,,HC,-'and DIT:have-been:.approved~<<
for this application in Reference 5-2.
5.2.2 Uncertainties in Measured Power Distributions The planar radial, power distribution measurement uncertainty of 5.3%, based on Reference 5-1, will be applied to the Cycle 3 COLSS and CPC on-line calculations '.which use planar radial power 'peaks.'he axial and three dimensional .power distribution measurement uncertainties are determined in conjunction with other monitoring
.and protection system measurement uncertainties,,as,was done for Cycle 2.
5.2.3 Nuclear Desi n Hethodolo The Cycle 3 nuclear design was performed. with two and three dimensional core models using the ROCS and HC computer codes employing DIT calculated cross sections. ROCS, MC, and'IT were described in Reference 5-2.
5-3
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TABLE 5-1 PVNGS-3 CYCLE 3 NOMINAL PHYSICS CHARACTERISTICS Reference Dissolved Boron Units ~Ccl e ~Cele 3 Dissolvedl Boron Concentration for Criticality, CEAs Withdrawn, Hot Full Power PPH 1083 1153 Equilibrium Xenon, BOC Boron Worth Hot Full Power, BOC PPH//hp 114 119 Hot Full Power, EOC PPH//8p 97 91 Moderator Tem erature Coefficients Hot Full Power, Equilibrium Xenon Beginning of Cycle 10-4hp/ F -'0. 5 -0.5 End of Cycle 10-4hp/ F -1.9 -3.2
,,Hot Zero Power, Beginning of Cycle 10-4hp/'F +0.3 +0.47 Do ler Coefficient Hot Zero Power, BOC 10-5hp/ F -2.0 -2.1 Hot Full Power, BOC 10-5hp/ F -1.6 -1.7 Hot Full Power, EOC 10-5hp/ F -1.9 -1.9 Total Dela ed Neutron Fraction eff BOC .0065 .0062 EOC .0053 .0051 Prom t Neutron Generation Time 1*
BOC 10-6 sec 22.6 20'. 9 EOC 10-6 sec ,28.8 28.2 5-4
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TABLE 5-2
'PVNGS-3 CYCLE 3 L'IHITING VALUES OF REACTIVITY WORTHS, AND ALLOWANCES FOR HOT FULL POWER STEAN LINE BREAK, Qp) END-OF-CYCLE (EOC)
Reference
~Ccl e ~Cele 3 Worth of all CEAs Inserted -16.3 -17.5 Stuck CEA Allowance +4.5 +3.8 Worth of all CEAs Less Highest Worth CEA Stuck Out -11.8 -13.7 Full Power Dependent Insertion Limit CEA Bite +0.2 +0.25
'Calculated Scram Worth -11.6 -13.5 Physics Uncertainty +1.2 +1.3 Other Allowances +0.1 +1.3 Net Available Scram Worth -10.3 -10.9 Scram Worth Used in Safety Analysis -10.0 -10.2 5-5
0 TABL'E 5-3
-PVNGS-'3 CYCLE 3 REACTIVITY WORTH OF 'CEA REGULATING 'GROUPS AT HOT FULL POWER, %%u8p Be innin of C cle ~Fd f Regulating Reference Reference CEAs ~Ccl e ~Cele 3 ~Ccl e ~Cele 3 Group 5 -.26 ~ 33 - e32 -.41 Group 4 -.45 -.33 -.48 -.35 Group 3 -.76 -.93 -.80 -1.13 Note:
Values shown assume sequential group insertion.
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CER INSERTION LIMITS VS. THERMRL PQWER CYCLE 3 (CQLSS IN SERVICE) 1.0 0.9 1 g
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I Vl I-g o o 0.5 TAANSIENT INSEATION LIMIT o o4 0 l3 X
"I 0.3 0 0.2 0.1 0.0 BANK BANK 5 BANK 3
'3P " 0 150 120 90 60 30 0 150 120 90 60 30 P 150 120 90 6P BANK 4 BANl( 2 150 120 90 60 30 0 150 120 90 60 30 0
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PRRT LENGTH CER POSITION VS.
THERHRL POWER 1.0 UNRCCEPTRBLE OPERRTION 0.8 RCCEPTRBLE OPERRT ION 0.6 0.50 0.4 LONG TERM TRRNS IENT INSERTION INSERTIDN LIMIT LIMIT IQ I 0.2 OJ I I
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0 .0 150 140 130 120 110 100 90 80 70 60 50 40 30 20 10 PRRT LENGTH CER POSITION (Inches Wi t;hdrawn)
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'FIGURE 5-3 PALO VERDE U3C3 Fxy = 1.52 Box = 47 0.29 0.41 0.58 0.77 10 0.29 0.50 0.95 0.99 1.17 0.93 12 13 14 15 16 17 0.47 0.78 .1.16 1.13 1.27 0.98 1.24 18 19, ,20 22 23 24 25 0.29 0.78 1.16 1.08 1.30 1.22 1.31 1.17 26 27 28 29 30 31 32 33 0.50 1.16 1.08 1.25 1.07 1.34 1.12 1.30 34 35 36 37 38 39 40 41 42 0.29 0.95 1;13 1.30 1.07 0.96 1.07 1.28 1.02 44 45 46 47 48 49 50 0.41 0.99 1.27 1.22 1.34 1.07 1.32 1.16 1.26 52 53 54 55 56 57 58 59 60 0.58 1.17 0.98 1.31 1.12 1.28 1.16 1.25 1.03 61 62 63 64 65 66 67 68 69 0.77 0.93 1.24 1.17 1.30 1.02 1.26 1.03 0.91 ARO Assembly ReIative Power Densities at Hot Full Power with Eq. Xe.
,BOC3
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FIGURE 5-4 PALO VERDE U3C3 Fxy = 1.52 Box = 63 0.33 0.46 0.61 0;79 '0 0.32 0.52 0.97 0.98 1.20 0.97 12 13 14 15 16 17 0.48 0.76 1.12 1.06 1.27 1.00 1.34 '5 18 19 20 21 22 23 24 0.32 0.76 1.10 1.00 1.16 1.16 1.34 1.18 26 27 28 29 30 31 32 33 0.52 1.12 1.00, 1.20 1.00 1.29 1.12 1.36 35 37 38 39 40 41 42 0.33 '3 0.97 1.06 1.16 1;00 0.94 1.05 1.34 1.09 45 46 47 48 49 50 51 OA6 0.98 1.27 1.15 1.29 1.05 1.32 1.18 1.35 52 53 54 55 56 57 58 59 60 0.61 1.20 1.00 1.34 1.12 1.34 1.18'7 1.33 1.09 62 63 64 65 66 68 69 0.79 '.97 1.34 1.18 1.36 1.09 1.35 1.09 0.99 ARO Assembly Relative Power Densities at Mot Full Power with Eq. Xe.
