ML20065P016
| ML20065P016 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/15/1982 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20065N996 | List: |
| References | |
| NUDOCS 8210220385 | |
| Download: ML20065P016 (152) | |
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Table 3-1 EO Calvert Cliffs Unit 2 Cycle 5 o Core Loading 125 gg
- s a Batch Average Poison Initial (2)
Total fiumber Total Initial Burnap MWD /T Rods Poison of Number
's Assembly Number of Enrichnot l?rra = Per Loading Poiron of Fuel
. Designation Assemblies wt% U-235 17,LCJ) Assembly wt% B,C Auds Rods D III 13 3.03 20,80" 0 0 0 2,288 F 40 3.65 13,100 0 0 0 7,040 F/ 88 3.03 19,700 8 3.03 704 14,784 G 48 4.00 0 0 0 0 8,448 G/ 28 3.55 0 8 3.03 224 4,704 TOTALS 217 928 37,264 II) Discharged from Calvert Cliffs Unit 2 Cycle 3.
10 (2) Shim B10 concentration equals .02685 gms 8 / inch. .I O, *e
1 2 G G 3 4 5 6 7 G G G FI F/ 8 9 10 11 12 13 G F- Fl F/ F D 14 15 16 17 18 19 20 G GI FI GI F/ GI F/ 21 22 23 24 25 26 27 28 G F F/ FI F/ F FI F 29 30 31 32 33 34 35 36 G FI GI F/ F FI GI D 37 38 39 40 41 42 43 44 45 G Fl Fl F F/ F Fl D G 46 47 48 49 50 51 52 53 54 Fl F GI F/ GI FI FI F G 55 56 57 58 59 W 61 62 FI D FI F D D F D
^ R" S
GAS E E T IC CO. CALVERT CLIFFS UNIT 2 C'YCLE 5 Nuc r ow Plant l
UNSHIMMED ASSEMBLY l l
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4 e 1 Fl-8 POISON ROD ASSEMBLY X X
-X X X X X, X
[ FUEL R0D LOCATION O POIS0N R'0D LOCATION
^ R" S
GAS EET cco. Cf.LVERT CLIFFS UNIT 2 CYCLE 5 Nuc ow "ha n, ASSEMBLY FUEL AND OTHER RCD LOCATIONS 3-2
,. I
G/ -8 POISCN R0D ASSEMBLY X X X X X X X T~ X _ FUEL R0D LOCATION 3 P01'.0d R0D LOCATION
^ R S **
GAS ELE T CCO. CALVERT CLIFFS UNIT 2 CYCLE 5
' cnns 3-3 gjj'[,, p ,,,
ASSEMBLY FUEL AND OTHER R0D LOCATIONS l
INITIAL ENRICHMENT WIO U-235 1 2 BOC5 BURNUP (MWD /f), E0C4 = 17,300 MWDIT 4.00 4.00 0 0 3 4 5 6 7 4.00 4.00 4.00 3.03 3.03 0 0 0 20,600 20,300 8 9 10 11 12 13 4.00 3.65 3.03 3.03 3.65 3.03 0 11,500 20,700 20,700 15,400 22,400 14 15 16 17 18 19 20 4.00 3.55 3.03 3.55 3.03 3.55 3.03 0 0 18,000 0 20,500 0 21,000 21 22 23 24 25 26 27 28 . 4.00 3.65 3.03 3.03 3.03 3.65 3.03 3.65 0 11,500 18,000 16,400 18,800 12,900 20,600 14,600 29 30 31 32 33 34 35 36 4.00 3.03 3.55 3.03 3.65 3.03 3.55 3.03 0 20,700 0 18,400 11,000 20,200 0 19,400 37 38 39 40 41 42 43 44 l
- ' 3.03 3.65 3.03 3.03 4.00 3.03 3.03 3.65 l 45 0 20,700 20,600 12,900 20,200 14,600 19,400 20,100 46 47 48 49 50 51 52 63 l~g'S
'O 3.03 3.65 3.55 3.03 3.55 3.03 3.03 3.65?-
54 20,600 15,400 0 20,700 0 19,100 16,300 11,100 0 55 56 57 58 59 60 61 6? 3.03 3.03 3.03 3.65 3.03 3.03 3.65 3.03
} 22,300 20,300 22,400 21,000 14,600 19,400 20,100 l11,100 R9ure GAS ELE T IC CO*
CALVERT CLIFFS UNIT 2 CYCLE 5 carvert cliffs ASSEMBLY AVERAGE BURNUP AT BOC 3-4 Nuclear Power Plant AND INITIAL ENRICHMENT DISTRIBUTION l L_
1 2 10,200 12,500 l 3 4 5 6 7 10,500 14,000 15,800 32,500 31,900 8 9 10 11 12 13 12,000 25,800 33,700 33,800 29,900 34,100 14 15 16 17 18 19 20 12,000 16,700 32,100 17,900 34 400 18,200 34,700 21 22 23 24 25 26 27 28 10,500 25,800 32,100 30,500 33,000 29,000 34,800 30,300 29 30 31 32 33 34 35 36 14,000 33,800 18,000 32,800 27,300 34,200 18,000 32,700
- 37 38 39 40 41 42 43 44 45 15,800 33,800 34,400 29,000 34,300 29,700 32,600 32,400 10,200 46 47 48 49 50 51 52 53 54 32,500 29,900 18,100 34,800 18,000 32,400 30,000 26,500 12,500 55 56 57 58 59 60 61 62 31,900 34,100 34,700 30,300 32,700 32,400 26,500 35,200 l
^ Figure GAS ELE T IC CO. CALVERT CLIFFS UNIT 2 CYCLE 5 ASSEMBLY AVERAGE BURNUP AT E0C (MWDIT) 3-5 Nuc r Plant l
1 1 - _, _ _ l
4.0 FUEL SYSTEM DESIGN 4.1 Mechanical Design The mechanical des!gn for the standard Batch G reload fuel is identical to that of the standard Batch H fuel described in the reference cycle submittal (Calvert Cliffs Unit 1 Cycle 6, Reference 1). It is also identical to that of the standard Batch F fuel used in Calvert Cliffs Unit 2 Cycle 4 (Reference 3) with the exception of a .200 inch reduction in the overall length of the fuel rods. The length reduction will provide additional clearance for fuel rod length increase during the lifetime of the Batch G fuel. The mechanical designs of the Batch D, and F fuel assemblies were described in References 7 and 3, respectively. C-E has performed analytical predictions of cladding creep collapse time for all Calvert Cliffs Unit 2 fuel batches that will be irradiated in Cycle 5 and has concluded that the collapse resistance of all standard fuel rods is sufficient to preclude collapse during their design lifetime. This lifetime will not be exceeded by the Cycle 5 duration (Table 4-1). These analyses utilized the CEPAN computer code (Reference
- 8) and the analysis procedures described in Reference 9 The analysis procedures described in Reference 9 were approved in Reference 10.