MOC3 5-10
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FIGURE 5-5 PALO VERDE U3C3 Fxy = 1.51 Box = 63 0.38 0.50 0.64 0.80 10 0.37 0.56 1.00 0.99 1.18 0.96 12 13 16 17 0.53 0.80 1.12 1.04 1.27 1.00 1.35
. 18 19 20 21 22 23 24 25 0.37 0.81 1.14 0.99 1.10 1.33 1.15 26 27 28 29 30 31 32 33 0.56 1.12 0.99 1.21 0.98 1.25 1.09 1.35 34 35 36 37 -
38 39 40 41 42 0.38 1.00 1.04 1.10 0.97 0.92 1.02 1.33 1.08 43 44 45 46 47 48 49 50 '51 0.50 0.99 1.27 1.25 1.02 1.29 1.15 1.35 52 53 54 55 56 57 58 "'59 60
'0.64 1.18 1.00 1.33 1.10 1.33 1.14 1.32 1.07 61 62 63 64 65 66 '67 68 69 0.80 0.96 1.35 1.15 1;35 1.08 1.35 1.07 0.97 ARO Assembly Relative Power Densities at Hot Full Power with Eq. Xe.
EOC3 5-1 1
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FIGURE 5-6 PALO VERDE U3C3 Fxy = 1.60 Box =.13 5
0.30 0.42 '.59 0.81 10 0.31 0.52 1.02 1.05 1.22 0.96 13 14 15 16 17 0.51 0.85 1.28 1.21 1.33 0.97 1.22 18 19 20 21 22 23 25 0.31'6 0.85 1.28 1.17 1.40 1.24 1.23 1.02 27 28 29 30 31, 32 33 0.52 1.28 1.17 1.35 1.33 0.96 0.75 34 35 36 37 38 39 40 41 42 0.30 1.02 1.21 1 40 0.94 1.03 1.17 0.85 43 44 45 46 47 48 49 50 51 52'.0553 0.42 1.33 54 1.24 55 1.33 56 1.03 1;29 58 1.10 59 1.19 60 0.59 61 1.22 62 0.97 63 1.23 64 65'.1766 0.96 1.10 67 1.21 68 0.99 69 0.81 0.96 1.22 1.02 0.75 0.85 1.19 0.99 0.86 Assembly Relative Power Densities at Hot Full Power with Eq. Xe.
.BOC3 With Bank 5 Inserted 5-12
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'PALO VERDE U3C3 Fxy ~ 1.59 Box = 49 0.27 0.38 0.56 0.77 10 0.28 0.47 0.92 0.95 1.16 0.92 12 13 14 16 0.46 0.76 1.16 1.09 1.14 0.94 1.24 18 19 20 21 22 23 24 25 0.30 0.76 1.15 1.05 1.30 1.21 1.33 1.20 26 27 28 29 30 31 32 33'4 0.47 1.16 1.06 1.14 1.06 1.39 1.15 1.36 35 36 37 38 39 40 41 42 0.27 0.92 1.09 1.30 1.06 0.96 1.36 1.06 43 44 45 46. 47 48 49 50 51 0.38 0.95 1.14 1.21 1.39 1.42 1.23 1.35 52 53 54 55 56 57 58 59 60 0.56 1.16 0.94 1.33 1.16 1.36 1.23 1.33 1.07 61 62 63 64 65 66 67 68 69 0.77 0.92 1.24 1.20 1.36 1.06 1.35 1.07 0.86 Assembly Relative Power Densities at Hot Full Power with Eq. Xe.
BOC3 with PLCEA's Inserted 5-13
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FIGURE 5-8 PALO VERDE U3C3 Fxy = 1.58 Box = 27 0.29 0.41 0.59 0.82 10 0.31 0.52 1.00 1.01 1.21 0.96 12 13 15 16 17 0.51 0.85 1.27 1.17 1.19 0.95 1.22 18 19 20 21 22 23 24 25 0.31 26 0.85 27 1.27 28 1.14 29 1.37 30 31'.2332 1.22 1.03 33 34'.5235 1.27'6 1;14 37 1.21 38 1.09 39 1.36 40 1.00 41 0.78 42 0.29 1.00 1.17 1.37 1.09 0.96 1.08 1.23 0.90 43 44 45 46 47 48 49 50 51 0.41 1.01 1.19 1.22 1.36 1.08 1.37 1.18 1.27 52 53 54 55 56 57 58 59 60 0.59 1.21 0.95 1.24 1.00 1.23 1.17 1.29 1.04 61 62 63 64 65 66 67 68 69 0.82 0.96 1.22 1.03 0.78 0.90 1.27 1.04 0.84 Assembly Relative Power Densities at Hot Full Power with Eq. Xe.
BOC3 with. Bank 5 and PLCEA's Inserted 5-14
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~ t lr
FIGURE 5-9 PALO VERDE U3C3 Fxy=1 48 Box = 17 0.39 0.51 0.67 0.85 10 0.39 12 0.60 13 1.09 14 1.05 15 16'.00 1.25 0.58 0.88 1.24 . 1.12 1.35 1.00 1.37 19 20 21 22 23 24 25 0.39 0.89 1.28 1.08 1.17 1.12 1.26 1.00 26 27 28 29 30 31 32 33 0.60 1.24 1.08 1.31 0.99 1.23 0.92 0.75 34 35 36 37 38 39 40 41 42 0.39 1.09 1.17 0.99 0.90 0.97 1.21 0.89 43 44 45 46 47 48 49 50 0.51 1.05 1.35 1.23 0.97 1.25 1.09 1.29 52 53 54 55 56 57 58 59 60 0.67 ~ 1.25 1.00 1.26 0.92 1.21 1.09 1.31 1.03 61 62 63 64 65 66 67 68 69 0.85 1.00 1.37 1.00 0.75 '.89 1.29 1.03 0.93 Assembly Relative Power Densities at Full Power EOC 3 with Bank 5 Inserted 5-15
4l Wlf4llb ~ I. 1 0
D D
- % +! 4 "4'l 4
J 0
FIGURE 5-10 PALO VERDE U3C3 Fxy = 1.59 Box = 51 0.34 0.46 '0.62 0.79 10 0.34 0.52 0.95 0.93 1.17 0.95 12 13 14 16 17 0.51 0.77 1.09 0.97 0.97 1.40 18 20 21 22 23 24 25 0.34 0.77 0.93 1.06 1.09 1.39 1.21 26 27 28 29 30 31 '32 33 0.52 1.09 0.93 1.05 0.94 1.31 1.16 1.48 34 35 36 37. 38 39 40 41 42 0.34 0.95 0.97 1.06 0.94 0.92 1.08 1.48 43 44 45 46 47 49 50 51 0.46 0.93 1.08 1.31 1.08 1.43 1.27 1.52 52 53 54 55 56 57 58 59 60 0.62 1.17 0.97 1.39 1.16 1.48 1.27 1.48 1.15 61 62 63 64 65 66 67 68 69 0.79 0.95 1.40 1.21 1.48 1.17 1.52 1.15 0.92 Assembly Relative Power Densities at Full Power EOC3 with PLCEA's Inserted 5-16
I 0
V
,"a +
~Q ~
Cl
FIGURE 5-11 PALO VERDE U3C3 Fxy = 1.50 Box = 59 0.38 0.51 0.67 0.85 10 0.39 0.59 1.06 1.00 1.24 1.00 12 13 15 16 17 0.59 0.88 1.22 1.06 1.17 0.97 1.37 18 19 20 21 22 23 24 ,25 0.39 0.88 1.26 1.03 1.13 1.09 1.27 '1;02 26 27 28 29 30 31 32 33 0.59 1.22 1.03 1.14 0.97 1.26 0.97 0.79 34 35 36 37, 38 39 40 41 42 0.38 1.06 1.06 1.13 0.97 0.92 1.03 1.31 0.97 43 44 45 46 47 48 49 50 51 0.51 1.00 1.17 1.08 1.26 1.03 1.36 1.20 52 53 54 55 56 57 59 60 0;67 1.23 0.97 1.27 0.97 1.31 1.19 1.43 61 62 63 64 65 66 67 68 69 0.85 1.00 1.37 1.02 0.79 0.97 1.42 0.90 Assembly Relative Power Densities at Full Power EOC3 with Bank 5 and PLCEA's inserted 5-17
p
'I il.'I E
6.0 THERMAL-HYDRAULIC DESIGN 6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 3 at the rated power level of 3800 MWT have been 'performed using the,iTORC .computer-'-code"described;.=
in Reference 6-1,, the CE-1 critical heat flux correlation"described-in References 6-2 and 6-8, and the CETOP code described in Reference 6-3.
Table 6-1'ontains a list of pertinent thermal-hydraulic design parameters. The 'Modified Statistical Combination of Uncertainties (MSCU) methodology presented in Reference 6-4 was applied with Palo Verde-3 specific data using the calculational factors listed in-Table 6-1 and other uncertainty factors to define overall uncertainty penalty factors to be applied in the DNBR calculations performed by the Core Protection Calculators (CPC) and Core Operating Limit Supervisory System (COLSS) which, when used with the Cycle 3 DNBR limit of 1.24, provide assurance at the 95/95 confidence/probability level that the hot rod will not experience DNB. The 1.24 DNBR limit was calculated using the methodology of Reference 6-5 as was done for the Reference Cycle.
This Cycle 3 DNBR limit includes the following allowances:
- 1. NRC imposed 0.01 DNBR penalty for HID-1 grids as discussed in Reference 6-6.
- 2. Rod bow penalty as discussed in Section 6.2 below.
Other penalties imposed by NRC in the course of their review of the Cycle 1 Statistical Combination of Uncertainties (SCU) analysis discussed in Reference 6-5 (i.e., TORC code uncertainty and CE-1 CHF correlation cross validation uncertainty, as discussed in Reference 6-6) are included in the overall uncertainty penalty factors derived in the Cycle 3 MSCU analysis.
6-1
II
'e f~
EFFECTS OF FUEL ROD BOWING ON DNBR MARGIN'ffects of fuel rod bowing on DNBR margin have been incorporated in the safety and setpoint .analyses in the manner discussed in Reference 6-7. The, penalty used for this analysis, 1.75%MDNBR, is valid for bundle burnups. up to 30,000"MWD/MTU. This penalty is . - .
incl,uded in the 1.24 DNBR limit.
- for assemblies with burnup greater than 30 GWD/T sufficient available margin exists to offset rod 'bow penalties due to the lower radial power peaks in these higher burnup batches. Hence the rod bow penalty based upon Reference, 6-7 for 30 GWD/T is applicable for
.all assembly burnups. expected for Cycle 3.
6-2
V' L
ik Cl
je TABLE 6-1 PVNGS-3 Cycle 3 draul'ic Parameters at Full Thermal H Power Reference General Characteristics Units ~Ccl e ~Cele 3 Total Heat Output (Core only) MWg 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) 0F 565.0 565.0 Total Reactor Coolant Flow 423,300 423,300 gpss (Minimum Steady State) 10 lb/hr 155.8 155.8 Coolant;Flow Through"Core (Minimum) 10 lb/hr '151. 1 151.1 Hydraulic Diameter (Nominal Channel) ft 0.039 0.039 Average Mass Velocity 10 1 b/hr- ft 2.49 2.49 Pressure Drop Across Core (Minimum psl 14.5 14.5 steady state flow irreversible hP over entire fuel assembly)
Total Pressure Drop Across Vessel psl 51.3 51.3 (Based on nominal dimensions and minimum steady, state flow)
Core Average Heat Flux (Accounts, BTU/hr- ft 186,400* 185,900**
for fraction of heat generated in fuel rod and axial densifica-tion factor)
Total Heat Transfer Area (Accounts 67,800* 68,000**
for axial densification factor)
Film Coefficient at Average Conditions BTU/hr- ft F 6100 6100 Average Film Temperature Difference F 31 30.5 Average Linear Heat Rate of Unden- kw/ft 5.5 5.4 sified Fuel Rod (Accounts for fraction of heat generated in fuel rod) 6-3
0 II 0
TABLE 6-1 (continued)
General Characteristics Units Refer ence
~Cele 'Cele 3 Average Core Enthalpy Rise BTU/lb 85.9 85.9 Haximum Clad Surface Temperature 0F 656 656 Engineering. Heat Flux Factor 1.03+ 1.03+
Engineering Factor'on'ot Channel 1.'03+ '
'1.03+'-"-
Heat Input Rod Pitch, Bowing and Clad Diameter 1.05+ 1.05+
Factor Fuel Densification Factor (Axial) 1.002 1.002 NOTES:
- Based on,2512 poison rods.
- Based on 2352 poison rods.
+ These factors have been combined statistically with other uncertainty factors as described in, Reference 6-4 to,define overall .uncertainty penalty factors to be applied in the DNBR calculations in 'COLSS and CPC which, when used in conjunction with the appropriate DNBR limit for that cycle .provide assurance at the 95/95 confidence/probability level that the hot rod will not experience DNB.
++
Tech. Spec. minimum flow rate.
6-4
0 e
t.'A t
II
7.0 ,NON-LOCA SAFETY ANALYSIS 7.0.1 Introduction This section presents the results .of,the, Palo Verde;Nuclear
, Generating,,Station Unit 3 (PVNGS-3), Cycle. 3.:Non.-.LOCA .safety analyses at 3800 HWt.
The Design Basis Events (DBEs) considered in the safety analyses are listed in Table 7.0-1. These events are categorized into three groups: Hoderate Frequency, Infrequent, and Limiting Fault events.
For the purpose of this report," the-Moderate-Frequency and Infrequent Events will be termed Anticipated Operational
= Occurrences. The'BEs 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 3. The DBEs chosen for. analysis for each-criterion are'he limi.ting events wi,th.
respect to that criterion.
7.0.2 Methods of Anal sis The analytical methodology used for PVNGS-3 Cycle 3 is the same as the Cycle 2 (Reference, Cycle), methodology (References 7-1, 7-2 and 7-9) unless otherwise stated in the event presentations. Only methodology that has previously been reviewed and approved on the PVNGS dockets (References 7-10 and 7'-ll), the CESSAR docket (Reference 7-2), or on other dockets is used.