TABLE 4-1 Batch Minimum EOC 5 Collapse Time Exposure D > 35,000 EFPH 26,000 EFPH F > 35,000 23,200 G > 27,500 10,400 All batches of fuel were also reviewed for dimensional changes using the SIGREEP model described in Reference 11. All clearances were found to be adequate during Cycle 5 The SIGREEP model described in Reference 11 was approved in Reference 10. The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch G fuel are identical to those of the Batch D and F fuels to be included in Cycle 5 Thus, the chemical or metallurgical performance of the Batch G fuel will remain unchanged from the performance of the Cycle 4 fuel. 4.2 Hardware fodifications to Mitigate Guide Ibbe Wear All standard fuel assemblies which will be placed in CEA locations in Cycle 5 will have stainless steel sleeves installed in the guide tubes to prevent guide tube wear. A detailed discu.ssion of the design of the sleeves and their effect on reactor operation is contained in Reference 12. S 5
4 3 Thermal Design The thermal performance of composite fuel pins that envelopes the :-ious fuel assemblies present in Cycle 5 (fuel batches D, F and G) has been evaluated using the FATES 3 version of the fuel evaluation model (References 13 and 14). The analysis was performed with a power history that modeled 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 an11yzed is in excess of that expected at the end of Cycle 5. The FATES 3 power-to-centerline melt limit was determined for Cycle 5 by taking some credit for the decrease in power peaking which is characteristic of highly burned fuel. Since a gradual decrease in the calculated power-to-melt (due to a decrease in the fuel melt temperature) alic accompanies burnup, the most limiting power-to-centerline melt has been found to occur within an intermediate burnup range. Using censervative estimates of the burnup point at which the power peaking begins to decrease and the rate at which it decreases for Cycle 5, the most limiting power-to-centerline melt has been determined to be in excess of 22 kw/ft and to occur at a rod average burnup of approximately 33,000 E'D/MTU. S d
5.0 . NUCLEAR DESIGN Cycle 5 of Unit 2 will M the second fuel cycle for either Calvert Cliffs Unit for which the average discharge exposure will be as high as 33,700 MWD /T. All analyses address fuel exposure explicitly and, r.s in the reference cycle, the small increase in average discharge exposure over that of the previous cycles has not yielded, of itself, significant changes in core parameters. Furthermore, those assemblies with burnups in excess of 24,000 MWD /T at EOC5 contain maximum 1-pin peaks which are substantially below the maximum 1-pin peak in the core (See Section 6.2). 5.1 Physics Characteristics 5.1.1 Fuel Management The Cycle 5 fuel management employs a mixed central region as described in Section 3, Figure 3-1. The fresh Batch G fuel is comprised of two sets of i sssemblies, each having a unique enrichment in order to minimize radial power peaking. There are 48 assemblies with an enrichment of 4.00 wt% U-235 and 28 assemblies with an enrichment of 3 55 wt% U-235 and 8 poison shims per assembly. With this loading, the Cycle 5 burnup capacity for full power operation is expected to be between 13,100 MWD /T and 13,700 MWD /T, depending on the final Cycle 4 termination point. The Cycle 5 core characteristics have been examined for Cycle 4 terminations between 16,300 and 17,300 MWD /T and limiting values established for the safety analyses. The loading pattern (see Section 3) is applicable to any. Cycle 4 termination point between the stated extremes. 1 Physics characteristics including reactivity coefficients for_ Cycle 5 are
- listed in Table 5-1 along with the corresponding values from the reference f
cycle. Please note that the values of parameters actually employed in safety analyses are different from those displayed in Table 5-1 and are typically chosen to conservatively bound predicted- values with accommodation for appropriate uncertainties and allowances. Table 5-2 presents a stamary of CEA shutdown worths and reactivity allowances for the end of Cycle 5 zero power steam line break accident and a comparison to reference cycle data. He EOC zero power steam line break was selected since -it is the most limiting zero power. steam line break accident and, thus, provides the basis for establishing the - Technical Specification required shutdown margin. De power dependent insertion limit (PDIL) for bank 2 insertion has changed relative to Cycle 4. The new PDIL curve, which is the same as the, curve for the reference cycle, is -discussed in Section 9 Table 5-3 shows the reactivity worths of various CEA groups calculated at fbil power conditions for Cycle 5 and the reference cycle. 5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC5, MOC5 'and EOC5, respectively, that are characteristic of the high burnup end of the . Cycle 4 shutdown window. These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80 percent of the fuel l E l-_.,-.- .-- . . . _ , - - . _ - - . . ~
height. He high burnup end of the Cycle 4 shutdown window tends to increase the power peaking in this axial central region of the core for Cycle 5. De planar radial power distributions for the above region with CEA Group 5 fully inserted at beginning and end of Cycle 5 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 4 shutdown window. The radial power distributions described in this section are calculated data without uncertainties or other allowances. However, the single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice. For both DNB and kw/ft safety and setpoint analyses in either rodded or unrodded configurations, the power peaking values actually used are higher than those expected to occur at any time during Cycle 5 Rese conservative values, which are used in Section 7 of this document, establish the allowable limits for power peaking to be observed during operation. The range of allowable axial peaking is defined by the limiting conditions for operation covering axial shape index (ASI). Within these ASI limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum three-dimensional or total peaking factor anticipated in Cycle 5 during normal base load, all rods out operation at full power is 1.88, not including uncertainty allowances and augmentation factors. 5.1 3 Safety Related Data The Cycle 5 safety related data for this section are identical to the safety related data used in the Reference Cycle, as presented in Section 5.1 3 of Reference 1. 52 Analytical Input to In-Core Measurements In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the manner described in Reference 15, which is the same method used for the reference cycle. 53 Nuclear Design Methodology Analyses have been performed in the same manner and with the same methodologies used for the reference cycle analyses. 5.4 Uncertainties in Measured Power Distributions The power distribution measurement uncertainties contained in Reference 15 which are applied to Cycle 5 are: Total 3-D peaking factor (Fq) uncertainty = 6.2 percent where Fq = Fxy X F2 , local power density Integrated radial peaking factor (Fr) uncertainty = 6.0 percent These values are to be used for monitoring power distribution parameters during operation.
Table 5-1 Calvert Cliffs Unit 2 Cycle 5 Nominal Physics Characteristics Units Reference Cycle Cycle 5 Dissolved Boron Dissolved Boron content for Criticality, CEAs Withdrawn Hot Full Power, Equilibriun Xenon, BOC PPM 1025 1032 Boron Worth Hot Full Power BOC PPM /%ap 106 105 Hot Full Power EOC PPM /$ap 85 85 Reactivity Coefficients (CEAs Withdrawn) Moderator Temperature Coefficients, Hot Full Power, Equilibriun Xenon Beginning of Cycle 10 4 aoMF -0.2 -0.1 End of Cycle 10-Nap /0F -2.1 -2.1 Doppler Coefficient Hot Zero Power BOC 10-5aoMF -1.48 -1.48 Hot Full Power BOC 10-5aoMF -1.24 -1.27 Hot Full Power EOC 10-53ppy _1,47 _1,47 Total Delayed Neutron Fraction,..seff BOC .00609+ .00609 EOC .00522+ .00521 Neutron Generation Time, g BOC 10-6 sec 23 8 24.0 ECC 10-Osec 30.2 30.5
+These values are corrections to those reported in Reference 1 for Unit 1 Cycle 6. ** Unit 1 Cycle 6 5
I Table 5-2 Calvert Cliffs Unit 2 Cycle 5 Limiting Values of Reactivity Worths and Allowances for Hot Zero Power Steam Line Break, %ap End-of-Cycle (EOC) Reference Cycle Cycle 5
- 1. Worth of all CEA's Inserted 10.2 94
- 2. Stuck CEA Allowance 2.6 1.8 '
3 Worth of all CEA's Less Worth of CEA Stuck Out** 7.6 7.6
- 4. Zero Power Dependent Insertion Limit CEA Bite 17 1.8 5 Calculated Scram Worth 59 5.8
- 6. Physics Uncertainty (10% of Item 5) 0.6 0.6 7 Net Ava'lable Scram Worth (Item 5 minus Item 6) 53 5.2
- 8. Technical Specification Shutdown Margin 53 5.2 9 Margin in Excess of Technical Specification Shutdown Margin 0.0 0.0
- Unit 1 Cycle 6
** Stuck CEA is one which yields worst results for HZP SLB, i.e., worst combination of scram worth and reactivity insertion with cooldown.