7-1
0 i~ h h
0 0
'Mathematical Models The mathematical models and computer codes used in the Cycle 3-Non-LOCA safety analysis are the same as .those used in the Reference Cycle analysis (References 7-1, 7-2 and 7-9). Plant response for Non-LOCA Events 'was simulated-using the CESEC III computer"code (Reference 7-3). Simulation of the flui'd. conditions, with'in.the-'hot'hannel of the reactor core and calculation of DNBR was performed using the CETOP-D computer code described in Reference 7-4.
The TORC computer code was used
'I to simulate the fluid conditions within the 'reactor core and to calculate fuel"pin DNBR for the RCP Shaft Seizure and Sheared Shaft event. The TORC code is described in References 7-6 .and 7-7.
The number of fuel pins predicted,to experience clad failure is taken as the number of pins which have a CE-1 DNBR value below 1.24.
The only exceptions are the CEA Ejection, the Shaft Seizure and Sheared Shaft events for which the statistical convolution method, described in Reference 7-8, was used. Reference 7-8 has been approved by the NRC and has been used in References 7-1, 7-2 and 7-9.
The HERMITE computer code (Reference 7-5) was used to simulate the reactor core for analyses which required more spatial detail than is provided by a,point kinetics model. Reference 7-5 has been approved by the NRC and has been used .in, References 7-1, 7-2 and .7-9.
HERMITE was also used to generate input to the CESEC point kinetics model by partially crediting space-time effects so that the CESEC calculation of core power during a reactor scram is conservative relative to HERMITE.
7-2
44ir, 'i'v r- I Ps vt,
+,* lite Ifi ~
0
,7.0.4 In ut Parameters and 'Anal sis Assum tions Table 7.0-6 summarizes the core parameters assumed in the Cycle 3
,.transient. analys'is and;compares them to the values used in-the Reference Cycle. Specific initial conditions for each event are tabulated 'in the section of the report summarizing that "event. '-Tech Spec changes are described in Section 10. The 'effects. of 'these'-..'...;
changes were considered for each DBE and were included as appropriate. For some of the DBEs presented; certain initial core parameters were assumed to be more limiting than the actual calculated Cycle 3 values. Such assumptions resulted in more
,adverse consequences. Events which have credited CPC trip protection have assumed instrument channel response times which are conservative relative to the Cycle 3 Technical'pecifications. . '.-
7.0.5
~ ~ Conclusion All~ DBEs have: been evaluated for PVNGS-3, Cycle 3 to determine whether their results are bounded by the Reference Cycle.
7-3
0 r~
f$
Y Age 1
0
Table 7.0-1 PVNGS Unit 3 Desi n Basis Events Considered in the C cle 3 Safet Anal sis 7.1 Increase in Heat Removal by the Secondary System 7.1.1 Decrease in Feedwater Temperature-7.1.2 Increase in Feedwater Flow 7.1.3 Increased Hain 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 7.2.6* Feedwater System Pipe Breaks
~.s 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 Hisoperation Events 7.4.4 CVCS Malfunction (Inadvertent Boron Dilution) 7.4.5 Startup of an Inactive Reactor Coolant System Pump 7.4.6* Control Element Assembly Ejection 7.5 Increase in Reactor Coolant System Inventory 7.5.1 CVCS Malfunction 7.5.2 Inadvertent Operation of the ECCS During Power Operation
- Categorized as Limiting Fault Events 7-4
0 U.
ik 0
Tabl'e 7;0-1,(continued)
.7. 6 . Decrease in .Reactor Coolant System Inventory 7.6. 1 Pressurizer 'Pressure Decrease Events 7.6.2* Smal.l 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 Fault Events 7-5
II Table 7.0-2 DBEs Evaluated with Res ect to Offsite Dose Criterion Section Event Results A) Anti'cipated. Operational Occurrences 7.1.4 1) Inadvertent opening of a 'Steam ';Bounded 'by Generator Safety Valve or Atmospheric Reference Cycle 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 ') 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 Presented Shaft Seizure/Sheared Shaft 7.4.6 4) Control Element Assembly Ejection Bounded by Reference Cycle 7.6.2 5) Small Primary Line Break Outside Bounded by Containment Reference Cycle 7.6.3 6) Steam Generator Tube Rupture Bounded by Reference Cycle 7-6.
tkr Table 7.0-3 DBEs Evaluated with Res ect to RCS Pressure Criterion Section Event Results A) Anticipated Operational: Occurrences 7.2.1 1) Loss of External Load Bounded by Reference Cycle 7.2.2 2) Turbine Trip Bounded by Reference Cycle 7.2.3 3) Loss of Condenser Vacuum Bounded by Reference Cycle 7.2.4 4) Loss of Normal AC Power Bounded by Reference Cycle 7.2.5 5) Loss of Normal Feedwater Bounded by
'Reference Cycle 7.4.1 6) Uncontrolled CEA Withdrawal from Bounded by Subcritical or Low Power Condition- Reference Cycle 7.4.2 7) Uncontrolled CEA Withdrawal at Power Bounded by Reference Cycle 7.5.1 8) CVCS Malfunction Bounded by .
Reference Cycle 7.5.2 9) Inadvertent Operation of the Bounded by ECCS During Power Operation Reference Cycle B) Limiting Fault Events 7.2.6 1) Feedwater System, Pipe Breaks 'Bounded by Reference Cycle 7.4.6 2) Control Element Assembly Ejection Bounded by Reference Cycle 7-7
<<5 A
0 0
Table 7.0-4 DBEs Evaluated with Res ect to Fuel Performance Section Event Results
- "A) Anticipated .Operational Occurrences 7.1.1 1) -
Decrease "in Feedwater Temperature'* - "'ounded",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 7.1.4. 4) Inadvertent Opening of a Steam Presented
- Generator Safety Valve or Atmospheric Dump Valve 7.3.1
~ ~ .5) Total Loss of Forced Reactor Bounded by Coolant Flow Reference Cycle 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 8) CEA Misoperation Events Bounded by Reference Cycle 7.6.1 9) Pressurizer Pressure Decrease Bounded by Events Reference Cycle 7.7.1 10) .Asymmetric Steam Generator .Events Bounded .by Reference Cycle B) Limiting Fault Events
- 1) Steam System Piping Failures: Bounded by Reference Cycle 7.1.5a a) Pre-Trip Power Excursions 7.1.5b b) Post-Trip Return to Power 7-8
0
+4'-
0 4,
Table 7.0-4'continued)
Section Event Results 7.3.2 2) Single Reactor Coolant Pump - Presented Shaft Seizure/Sheared Shaft 7.4.6 3) Control Element Assembly, Ejection '.Bounded by Reference Cycle
- The Base Case is bounded by Reference Cycle. Results of the Event with Loss of Offsite Power i's presented.
7-9
il 0
f
Table 7.0-5 DBEs Evaluated with Res ect 'to Shutdown Mar in Cri'terion Section Event Results
.A) Anticipated Operational'.Occurrences 7.1.4 1) Inadvertent 'Opening of a Steam Bounded, by Generator Safety Valve or Reference Cycle Atmospheric Dump Valve 7 ..4 . 4 2) CVCS Malfunction ( Inadvertent Bounded by Boron D i 1 ut i on ) Reference Cycle 7..4.5 3) Startup of an Inacti,ve Reactor Bounded by Coolant System Pump Reference. Cycle B) 'Limiting .'Fault Events
- 1) Steam System Piping Failures: Bounded by Reference Cycle 7.1.5b
~ ~ a) Post-Trip Return-to-Power 7-10
90 W 7
0 0
Table 7.0-6 PVNGS Unit 3 C cle 3
,Core Parameters In ut to Safet Anal ses
'Reference Cycle Safet Parameters Units Y 1 Total RCS Power HWt 3898 3898 (Core Thermal Power
+ Pump Heat)
Core Inlet Steady State F 560 to 570 560 to 570 Temperature (90% power and (90% power and above) above)
'550 to 572 550 to 572 (below 90% power) (below 90% power)
'Steady State psia 2000 - 2325 2000 - 2325
.RCS Pressure Minimum Guaranteed g Pill 423,320 423,320 Delivered Volumetric, Flow Rate Axial Shape Index LCO ASI Units -0.3 to +0.3 -0.3 to +0.3 Band Assumed ( >= 20% Power) ( >=20% Power)
-0.6 to +0.6 -0.6 to +0.6
( < 20% Power) ( < 20% Power)
Maximum, CEA,Insertion % Insertion 28 28 at Full Power of Lead Bank
% Insertion 25 25
.of Part-Length Haximum Initial Linear KW/f t 13.5 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 7-11
0 0
Table 7.0-6 (continued)
Reference Cycle
,Safet Parameters -Units Value Hinimum DNBR CE-1 (SAFDL) 1.24 1.24 Hacbeth (Fuel failure 1.30 1.30 limit for post-trip SLB with 'LOAC-References 7-12 and 7-13)
Initial Hoderator. 10 hp/,F Figure 7.0-1 Figure 7.0-1 Temperature Coefficient Shutdown Hargin (Value %hp -6'. 5 -6.5 Assumed in Limiting Hot Zero Power SLB) 7-12
li 4
V' P4 0
7.. 1'INCREASE IN HEAT 'REMOVAL BY THE SECONDARY SYSTEM
- 7. 1. 1 .,Decrease in Feedwater Tem erature The results are bounded by the Reference Cycle.
7.1.2 Increase in Feedwater Flow The. results are bounded by the Reference Cycle.
- 7. 1.3 Increased Main Steam Flow The results are bounded by the Reference Cycle.
- 7. 1.4 Inadvertent 0 enin of a Steam Generator Safet Valve or Atmos heric Dum Valve The amount of predicted failed fuel has increased for the Inadvertent Opening: of a Steam Generator Atmospheric Dump Valve with Loss of Offsite Power after Turbine Trip (IOSGADV+LOP) from 8% to 12%. The increase in failed fuel was the result of more adverse nuclear power distributions. 'All other reference cycle data and results related to the IOSGADV remain applicable.
7-13
f II 0
7.1.5 Steam'S stem Pi in Failures
- 7. 1.5a ,
Steam S stem Pi in Failures: Inside and Outside Containment, Pre-Tri Power Excursions The results are..bounded",by the Reference Cycle.
- 7. 1.5b, Steam S stem"Pi in Failures: Post-Tri Return to Power The results are bounded by the Reference Cycle .
7.2 DECREASE 'IN"HEAT REMOVAL BY THE SECONDARY. SYSTEM 7.2.1 Loss of External Load The results're bounded by the 'Reference Cycle.
7.2.2 Turbine Tri The results are bounded 'by the Reference Cycle.
7.2.3 'Loss of Condenser Vacuum The results are .bounded by the, Reference Cycle.
7.2.4 Loss of. Normal AC Power The results are bounded by the Reference Cycle.
7-1'4
I. 6 A 4
ik 0
7.2.5 Loss of 'Normal Feedwater The results are bounded by the Reference Cycle.
7.2.6 Feedwater S stem Pi e Breaks
,The. results, are bounded by th'e Reference Cycle.
7.3 DECREASE IN REACTOR COOLANT FLOWRATE 7.3. 1 Loss of Forced Reactor Coolant The results are bounded by the Reference Cycle.
7.3.2 Sin le Reactor Coolant 'Pum Shaft Seizure Sheared Shaft The amount of predicted failed fuel 'has increased for,,the Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft from 3.79% to 4.5%.
,As wi'th section 7. 1.4, the increase in failed fuel was the result of more adverse nuclear power distributions.. The resul:tant.
radiological consequences are a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site boundary thyroid dose of less 'than 240 Rem. This is within 10CFR100 guidelines.
7.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 7.4. 1 Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition The results are bounded by the Reference Cycle.
7-15
7.4.2 Uncontrolled CEA Withdrawal at Power
-The results are bounded by the 'Reference Cycle.
7.4.3 CEA Miso eration Event The results are bounded by the Reference Cycle.
7.4.4 CVCS MALFUNCTION INADVERTENT BORON DILUTION)
The results are bounded 'by the Reference Cycle.
7.4.5 Startu of an Inactive Reactor Coolant Pum Event The results are bounded by the Reference Cycle.
7.4.6 Control Element Assembl E ection The. results are bounded by, the Reference Cycle.
7.5 'INCREASE IN REACTOR COOLANT SYSTEM INVENTORY 7'.5.1 CVCS Malfunction The results are bounded 'by- the Reference Cycle.
7.5.2, Inadvertent 0 eration of the ECCS Durin Power 0 eration.
The results are bounded'y the Reference Cycle.
7-16
0 4
7.6 DECREASE IN REACTOR COOLANT SYSTEH INVENTORY 7.6. 1 Pressurizer Pressure Decrease Events The results. are bounded, by the .Reference Cycle..
7.6.2 Small Primar Line Pi e Break Outside Containment The results are bounded by the Reference Cycle.
7.6.3 . Steam Generator Tube Ru ture The results are, bounded by the Reference Cycle.
7.7 HISCELLANEOUS 7.7. 1 As mmett ic Steam Generator Events The resul.ts are bounded by the Reference Cycle.
7-17
FIGURE .0-1 RLLOWRBLE MTC MODES I RND 2 LL O
0.5 (OZ.O.5) (IOOZ,O.O) 0.0 I
Z LU II -0.5 LL LL -1.0 LU I
C3 LU
-1.5 ALLQWRBLE MTC CL I -2.0 CL CL LU CL
-2.5 (0%,-2.8)
LU I
-3.0 (100%,-3. )
O I
CL CL LLI
-3.5 CI 0 10 20 30 40 50 60 70 80 90 100 C3 CORE POWER LEVEL, % OF RRTED THERHRL POWER
Ei Cy 0
8.0 ECCS ANALYSIS 8.1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT 8.1.1 Introduction And Summar An ECCS performance analysis of the limiting break size was performed for PVHGS-3 Cycle 3 to demonstrate compliance with IOCFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled reactors (Reference 8-1). The analysis justifies an allowable Peak Linear Heat Generation Rate (PLHGR) of 13.5 kw/ft. The method of analysis and detailed results which support this value are presented herein.