5
Table 5-3 Calvert Cliffs Unit 2 Cycle 5 Reactivity Worth of CEA Regulating Groups at Hot Full Power, 5 ap Beginning of Cycle End of Cycle Regulating Reference. Reference CEA's Cycle Cycle 5 Cycle Cycle 5 Group 5 0.48 0.49 0.65 0.63 Group 4 0 31 0.27 0 33 0 36 Group 3 0.84 0 91 1.04 1.16 Notes Values shown assume sequential group insertion Unit 1 Cycle 6 3
0.77 0.93 0.78 1.08 1.23 0.86 0.82 X .. 0.89 1.08 0. M 0.93 -1.03 0.79 0.89 1.21 1.01 1.26 0.95 1.25 0.92 0.78 1.08 1.01 1.00 0.99 1.15 0. % 1.10 1.08 0.M 1.26 1.01 1.19 0. M 1.20 0.87 1.23 0.93 0.95 1.16 0.95 1.05 0.87 0.80 0.76 0.86 1.03 1.25 0.95 1.20 0.88 0.93 1.09 0.93 0.82 0.79 0.92 1.10 0.87 0.80 1.09 0.89 NOTE: X MAXIMUM 1-PIN PEAK - 1.57
^ CALVERT CLIFFS UNIT 2 CYCLE 5 Reure GAS EET cco' ASSEMBLY RELATIVE POWER DENSITY AT BOC, calvert cliff, 51 Nuclear Power Plant EQUILIBRIUM XENON l
h 0.74 0.90 0.75 1.00 1.13 0.84 0.82 l
~u._
0.87 1.03 0.91 0.91 1.03 0.83 O.87 1.21 1.00 1.29 0.97 1.31 0. % X 1.03 1.00 1.01 0.99 1.15 0.99 1.13 0.75 0.91 1.29 1.00 1.16 0.97 1.29 0.93 1.00 0.91 0.97 1.15 0.97 1.08 0.92 0.86 l 1.13 i 0.74 1.03 1.31 0.99 1.29 0.93 0.97 1.11 0.84 0.89 0.96 1.13 0.93 0.86 1.11 0.91 i 0.82 0.83 l l NOTE: X MAXIMUM 1-PIN PEAK =. L47 l Rsure CALVERT CLIFFS UNIT 2 CYCII 5 GAS E$#E T IC CO* ASSEMBLY RELATIVE POWER DENSITY AT ? GWDIT coiver+ cliff, 5-2 Nuclear Power Plant EQUILIBRIUM XENON l l E . . . _ . . _ _ . . _ _ _ . _ _ _ - _ .. - _ _ ___ __
0.76 0.91 - 0.77 0.99 1.11 0.85 0.34 0.90 1.03 0.92 0.92 1.03 0.85 0.90 1.26 1.00 1.32 0.97 1.34 0.98 X 0.77 1.03 1.00 0.99 0.97 1.11 0.99 1.11 0.99 0.92 1.32 0. % 1.10 0.95 1.31 0.94 1.11 0.92 0.97 1.11 0.95 1.04 0.91 0.86 0.76 0.85 1.C3 1.34 0.99 1.31 0.92 0.93 1.05 0.91 0.84 0.85 0.98 1.11 0.94 0.86 1.05 0.87 NOTE: X MAXIMUM 1-PIN PEAK = 1,51 CALVERT CLIFFS UNIT 2 CYCLE 5 Rsure GAS E E 7 IC CO* coivert Cliff, ASSEMBLY RELATIVE POWER DENSITY AT E0C' 5-3 Nuclear Power Plant EQUILIBRIUM XENON l
l . . 0.81 0.99 0.80 1.14 1.30 0.88 0.80 X 0.74 1.01 0.97 0.97 1.00
// /
0.74 0.91 1.28 1.00 1.28 0.92
///b 1.01 0.91 0.97 1.04 1.21 1.02 1.13 0.80 0.97 1.28 1.05 1.25 1.02 1.28 0.92 1.14 0.97 1.00 1.22 1.03 1.09 0.88 0.79 1.30 0.81 1.00 1.28 1.02 1.28 0.88 0.82 0.87 0.88 0.99 fl f/
0.80 f0.63d
/////
0.92 1.13 0.92 0.79 0.87 0.48 /
////
NOTE: X MAXIMUM 1-PIN PEAK- 1.65
/ 0 TI NS
^ CALVERT CLIFFS UNIT 2 CYCLE $ .