8.1.2 Method Of Anal sis The ECCS performance analysis for PVHGS-3 Cycle 3 consisted of an evaluation of the differences between Cycle 3 and PVHGS-3 Cycle 1.
For this reason PVNGS-3, Cycle 1 shall be referred to as the Reference Cycle in Section 8. Acceptable ECCS performance was demonstrated for the Reference Cycle in Reference 8-2 and approved by the NRC in Reference 8-3. As in the Reference Cycle, the calculations performed for this evaluation used the HRC approved C-E large break ECCS performance evaluation model which is described in reference 8-4 including the use of a more conservative axial power shape. The blowdown hydraulic calculations, refill/reflood hydraulics calculations, and steam cooling heat transfer coefficients of the Reference Cycle apply to PVNGS-3 Cycle 3 since there have been no significant adverse changes to RCS or ECCS hardware characteristics, or to core and system parameters.
Therefore, only fuel rod clad temperature and oxidation calculations are required to re-evaluate ECCS performance with respect to the changes in fuel conditions introduced by Cycle 3. The NRC approved STRIKIH-II (Reference 8-5) code was used for this purpose.
8-1
C J
Ar
'.Burnup dependent-calculations were performed with STRIKIN-II to determine the limiting conditions for the ECCS performance analysis.
The fuel performance data was generated with the 'FATES-3A fuel evaluation model (References 8-6 and 8-7) with the NRC grain-size restriction (Reference 8-8),. It was demonstrated that the'urnup with the highest initial fuel stored'nergy,was.'.limiting.:-This'- r--.
'occurred at a low burnup for the hot rod.
The Unit 3 Cycle 3 analysis considered a reduction of 470 gpm in LPSI runout flow relative to the Reference Cycle. The evaluation confirmed that there is adequate safety injection flow to maintain a
'full'owncomer with the reduced flow. Therefore, this reduction in LPSI flow will not affect the results.
The acceptable performance of Unit 3 Cycle 3 has also been confirmed
,with up to 400 plugged tubes per steam generator and with a reduction in system flow rate to 155.8 X 10 ibm/hr and a reduction in core flow rate to 151.1 X 10 ibm/hr. For the Unit 3 Cycle 3 analysis nuclear flux augmentation factors were set to unity. The allowable PLHGR of 13.5 kw/ft for Cycle 3 is a reduction of 0.5 kw/ft from the Reference Cycle.
The temperature and oxidation calculations were performed for the 1.0 Double-Ended Guillotine at Pump Discharge (DEG/PD) break. This break size is the limiting break size of the Reference Cycle and, as the hydraulics are identical, is the limiting break size for Cycle 3.
8.1. 3 Res ul ts The ECCS performance analysis for PVNGS-3 Cycle 3 showed that the reference analysis results conservatively apply. The peak clad temperature, maximum local clad oxidation, and core wide oxidation values of 2091 F, 9.0% and < 0.80%, respectively, for the reference analysis, are below the corresponding 10CFR50.46 acceptance criteria of 2200'F, 17%, and 1%, respectively.
8-2
+l "a
sf'g' 4'
- 8. 1.4
~ ~ Conclusion Conformance to the,ECCS criteria is demonstrated. by the analysis results. Therefore, operation of PVNGS-3 Cycle 3 at a core power level of,3876 'HWt (.102% of 3800 MWt) and .a.,PLHGR of 13.5 .kw/ft is, in compliance with 10CFR50.46.
8.2 SHALL BREAK:L'OSS-OF-COOLANT ACCIDENT A review of Cycle 3 fuel and core data confirmed that the reported small break loss-of-coolant accident results (Reference 8-9) for PVNGS-3 Cycle 1 bounds PVNGS-3 Cycle 3. Therefore, acceptable small break LOCA ECCS performance is demonstrated at a peak linear heat generation rate of 13.5 kw/ft and a reactor power level of 3876 HWT (102% of 3800 HWT). This acceptable performance has been confirmed-
,with up to 400 plugged tubes per steam generator.
The. reduction in delivered low pressure"safety injection flow (see Reference 8-10) does not impact the small break loss-of-coolant analysis. The fuel. cladding temperature excursion is either terminated .by the, high pressure safety injection..pump flow or .by the discharge of the safety injectio'n tanks.
i 8-3
i1 II iS
'I
9.0 REACTOR PROTECTION AND MONITORING SYSTEM
9.1 INTRODUCTION
.The Core Protection Calculator System"(CPCS) is designed to provide the low .DNBR and high Local Power 'Density (LPD) trips to.-('I)"ensure"
,that the specified acceptable fuel design limits .on departure;from -.
nucleate boiling and centerline fuel melting are not exceeded during Anticipated Operational Occurrences (AOOs) and (2) assist the Engineered Safety Features System in limiting the consequences of certain postulated accidents.
The CPCS in conjunction with the remaining Reactor Protection System (RPS) must be capable- of. providing protection for certain specified'esign basis events, provided that at the initiation of these occurrences the Nuclear Steam Supply System, its subsystems,
-components and parameters are maintained within operating limits and Limiting Conditions for Operation (LCOs).
9.2 CPCS SOFTWARE MODIFICATIONS The algorithms associated with the CPC Improvement Program (References 9-1, 9-2 and 9-3) which were implemented in Cycle 2', are applicable to this cycle. The values for the Reload Data. Block constants will be evaluated for applicability consistent with the cycle design, performance and safety analyses. Any necessary change to the RDB constants wi,l:1 be installed in accordance with Reference, 9-4.
9-1
4g rS Q ~
J 4 I 0
'DD SSABLE CONSTANTS Certain CPC constants are addressable so that they'can be changed as required during operation. Addressable constants include (I) constants that are. measured during.startup (e.g.,-. shape-anneal;ing matrix, boundary point power correlation coefficients, and adjustments for planar .radial peaking factors), (2)..uncertainty factors to,account .for. processing and,:measurement uncertainties in .
DNBR and LPD calculations (BERRO through BERR4), (3) trip 'setpoints and (4) miscellaneous items (e.g., penalty factor multipliers, CEAC penalty factor time delay, pre-trip setpoints, CEAC inoperable flag, calibration constants, etc.).
Trip setpoints, uncertainty factors and other addressable constants will,be determined for this cycle consistent with the "software'nd -=
methodology established in the CPC Improvement Program and the cycle design, performance and safety analyses. As for the Reference Cycle, uncertainty. factors will be determined using the modified statistical combination of uncertainties method (Reference 9-5).
9.4 DIGITAL MONITORING SYSTEM COLSS The, Core Operating Limit Supervisory System (COLSS), described in Reference 9-6, is a monitoring system that initiates alarms if the LCO's on DNBR, peak linear heat rate, axial shape index, core power, or core azimuthal tilt are exceeded. The COLSS data base and uncertainties will be updated, as required, to reflect the reload core design.