Rsure GAS ELE T CCO' calvert clifr, ASSEMBLY RELATIVE POWER DENSITY WITH B ANK 5 5-4 INSERTED, HFP, BOC Nuclear Power Plant -- d
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\
0.79 0.94 0.74 1.01 1.15 0.'85 0.80 O.71 0.94
~
0.92
~
0.95 l .'0.0 fll 0d
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0.71 0.89 1.34 1.02 1.38 0.96 7
////
~
1.02 1.'20 1.07 1.19
~
0.74 0.94 0.89 0.97
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1.01 0.92 1.34 1.03 1.19' 1.04 1.43 1.03' X 1.15 0.94 -1. 02. 1.20 1.04 1.14 0.97- 0.90 0.79 ,
- 0.85 0.99 1.38 1.07 1,44 0.97 0.89 0.91 0'94
'/
// 0.90 0.91 .48 0.80
//
.4 0.96 1.19 1.03 I//b NOTE X MAXIMUM 1-PIN PEAK- 1.64 CEA BANK 5
/ LOCATIONS
^ " CALVERT CLIFFS UNIT 2 CYCLE 5 Figure GAS ELE T !C CO~
coivert clim ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 5 Nuclect Power Plant INSERTED, HFP, E0C ,
l 6.0 THERMAL-HYDRAULIC DESIGN 6.1 DNBR Analysis Steady state DNBR analyses of Cycle 5 at the rated power level of 2700 MWt have been performed using the TORC computer code described in Reference 1, the CE-1 critical heat flux correlation described in Reference 2, and the simplified modeling methods described in Reference 3 A variant of TORC called CETOP, optinized for simplified modeling applications, was used in this cycle to devalop the " design thermal margin model" described generically in Reference 3 Details of CETOP are discussed in Reference 4. CETOP was approved for use on Calvert Cliffs Units-in Reference 5 CETOP is used only because it reduces computer costs significantly; no margin gain is expected or taken credit for. > Table 6-1 contains a list of pertinent thermal-hydraulic design parameters used for both safety analyses and for generating reactor protective system setpoint information. The calculational factors -(engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in Table 6-1 have been combined statistically with other uncertainty factors at the 95/95 confidence / probability - level to define a new design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6 and approved by the NRC in Reference 5. The applicability of this minimum DNBR limit has been verified for Cycle 5 Investigations have been made to ascertain the effect of the CEA guide tube wear problem and the sleeving repair on DNBR margins. The findings were reported to the NRC in Reference 7 which concluded that the wear problem j and the sleeving repair do not adversely affect DNBR margin. 6.2 Effects of Fuel Rod Bowing on DNBR Margin The fJel rod bowing effects on DNB margin for Calvert Cliffs Unit 2 Cycle ! 5 have been evaluated according to the guidelines set forth in Reference 8. A total of 141 fuel assemblies will exceed . the NRC specified DNB penalty threshold burnup of 24 GWD/T (Reference 8) during Cycle 5, the maximum assembly burnup reaching 35 2 GWD/T by the end of cycle. For those assemblies which will experience a burnup of betwean 24 and 35 2 GWD/T at any time during Cycle 5, the minimum best estimate margin available relative to more limiting peaking values present in other assemblies is greater than 115. The DNB rod bow penalty for this burnup range, as determined via an interpolation of the data contained in Reference 8, varies from 0 to 3 8%. In sumary, the magnitude of the margin available is considerably in excess of the corresponding DNB rod bow penalty and, consequently, no power penalty for fuel rod bowing is required in Cycle 5. l a l
Table 6-1 Calvert Cliffs Unit 2 Thermal-Hydraulic Parameters at Full Power Reference + General Characteristics Unit Unit 1. Cycle 6** cvele 5** Total Heat Output (core only) MgT 2700 2700 10 Btu /hr 9215 9215
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Fraction of Heat Generated in .975 .975 Fuel Rod . Primary System Pressure psia 2250 2250 (Nominal) Inlet Temperature *F 548 543 Total Reactor Coolant Flow gpg 381,600 381,600 (steady state) 10" lb/hr 1 43.8 1 43.8 Coolant Fiow Through Core 106 lb/hr 1 38.5 1 38.5 Hydraulic Diameter ft 0.044 0.044 (nominal channel) Average Mass Velocity 106 lb/hr-ft 2 2.59 2.59 Pressure Drop Across Core psi 11.1 11.1 (steady state flow irreversible ao over entire fuel assembly) Total Pressure Drop Across Vessel psi 34.7' 34.7 (based on steady state flow and nominal dimensions ) Core Average Heat Flux (Accounts Btu /hr-ft 2 184,266*** 185,532**** for above fraction of heat generated in fuel rod and . axial densification factor) 2 48,748*** 48,415 ** ** Total Heat Trar sfer Area (Accounts ft for axial densification factor) Film Coefficient at Average Btu /hr-ft 2 *F 5930 5930 Conditions
*F 31 Average Film Temperature Difference 31 5
l Table 6-1 (continued)
?
General Characteristics Unit Refe ence Cycle 5** Unit 1, Cycle 6** Average Linear Heat Rate of, kw/ft 6.16*** 6.20**** Undensified Fuel Rod (accounts for above fraction of heat getterated in fuel rod) Ave: age Core Enthalpy Rise Stu/lb 66.5 66.5 Maximum Clad Surface Temperature *F 6 57 6 57 Calculational Factors Reference - Unit 1, Cycle 6 Cycle 5 Engineering Heat Flux Factor 1.03* 1.03* Engineering Factor on Hot Channel 1.02* 1.02* Heat Input Rod Pitch and Clad Diameter Factor 1 .06 5* 1. 06 5* Fuel Densification Factor (axial) 1.01 1.01 , NOTES
- These factors have been combined statistically w,ith other uncertainty factors
- at 95/95 confidence / probability level (Reference 6-) to define a new design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6 and approved by the NRC in Reference 5. This limit has been verified to be applicable to Cycle 5.
** Due to the statistical combination of uncertainties described in References 6 9, and 10, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.
*** Based on Unit 1, Cycle 6 specific value of 672 shims.
**** Based on Unit 2, Cycle 5 specific value of 928 shims.
+ Reference cycle (Unit 1, Cycle 6) analysis is contained in Reference 11.
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l
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- 1
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70 Transient Analysis l
- This section presents the results of the Baltimore Gas & Electric Calvert Cliffs Unit 2, Cycle 5 Non-LOCA safety analysis at 2700 MWt. ,
J 1he Design Bases Events (DBEs) considered in the safety analysis are listed in Table 7-1. These events were categorized in the following groups:
- 1. Anticipated Operational Occurrences (A00s) for which the intervention of the Reactor Protection System (RPS) is necessary to prevent exceeding acceptable limits.
- 2. A00s for which the intervention of the RPS trips and/or initial steady
- state thermal margin, maintained by Limiting Conditions for Operation i
(LCO), are necessary to prevent exceeding acceptable limits. 3 Postulated Accidents
- 4. Postulated Occurrences The Design Basis Events (DBEs) considered in the Unit 2, Cycle 5 safety analyses are listed in Table 7-1. Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to melt fuel design 4 limits are presented in Table 7-2.
As indicated in Table 7-1, no reanalysis was performed for the DBEs for which key transient input parameters are within the bounds (conservative with respect to) of the reference cycle values (Unit 1, Cycle 6, Reference 1). For these DBEs the results and conclusions quoted in the reference cycle analysis are valid for Unit 2, Cycle 5 A re-evaluation of a number of DBEs, as indicated in Table 7-1, was performed - to determine the impact of lowering the Technical Specification minimum indicated pressure from 2225 psia to 2200 psia. For these DBEs, DNBR is the primary criterion; thus, they were re-evaluated to ensure that the transient minimum DNBR does not exceed the design limit of 1.23 The evaluation of these DBEs showed that sufficient initial steady state thermal margin is maintained by the Technical Specification DNB LCOs to assure that the transient minimum DNBR is equal to, or greater, than the design limit of 1.23 Therefore, the l conclusions reached in the reference cycle analysis remain valid, t l A re-evaluation of all DBEs was also performed to determine the impact of
- extended burnup. The evaluation showed that effects of extended burnup are
, no more severe than those reported in Reference 1. Thus, the conclusions l reached about extended burnup effects in Reference 1 are valid for Unit 2, Cycle 5 For the events reanalyzed, Table 7-3 shows the reason for the reanalysis, the acceptance criterion to be used in judging the results and a surunary of che results obtained. Detailed presentations of the results of the reanalysis are provided in the appropriate sections. f S I
- . _ - . - - . 4 -- _ _ _ . .. m . . _- : , - - _ -- _ . ,,
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TABLE 7-1 CALVERT CLIFFS UNIT 2, CYCLE 5 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS Analysis Status 71 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits: 7 1.1 Boron Dilution Not Reanalyzed 7.1.2 Startup of an Inactive Reactor Coolant Not Reanalyzed l Pump 7.1.3 Loss of Load Reanalyzed 7.1.4 Excess Load Reanalyzed 7.1.5 Loss of Feedwater Flow Not Reanalyzed 7.1.6 Excess Heat Removal due to Feedwater Not Reanalyzed Malfunction 7.1.7 Reactor Coolant System Depressurization Not Reanalyzed 7 1.8 Excessive Charging Event Analyzed 72 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial steady state thermal margin, maintaird by the LCOs, are necessary to prevent exceeding the acceptable limits: 7.1.1 Sequential CEA Group Withdrawal 2 Re-evaluated 7.2.2 Loss of Coolant Flow 3 Re-evaluated 7 2 3 Full Length CEA Drop Re-evaluated 7 2.4 Transients Resulting from the Re-evaluated 4 Malfunction of Oge Steam Generator 7.2.5 Loss of AC Power Re-evaluated 73 Postulated Accidents 731 CEA Ejection Reanalyzed 732 Steam Line Rupture Reanalyzed 733 SteamGeneragorTubeRupture Reanalyzed 734 Seized Rotor Re-evaluated l 7.4 Postulated occurrences 7.4.1 Fuel Handling Reanalyzed l I Technical Specifications preclude this event during operation. 2 Requires High Power and Variable High Power Trip. l 3 Requires Low Flow Trip. l 4 Requires trip on high differential steam generator pressure. 8 5
. . l l TABLE 7-2 CALVERT CLIFFS UNIT 2, CYCLE 5 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CIM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle Values (Unit 1, Unit 2, Physics Parameters Units Cycle 6) {ycle 5 Values Radial Peaking Factors Forf}NBMarginAnalyses (F r Unrodded Region 1 70+'* 1 70+'* Bank 5 Inserted 1.87+'* 1.87+'* Forp)anarRadialCompoent (F of 3-D Peak (CN Limit Analyses) Unrodded Region 1.70+'* 1.70* Bank 5 Inserted 1.87+'* 1.87* Maximum Augmentation 1.055 1.055 Factor Moderator Temperature 10-4ap/0F -2.5**++.5 -2 555* .5 Coefficient Shutdown Margin (Value %ap -5 3 -5 2 Assumed in Limiting EOC Zero Power SLB) Tilt Allowance % 30 3.0
#For DNBR and CTM calculations, effects of uncertainties on these - parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c. These procedures have been approved by NRC for the Calvert Cliffs Units in Reference 3 d'Ihe effective initial MTC assumed for the SLB event is -2.2x10~4ap /0F.
+1he values assumed are conservative with respect to the Technical Specification limits.
1 TABLE 7-2 (continued) Reference Cycle Values (Unit 1, Unit 2, Safety Parameters Units Cycle 6) Cycle 5 Values Power Level MWt 2700* 2700* Maximum Steady State CF 548' 548* Temperature Minimum Steady State psia 2225' 2200* RCS Pressure Peactor Coolant Flow 1061bm/hr 138.5* 138.5' Negative Axial Shape I p .15* .15' LCO Extreme Assumed at Full Power (Ex-Cores) Maximum CEA Insertion % Insertion 25 25 at Full Power of Bank 5 Maximum Initial Linear KW/ft 16.0 16.0 Heat Rate for Transient Cther than LOCA Steady State Linear KW/ft 21 3 22.0 Heat Rate for Fuel CTM Assumed in the Safety Analysis CEA Drop Time from sec 31 31 Removal of Power to Holding coils to 90% Insertion Minimum DNBR (CE-1) 1.23* 1.23* l ! #For DNBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to CNB and CTM limits are detailed in References 2a, 2b, 2c. These procedures have been approved by NRC for the Calvert Cliffs Units in Reference 3 i I
1 TABLE 7-3 i DESIGN BASIS EVENT REANALYZED FOR UNIT 2, CYCLE 5 Reason for Acceptance Summary Event Reanalysis Criterion of Results (changes relative to reference cycle) Loss of Increase in initial Peak pressure Peak pressure calcu-Load pressurizer liquid less than upset lated to be 2617 level and decrease pressure limit psia. Further in initial of 2750 psia. details in Section pressurizer pressure. 7.1 3 Excess Decrease in initial Impact of Reactor Results acceptable. Load liquid level and Vessel Upper Head Further details in decrease in initial voiding on reactor Section 7.1.4. pressurizer pressure. coolant circula-tion and DNBR limits not exceeded. Excessive Increase in initial Time to fill Results show that Charging pressurizer liquid pressurizer. operator has at level. least 15 minutes from the initiation of high pressurizer level alarm to terminate the event. Further details in Section 7 1.8. CEA Lower available scram Fuel failure Results show that Ejection worth at trip. Changes small fraction no pin experiences in post ejected 3-D of 10CFR100. clad damage. peak and ejected CEA Further details worth, in Section 7 3 1. Steam Line Changes in moderator Post trip D'!3R Results show minimum Rupture cooldown curve and greater than 1 3 DNBR is equal to available scram worth with MacBeth 135 for HFP and 2.0 at trip. correlation. for HZP. Further details in Section 7 3 2. 5 Y
l TABLE 7-3 (continued) ' Reason for Acceptance Summary Event Reanalysis, Criterion of Results (change 3 relative to reference cycle) Steam Increase in initial Site boundary Results acceptable. Generator pressurizer liquid dose less than Further details in Tube Rupture level. 10CFR100 limits. Section 7 3 3 Fuel High burnup. Site boundary Results acceptable. Handling dose less than Further details in 10CFR100 limits. Section 7.4.1. r 9
, . . - , . . - , . . ~ . , , , - , ,
)
l I 7.1 3 Loss of Load Event The Loss of load event was reanalyzed for Cycle 5 to determine that the RCS pressure upset limit of 2730 psia is not exceeded. The transient DNBR was also evaluated to ensure that the results are within the design limit of 1.23 The methods used to analyze .this event are consistent with those reported in the reference cycle (Reference 1). Be assumpticns used to maximize RCS pressure during the transient are:
- a. The event is assumed to result from the sudden closure of the turbine stop valves without a simultaneous reactor trip. Bis asstanption causes the greatest reduction in the rate of heat removal from the reactor coolant system and thus results in the most rapid increase in primary pressure and the closest approach to the RCS pressure upset limit.
- b. The steam dump and bypass system, the pressurizer spray system, and the power operated pressurizer relief valves are assumed not to be operable.