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TECHNICAL SPECIFICATIONS This section provides a summary of the .proposed. changes to the Technical, Specifications for"PVNGS-3 Cycle 3. The changes are arranged in numerical, order.
10-1
IP Section Ti tl e 'Natur e of Chan e 3.1.1.2, Shutdown Hargin Revise Shutdown Hargin Requirements Figure 3.1-1A T greater than consistent with Cycle 3 analyses.
210'F 3.1.2.7, Boron Dilution Revise Tables 3. 1-2, 3. 1-3, and 3. 1-5 on Tables 3.1-2, Alarms monitoring frequency with 1 or 2 boron 3.1-3, and dilution alarms inoperable, due to 3.1-5 increased BOC critical boron concentrations for. Cycle 3.
3.2.4 DNBR Hargin .Revise Figures 3.2-3 and 3.2-2A to reflect Figures 3.2-2 Cycle 3 core characteristics.
and 3.2-2A 3.2.7 Axial Shape Index Change COLSS ASI limits to +0.27.
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11,. 0 STARTUP TESTING
, 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 (Reference 11-1) "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. PHYSICS TESTS 11.1.1 In'itial Criticalit Initial criticality will be achieved by one of two methods. By the first method, all CEA groups would be fully withdrawn with the exception of the lead regulating group which would be positioned at approximately mid-core. The boron concentration of the reactor coolant would then be reduced until criticality is attained. By the second method the boron concentration is adjusted to the expected critical concentration with the shutdown and Part-Length CEA groups fully withdrawn. The regulating CEA groups would be withdrawn to achieve criticality.
11.1.2 Cr i ti cal Boron Concentr ati on CBC The CBC will be determined for the unrodded configuration and for a partially rodded configuration.. The measured ,CBC values wil.l be verified to be within +1/ hk/k of the predicted values.
11-1
II
- 11. 1.3 Tem erature 'Reactivit Coefficient The isothermaltemperature coefficient (ITC) wil.l'e measured at .the Essentially All Rods Out (EARO) configuration and't a partially rodded configuration. The coolant temperature wil:1 =-be-'varied -and the resulting reactivity .change will 'be measured.. '-The measured values will be verified to be wi.thin +0.3 x '10 hk/k/ F ofthe predicted values.
This technique consists of measuring the worth of a "Reference Group" via standard boration/dilution techniques and then exchanging this group with other groups to measure their worths. All full-length CEAs will be included in the measurement. Due to the large differences in CEA group worths, two reference groups (one with high worth and one with 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.
11.2 Power Ascension Testin Following completion of the Low Power, Physics Test sequence, reactor power will be increased in accordance with normal operating procedures. The power ascension will be monitored-.through use of an off-line NSSS performance and data processing computer algorithm.
This computer code will be executed in parallel with the power
,ascension to,,monitor CPC and COLSS performance relative to the 11-2
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processed plant data against which they are normally calibrated. If
,necessary, the power ascension will be suspended while necessary data reduction and equipment calibrations are performed. The following measurements will be performed during the program.
11.2.1 Flux S mmetr Verification Core power distribution, as determined .from fixed incore detector data, will be examined prior to exceeding 30% power to verify that no detectable fuel misloadings exist. Differences between measured powers in symmetric, instrumented assemblies will be verified to be within 10% of the symmetric group average.
11.2.2 Core Power Distribution Core power distributions derived from the fixed incore neutron detectors will be, compared to predicted distributions at two power plateaus. These comparisons serve to further verify proper fuel loading and verify consistency between the as-built core and the engineering design models. Compliance with the acceptance criteria at the intermediate power plateau (between 40% and 70% power) provides reasonable assurance that the power distribution will remain wi,thin the design limits while reactor power is increased to 100%, where the second comparison will be performed.
The measured results will be compared to the predicted values in the following manner for both the intermediate and the full power analyses:
A. The root-mean-square (RHS) of the difference between the measured and predicted relative power density (axially integrated) for each of the fuel assemblies will be verified to be less than or equal to 5%.
(I 11-3
'k 4Z pP 4
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B. The RHS of the difference between the measured and predicted .core average. axial power distribution for each axial node will be verified =to be less than or equal to 5'.
C. The measured values of planar radial peaking factor (Fxy),
,integrated radial peaking factor (Fr), core average .axial peak (F ), .and .the 3-0 power, peak (Fq) wi,l.l,be ver.ified=;to be:within +10/ of their predicted values.
11.2.3 Sha e Annealin Matrix SAH and Boundar Point Power Correlation Coefficients BPPCC Verification The SAM and BPPCC values will be determined from a linear regression analysis of the measured excore detector readings and corresponding core power distribution determined from incore detector signals.
Since these values must be. representative for a rodded and unrodded core throughout the cycle, it is desirable to use as wide a range of axial shapes as is available to establish their values. The spectrum of axial shapes encountered during the power ascension has been demonstrated to be adequate for the calculation of the matrix elements. The necessary data will be compiled and analyzed through the power ascension by the off-line NSSS performance and data processing algorithm. The results of the analysis will be used to modify the appropriate CPC constants, if necessary.
11.2.4 Radial Peakin Factor RPF and CEA Shadowin Factor RSF Verification The RPF and RSF values will be determined using data collected from the fixed incore detectors and the excore detectors. Values will be determined for unrodded as well as rodded (lead regulating group and part-length group only) operating, conditions. Appropriate CPC and/or COLSS constants will be modified based upon the calculated values.
The rodded portions of this measurement may be deleted from the test program if appropriate adjustments are made to CPC and COLSS constants.
11-4
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11.2.5 Tem erature Reactivit Coefficients at Power The isothermal temperature coefficient .(ITC) .will,be .measured,.at approximately full power. The ITC will be measured by changing coolant temperature, compensating with CEA. motion, and maintaining power steady. The ITC will be verified to be within +0.3,x 10 hk/k/'F of the predicted value.
11.2.6 Critical Boron Concentration The CBC will be determined for conditions of full power, equilibrium xenon. The measured CBC will be verified to be within +50 ppm of the predicted value after adjustment for the .bias observed between measured and predicted CBC values at zero power.
11.3 PROCEDURE IF ACCEPTANCE CRITERIA ARE NOT 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.
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REFERENCES SECTION
1.0 REFERENCES
(1.-1) "Palo Verde .Nuclear Generating Station Unit No. 2, Final Safety Analysis Report," Arizona Public Service Company,,
Docket No. 50-529.
,( 1'-2). 161-01575-DBK/PGN, "Palo Verde Nuclear. Generating Station (PVNGS) Unit 3 Docket No. STN 50-530,, Submittal of the Reload Analysis Report for, Unit 3 Cycle 2," December 27, 1988.
t~
SECTION
2.0 REFERENCES
None
.SECTION '
3.0 REFERENCES
None SECTION
4.0 REFERENCES
(4-1) 161-00730-EEVB/LJM, "Final Surveil.l ance Test Results for PVNGS-1',, Cycle 1," January 8, 1988.
161'-01102-EEVB/PGN, "Fuel Surveillance Test Results for PVNGS-2 Cycle 1," Combustion Engineering Inc., June 9, 1988.