This too maximizes the primary system pressure reached during the transient. The Loss of Load event was initiated at the conditions shown in Table 71 3-1. The combination of parameters shown in Table 7.1 3-1 maximizes the calculated peak RCS pressure. As can be inferred from the table, the key parameters for this event are the initial primary and secondary pressures and the moderator and fuel temperature coefficients of reactivity. The initial core average axial power distribution for this analysis was assumed to be a bottom peaked shape. Bis distribution is assumed because it minimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure an The Moderator Temperature Coefficient (KrC) of +0.5x10 g heat 40/ Fflux wasincreases. assumed in this analysis. Bis MTC, in conjunction with the increasing coolant temperatures, maximizes the rate of change of heat flux and the pressure at the time of reactor trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions was used in the analysis. Bis FTC causes the least amount of negative reactivity feedback to mitigate the transient increases in both the core heat flux and the pressure. Be uncertainty on the FTC used in the analysis is shown in Table 7.1 3-1. The lower limit on initial- RCS pressure is used to maximize the rate of change of pressure, and thus peak pressure, following trip. The Loss of Load event, initiated from the conditions given in Table 71 3-1, results in a high pressurizer pressure trip signal at 6.8 seconds. At 10.1 seconds, the primary pressure reaches its maximum value of 2617 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 5.8 seconds. The secondary pressure reaches its maximum value of 1047 psia at 10.6 seconds after initiation of the event. Table 7.13-2 presents the sequence of events for this event. Figures 7.1 3-1 to 71 3-4 show the transient behavior of power, heat flux, RCS pressure, and RCS coolant temperatures. Y-
The event was also reanalyzed with the initial conditions listed in Table 7 1 3-3 to determine that the acceptable DNBR limit is not exceeded. The minimum transient DNBR calculated for the event is 134, compared to the design limit of 1.23 The results of this analysis demonstrate that during a Loss of Load event the peak RCS pressure and the minimum DNBR do not exceed their respective design limits. i
+
1 8 I
TABLE 7.1 3-1 KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD ANALYSIS TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Reference
- Parameter Units Cycle Cycle 5 Initial Core Power Level MWt 2754 2754 Initial Core Inlet Coolant OF 550 550 Temperature Core Coolant Flow x1061bm/hr 133 9 133 9 Initial Reactor Coolant psia 2200+ 2154++
System Pressure a Initial Pressurizer ft3 800 975 Liquid Level at Full Power Initial Steam Generator psia 864 864
, Pressure Moderator Temperature x10~4Ap / F +0 5 +0.5 Coefficient IXappler Coefficient 0.85 0.85 Multiplier CEA Worth at Trip %ap -4.7 -4.7 Time to 90% Insertion sec 31 31 of Scram Rods Reactor Regulating System Operating Mode Manual Manual Steam Dtanp and Bypass Operating Mode Inoperative Inoperative
' Unit 1, Cycle 6 (Reference 1)
+ Corresponds to Technical Specification minimum indicated pressure of 2225 psia. The value includes an uncertainty of 25 psia.
" Corresponds to Technical Specification minimum indicated pressure of 2200 psia. The value includea n m eertainty of 46 psia.
E 5
t TABLE 7.1 3-2 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Time (sec) Event Setpoint or Value 0.0 Loss of Secondary Load - 5.8 Steam Generator Safety Valves 1000 psia Open e-6.8 High Pressurizer Pressure Analysis 2422 psia Trip Setpoint is Reached 8.2 CEAs Begin to Drop Into Core - 79 Pressurizer Safety Valves Open 2500 psia 10.1 Maximum RCS Pressure 2617 psia 10.6 Maximum Steam Generator Pressure 1047 psia 12 3 Pressurize Safety Valves are 2500 psia Fully Closed i 6 S I
TABLE 7 1.3-3 KEY PARAMETERS ASSUMED IN THE LOSS CF LOAD ANALYSIS TO CALCULATE TRANSIENT MINIMUM DNBR Reference
- Unit 2, Parameter Units Cycle Cycle 5 Initial Core Power Level MWt 2700** 2700**
CF Initial Core Inlet Coolant 548** 548** Temperature Core Coolant Flow x10kbm/hr 138.55* 138.5** Initial Reactor Coolant psia 2225** 2200** System Pressure Initial Steam Generator psia 864 864 Pressure Integrated 1 75**'+ 1.75**'+ Factors,FgadialPeaking (Bank 5 Inserted 2h%) Moderator Temperature x10-$p/F0
+0.5 +0.5 Coefficient Doppler Coefficient 0.85 0.85 Multiplier CEA Worth at Trip %ap -4.7 -47 Time to 90% Insertion sec 31 31 of Scram Rods Reactor Regulating Operating Mode Manual Manual System Steam Dunp and Bypass Operating Mode Inoperative Inoperative System
' Unit 1, Cycle 6 (Reference 1)
** Effects of uncertainties on these parameters were accounted for statistically. (See Reference 2)
+7he values assumed are conservative with respect to the Technical Specification limits.
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I 7 1.4 Excess Ioad Event The Excess Lead event was reanalyzed for Unit 2, Cycle 5 to evgluate the impact The analysis of initiating the event from a pressurizer level of 270 ft . evalcated the impact of reactor vessel upper head (RVUH) voiding on reactor coolant circulation and fuel design limits. he analysis included the automatic initiation of auxiliary feedwater three minutes after initiation of reactor trip signal and a manual trip of the reactor coolant pumps (RCPs) following a safety injection actuation signal (SIAS) due to low pressurizer pressure. he RCP coastdown results in a proportionately reduced RVUH flow until natural circulation is established; at that time all flow to the RVUH is assumed to terminate. he Excess Load event is initiated by the instantaneous opening of steam dump and bypass valves which have a combined capacity of 45% of nominal full power steam flow. This Excess Load event persists until the steam generators are isolated on steam generator isolation signal (SGIS) due to low secondary pressure. The full power event maximizes primary cooldown, shrinkage and consequently RVUH voiding. The magnitude of RVUH voiding at full power is significantly greater than at zero power because of the higher RVUH coolant temperatures and the larger primary coolant shrinkage that occurs at the higher system temperatures. Therefore, only the full power excess load transient results are described herein. The key parameters assmed in the analysis to maximize RVUH is given in Table 7.1.4-1. The key parameters assumed to minimize DNBR is given in Table 7 1.4-2. The analysis to maximize RVUH voi conservatively assumed a Moderator Temperature Coefficient of 4.5x10 ging ao / F. Bis MTC in combination with decreasing coolant temperatures inserts negative reactivity and causes the core power to decrease. The decreasing core power does not allow either the High Power trip or IN/LP trip to be initiated and thus the time of reactor trip is delayed until a low pressurizer pressure trip (i.e., floor of the TM/LP trip) is generated. The longer time required to initiate reactor trip causes the pressurizer to drain and thus maximizes RVUH voiding. The analysis conservatively assumed that all three charging pumps were inoperable and that one High Pressure Safety Injection (HPSI) pmp fails to start on SIAS due to low pressurizer pressure. The effect of auxiliary feedwater was explicitly evaluated by analyzing the event both with and without auxiliary feedwater initiated three minutes after reactor trip signal is generated. An auxiliary feedwater flow of 172 lbm/see to each steam generator is conservatively assumed (i.e., 10.5% of fbli power l main feedwater flow per generator).- The maximum auxiliary feedwater flow l causes the fastest primary cooldown and thus enhances the bubble formation in the upper head. This analysis shows that some RVUH voiding will occur as a result of the RCS depressurization caused by an Excess Load transient. Voiding in the RVUH starts when RCS pressure control is lost due to the primary coolant shrinkage which drains the pressurizer. During the period of RVUH voiding, RCS pressure is controlled by the saturation pressure within the upper head. - This reduces t 5~
l l the RCS depressuri::ation in the latter part of the transient as compared to analyses which do not explicitly model RVUH voids. RVUH coolant temperatures initially follow the core outlet temperature and therefore decrease imediately following a reactor trip. The decrease in RVUH coolant flow reduces the convective heat transfer and inhibits upper head cooldown. This leads to elevated RVUH coolant temperatures which raise the upper head saturation pressure and therefore increase RCS pressure during the period of voiding. The increased RCS pressure does not adversely impact the approach to any SAFDL; however, safety injection flow is decreased. The reduced safety injection flow does not result in a return to criticality. However, the decreased flow diminishes the mitigating effect of safety injection on coolant shrinkage and, therefore, enhances voiding. Subsequent RVUH cooldown is accomplished through an exchange of coolant between the RVUH j and the core outlet plenurt. This exchange of coolant is driven by the expansion and contraction of the steam bubble. Additional RVUH cooling is accomplished through heat ccnduction across the upper guide structure. At the time of maximum RVUH voiding approximately 63% of the head is occupied by steam. Since this steam bubble does not expand beyond the upper head, primary coolant circulation is unaffected. Tables 7.1.4-3 and 7 1.4-4 present the sequence of events for the event initiated without and with auxiliary feedwater flow. Figures 7.1.4-1 through 7.1.4-14 present the transient behavior of the system variables during the event. The analysis demonstrates that the addition of auxiliary feedwater prolongs the duration of RVUH voiding and delays repressurization of the RCS. However, since auxiliary feedwater is delivered after the time of maximum voiding, the peak void fraction is unchanged. The Excess Load event initiated from the conditions given in Table 7.1.4-2 resulted in a minimum DNBR of 1.46 compared to the design limit of 1.23 (CE-1 correlation). wigh an Excess. Load In conclusion, the potential RVUH voiding associated transient initiated from a pressurizer level of 270 ft does not extend beyond the upper head and therefore will not affect primary coolant j circulation. In addition, approach to SAFDLs are not impacted by RVUH
- voiding. The results of the analysis also shows that the NSSS achieves stable
! conditions and that shutdown cooling procedures can be initiated if deemed necessary. l \ . Y
, , _ _ . _ . _ , ._ _ , , __ _,.~. , . . - __ . _ _ _ ..