(4-3) CENPD-139-P-A, "C-E Fuel Evaluation Model," Combustion Engineering Inc., July, 1974.
(4-4) CEN-161'(B) -P-A, "Improvements to Fuel'valuation Model,"
August, 1989.
12-1
II II
(4-5) NOT USED (4-6) 161-00453-JGH/SGB, ",Fuel Assembly Guide Tube =Wear- 'Program for PVNGS Unit 3.," 'August 20, 1987.
(4-7) "Palo Verde Nuclear Generating Station Unit .No. 1, Final Safety Analysis Report,." Arizona Publ.ic Service Company, Docket No.. 50-528, Section 4.2.4.
(4-8) CESSAR SSER2, Section 4.2.5, "Guide Tube Wear Surveillance".
12.5 SECTION 5.0 'REFERENCES (5-1) CENPD-153-P, Rev. 1-P-A, "INCA/CECOR Power Peaking Uncertainty," May, 1980.
(5-2) CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design," April, 1983.
12.6 . ,SECTION
6.0 REFERENCES
(6-1) CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core", April, 1986.
CENPD-162-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution," September, 1976.
(6-3) 161-01867-DBK/JRP, "Generic Applicability of the CETOP-D Code for PWR Core T-H Analysis", April 26, 1989.
(6 4) CEN-356(V) -P-A, Rev. 01-P-A, "Modified Stati stical Combination of Uncertainties", May, 1988.
12-2
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'U gl ~t t,C I,
Enclosure 1-P to LD-82-054, ."Statistical Combination of System Parameter Uncertainties in Thermal Margin .Analyses for System 80", submi.tted by letter from A.= E. Scherer (C-E)'o D. G. Eisenhut (NRC), May 14, 1982.
CESSAR SSER 2 Section 4.4.6, Statistical Combination of Uncertainties.
CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983.
CENPD-207-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-uniform Axial Power Distribution;" December, 1984.
SECTION
7.0 REFERENCES
(7-1) "Palo Verde Nuclear Generating Station Unit No. 3, Final Safety Analysis Report," Arizona Public Service Company, Docket No. 50-530.
"CESSAR, Combustion Engineering, Standard Safety Analysis Report," Docket No. 50-470.
"CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981, Enclosure 1-P to LD-82-001, January 6, 1982.
(7-4) 161-01867-DBK/JRP, "Generic Applicability of the CETOP-D Code for PWR Core T-H Analysis," Apri'1 26, -1989.
(7-5) CENPD-188-A, "HERMITE Space-Time Kinetics," July, 1975.
(7-6) CENPD-161.-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April, 1986.
12-3
0
\
jt1 0
(7-7) CENPD-206-P-A, "TORC Code Verification and Simplified Modeling Methods," June 1981.
(7-8) 'ENPD-183-A, "Loss of Flow - C-E Methods for Loss of Flow
'Analysis," June 1984.
(7-9) 161-01575-DBK/PGN, '"Palo"Verde Nuclear Generating 'Station (PVNGS) Unit 3 Docket .No. STN 50-530, Submittal of the Reload Analysis .Report for Unit 3 Cycle 2," December 27, 1988.
(7-10) USNRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 24 to Facility Operating, License No. NPF-41, Arizona Public Service Company, et. al. Palo Verde Nuclear Generating Station, Unit No. 1 Docket No. STN50-528," October 21, 1987.
(7-11) USNRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related-to Amendment No. 19 to Facili.ty Operating License No. NPF-51, Arizona Public Service Company, et. al. Palo Verde Nuclear Generating Station, Unit No. 2 Docket No. STN50-529," May 5, 1988.
(7-12) 'R. V. MacBeth, "An Appraisal of Forced Convection Burn-out Data", Proc. Instn. Mech. En rs., Vol 180, Pt. 3C, pp 37-50, 1965-66.
(7-13) D. H. Lee, "An Experimental Investigation of Forced Convection Burn-out in High Pressure Water - Part IV, Large Diameter Tubes at About 1600 psia", A.E.E.W.,
Report R479, 1986.
SECTION 8. 0 REF ERENC ES- ECCS ANALYSIS (8-1) "Acceptance Criteria for Emergency Core Cooling Systems for, Light Water Cooled Nuclear Power Reactors," Federal Register, Vol. 39, No. 3, Friday, January 4, 1974.
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(8-2) ANPP-33584-EEVB/KLM, "Limiting Large Break, LOCA Analysis
.Results - Chapter 15 Reanalyses", September 27, 1985.
ANPP-33650-EEVB/KLM, "Large Break LOCA Evaluation Model-Reanalysis Results", October 3, 1985.
(8-3) PVNGS Safety Evaluation Report, NUREG - 0857, Section 6.3, Supplement 9, December 1985, (8-4) CENPD-132-P, "Calculational Methods for the C-E Large Break LOCA Evaluation Model", August 1974.
CENPD-132-P, Supplement 1, "Calculational Methods for the C-E Large Break LOCA Evaluation Model", February 1975.
CENPD-132-P, Supplement 2-P', "Calculational Methods for the C-E Large Break LOCA Evaluation Model", July 1975.
Letter 0. D. Parr (NRC) to F. M. Stern (C-E), dated June 13,, 1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation, Model).
Letter 0. D. Parr (NRC) to A. E. Scherer (C-E) dated December 9, 1975 (NRC Staff Review of 'the Proposed Combustion Engineering ECCS Evaluation Model Changes).
(8-5) CENPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1974.
CENPD-135, Supplement 2P, "STRIKIN-II, A.Cyl indrical 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.
12-5
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(8-6) CENPD-139'-P-A, "C-'E Fuel'valuation Model.", July, 1974.
(8-,7) CEN-161(B)-P, " Improvements to Fuel Evaluation Model",
July, 1981.
Letter from R. A. Clark (NRC) to A. 'E. Lundvall, Jr.,
(BGEE), "Safety Evaluation of BEN-161 (FATES 3),"
March 31, 1983.
(8-9) ANPP-33609-EEVB/KLH, dated September 30,, 1985, ".Limiting Small Break LOCA Analysis - Additional Information".
'(8-10) 161-00890-.EEVB/BJA, dated March 16, 1988, '"Proposed Technical Specifications Change - LPSI" Flow:Requirements".
161-01155-EEVB/BJA, dated July 6, 1988,, "LPSI Flow
'Requirements".
Letter from the NRC, dated, October 1'7, 1988, "Amendment to PVNGS-1, 2,E 3 Operating License for LPSI Flow".
SECTION
9.0 REFERENCES
(9-1) CEN-304-P, Rev. 01-P, "Functional Design Requirement for a Control Element Assembly Calculator,," Hay, -1986.
(9-2) CEN-305-P, Rev. 01-P, "Functional Design Requirement for a Core Protection Calculator," Hay, 1986.
(9-3) CEN-330-P-A, "CPC/CEAC Software Modifications for the CPC Improvement Program Reload Data Block," October, 1987.
(9-4) CEN-323-P-A, "Reload Data Block Constant Installation Guidelines," September, 1986.
12-6
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