4 TABLE 7 1.4-1 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD EVENT ANALYSIS TO MAXIMIZE RVUH VOIDING Parameter Units Value Initial Core Power Level MWt 2754 Core Inlet Temperature F 550 Reactor Coolant System Pressure psia 2300 Core Mass Flow Rate x1061bm/hr 133 6 Moderator Temperature x10-4Ap/cF +.5 Coefficient CEA Worth Available at Trip %Ao -4.7 i Auxiliary Feedwater Flow Rate Ibm /sec 172.0/S.G. kw Pressurizer Pressure psia 1728 Analysis Trip Setpoint SIAS Analysis Setpoint psia 1556 SGIS Analysis Setpoint psia 548 Initial Pressurizer Level ft3 270 S
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TABLE 7 1.4-2 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD EVENT ANALYSIS TO CALCULATE TRANSIENT MINIMUM DNBR Reference Unit 2, Parameter Units Cycle
- Cycle 5 Initial Core Power Level MWt 2700+ 2700+
Core Inlet Temperature F 548+ 548+ Reactor Coolant System psia 2225+ 2200+ Pressure Core Mass Flow Rate x100lbm/hr 138.5+ 138.5+ Moderator Temperature x10-4apFF -2.5 -2.5 Coefficient CEA Worth Available %ap -4 3 -4 3 at Trip J Doppler Multiplier 0.85 0.85 , Inverse Boron Worth PPM /% 105 105 - Auxiliary Feedwater ltxr/sec 175.0/S.G. 175 0/S.G. Flow Rate High Power Level Trip % of full power 110 110 Setpoint Low S.G. Water Level ft 30 9 30 9 Trip Setpoint RTD Response Time see 12.0 12.0
# Reference cycle is Unit 1 Cycle 6, Reference 1.
+For DNBR calculations, effects of uncertainties on these parameters were combined statistically (see Reference 2).
l
TABLE 7 1.4-3 SEQUENCE OF EVENTS FOR THE EXCESS LOAD EVENT WITHOUT AUXILIARY FEEDWATER Time (sec) Event Setpoint or Value 0.0 Steam Dtap and Bypass Valves - Fully Open 24.2 Pressurizer Empties - 28.9 Low Pressurizer Pressure Analysis 1728 psia Trip Setpoint is Resched 29 8 Trip Breakers Open 30 3 CEAs Begin to Drop Into Core Feedwater Starts Rampdown - 32.6 SIAS is Initiated 1556 psia Reactor Coolant Pumps Manually Tripped 67.1 SGIS is Generated 548 psia 68.0 Main Steam Isolation Valves - Begin to Close 80.0 Main Steam Isolation Valves - are Closed 90 3 Feedwater Rampdown to 5% is Completed 107.6 Maximum RVUH Void is Reached 63% ( 210 3* Main Feedwater Isolated 521.6 Minimum RCS Pressure 728.50 717.8 Upper Head Void is Zero l Pressurizer Starts to Refill i
# Main feedwater would have been isolated 80 seconds after SGIS is initiated (i.e., at 147 1 seconds). The analysis conservatively assumed that main feedwater is isolated at 210 3 seconds to prolong the duration of RVUH voiding.
I
TABLE 7.1.4-4 SEQUENCE OF EVENTS FOR THE EXCESS LOAD EVENT WITH AUXILIARY FEEDWATER Time (sec) Event Setpoint or Value 0.0 Steam Dep and Bypass Valves - Fully Open 24.2 Pressurizer Empties - 28.9 Low Pressurizer Pressure Analysis 1728 psia Trip Setpoint is Reached 29 8 Trip Breakers Open 30 3 CEAs Begin to Drop Into Core Feedwater Starts Repdown - Turbine Valves Begin to Close 32.6 SIAS is Initiated 1556 psia Reactor Coolant Pumps Manually Tripped 67 1 SGIS 13 Generated 548 psia 68.0 Main Steam Isolation Valves Begin to Close 80.0 Main Steam Isolation Valves are Closed 90 3 Feedwater Rampdown to 5% is Completed 107.6 Maximum RVUH Void is Reached 63% 210 3* Main Feedwater Isolated 172 lbm/S.G. Auxiliary Feedwater is Initiated 900.0 Operator Action To Isolate Auxilary Feedwater to Stean Generators
# Main feedwater would have been isolated 80, seconds after SGIS is initiated (i.e., at 147.1 seconds). The analysis conservatively assumed that main feedwater is isolated at 210 3 seconds to prolong the duration of RVUH voiding.
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7 1.8 Excessive Charging Event The Excessive Charging event initiated from gl pressurizer level was analyzed for Unit 2 Cycle 5 to assure that the o)imumptrator has at least fifteen (15) minutes from initiation of a high prepurizer level alarm to take corrective action and terminate the event prier to filling the pressurizer solid. The Excessive Charging event is assumed to occur by inadverteit initiation of charging flow. The time required to fill the pressurizer solid was calculated using Equation 7 1.8-1. V3-VSL - VT v Eq. 7 1.8-1 CH LD) I where: V3 = steam volume in the pressurizer V3t = equivalent saturated liquid volume of pressurizer steam volume VT
= volume above the spray nozzles FCH = charging flow rate Fpt = letdown flow rate vj = specific volume of liquid at charging and letdown conditions v
2 = specific volume of liquid at pressurizer conditions The analysis was performed for three combinations of charging and letdown flows. Table 71.8-1 presents the initial conditions assumed in the analysis and the results of the analysis. The initial conditions assumed in the analysis are consistent with Technical Specification limits on initial pressurizer level including appropriate uncertainties. As seen from the table, all three combinations of charging and letdown flows analyzed provide at least fifteen minutes after initiation of high level alarm for the operator to take corrective actions and terminate the event prior to filling the pressurizer solid. S I
TABLE 7.1.8-1 fiaximum Initial Pressurizer Level liigh Level Alarm Time to Fill (j) Volume Control Assumptions Assumed in Analysis Analysis Setpoint Pressurizer Chargin Flow Letdown Flow Liquid Volume Levef3) Liquid Volunn. Level (3) (minutes) (GPM (GPM) (ft 3) fin) 3 (ft ) (in)
- 1. 132 0 915 227 920 228 15 *
- 2. 132 29 975(2) 242 1040 257 15
- 3. 88 0 975(2) 242 1100 272 15 II)From time of initiation of high pressurizer level alarm (2)tiaximum limit based on Loss of Load event (3) Referenced to the 1" level nozzle at the bottom of the pressurizer
- c. .
731 CEA Ejection Event Be CEA Ejection event was reanalyzed for Unit 2 Cycle 5 to determine the fraction of fuel pins that exceed the criterion for clad damge. The analytical method employed in the reanalysis of this event is the NRC approved Combustion Engineering CEA Ejection method whtet U described in CENPD-190-A, (Reference 4). Be key parameters used in this event are listed in Tabic ?.3,i-1. With these key parameters, selected to add conservatism, the procedure outlined in Figure 2.1 of Reference 4 is then used to determine tha ava aEe and centerline enthalpies in the hottest spot of the rod. The calculate 3 enchalpy values are ecmpared to threshold enthalpy values to determine the aoount sf fuel exceeding these thresholds. These threshold enthalpy values are (Ref; inces 5, 6, and 7). Clad Damage Threshold: Total Average Enthalpy = 200 cal /gm Incipient Centerline Melting Threshold: Total Centerline Enthalpy = 250 cal /gr Fully Molten Centerline Threshold: Total Centerline Enthalpy = 310 cal /gm To bound the most adverse conditions during the cycle, the most limiting of either the Beginning of Cycle (BOC) or End of Cycle (EOC) parameter values were used in the analysis. A BOC Doppler defect was used since it produces the least amount of negative reactivity feedback to pitigate the transient. A BOC moderator temperature coefficient of +0.5x10- ap / F was used because a positive MIC results in positive reactivity feedback and thus increases coolant temperatures. An EOC delayed neutron fraction was used in the analysis to produce the highest power rise during the event. The zero power CEA Ejection event was analyzed assuming the core is initially operating at 1 Wt . At zero power, a Variable High Power trip is conservatively assumed to initiate at 40% (30% + 10% uncertainty) of 2700 Wt and terminates the event. The full and zero power e ries were analyzed, assuming the value of 0.05 seconds for the total ejection tine, which is consistent with the FSAR and previous reload submittals. The power transient produced by a CEA ejection initiated at the maximum allowed power is shown in Figure 7 3 1-1. Similar results for the zero power case are shown in Figure 7 3 1-2. Be results of the two CEA ejection cases analyzed (Table 7.31-2) show that the maximum total energy deposited during the event is less than the criterion for clad damage (i.e. , 200 cal /gm). Also, an acceptably small fraction of the fuel reaches incipient centerline melt threshold. I
t TABLE 7 3 1-1 KEY PARAMETERS ASSUMED IN THE CEA EJECTION ANALYSES Reference Unit 2 Parameter Units Cycle
- Cycle 5 Full Power Core Power Level Wt 2754 2754 Core Average Linear Heat KW/ft 6.12 6 36 Generation Rate at 2754 Wt Moderator Temperature 10-46p / F +.5 +.5 Coefficient Ejected CEA Worth kp 32 .28 Delayed Neutron .0047 .0044 Fraction, Post-Ejected Radial 3 36 3.6 Power Peak Axial Power Peak 1 39 1.44 CEA Bank Worth at Trip kg -3 88 -3 0 Tilt Allowance 1.03 1.03 Doppler Multiplier 0.f5 0.85 Zero Power Core Power Level Wt 1.0 1.0 Ejected CEA Worth kp .60 .63 Post-Ejected Radial 9 83 9 40 Power Peak l
- Axial Power Peak 1.60 1.60 CEA Bank Worth at Trip to -2.58 -1.50 j Tilt Allowance 1.10 1.10 CEA Drop Time sec 31 31 Doppler Multiplier 0.85 0.85
' Reference cycle is Unit 1, Cycle 4 (Reference 8).
t .
, , s s
s g TABLE 7 3 1-2 CEA EJECTION EVENT RESULTS Reference Cycle Unit 2 Full Power Unit 1, Cycle 4 Cycle 5 s Total Average Enthalpy of Hottest Fuel 198. 185 Pellet (cal /gm) Total Centerline Enthalpy of Hottest 268. 293 Fuel Pellet (cal /gm) Fraction of Rods that Suffer Clad 0 0 , Damage (Average Enthalpy 1200 cal /gm) l? Fraction of Fuel Having a Least .01 .08 Incipient Centerline Melting (Centerline Enthalpy > 250 cal /gn) Fraction of Fuel Having a Fully Molten 0 0 Centerline Condition (Centerline Enthalpy 1310 cal /gm) Reference Cycle Unit 2 Zero Power Unit 1, Cycle 4 Cycle 5 , Total Average Enthalpy of Hottest Fuel
- 177 145 Pellet (cal /gm) -
Total Centerline Enthalpy of Hottest 177 199 Fuel Pellet (cal /gm) Fraction of Rods that Suffer Clad 0 0 Damage (Average Enthalpy 1 200 cal /gn) Fraction of Fuel Having a Least 0 0 Incipient Centerline Melting (Centerline Enthalpy 1 250 cal /gm) Fraction of Fuel Having a Fully Holten 0 0 Centerline Condition (Centerline
- Enthalpy > 310 cal /gs) Total 8
e
i i i . FULL POWER i EJECTED WORTH = 0'.28%Ap tE 2.0 " - r . 8 s es E D o si u_
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d S . 0 0 1 2 3 4 5 TIME, SECONDS BALTIMORE GAS & ELECTRIC CO. CEA EJECTION EVENT FIGURE calvert c11rts CORE POWER VS TIME 7.3.1-1 Nuclear Power Plant
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l
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732 Steam Line Rupture Event he Steam Line Rupture (SLB) event was analyzed for Cycle 5 to determine that the critical heat flux is not exceeded during this event. The analysis included the effect of automatic initiation of auxiliary feedwater flow in three (3) minutes from the initiation of a Low S.G. level signal, a manual trip of the Reactor Coolant Pumps on Safety Injection Actuation Signal due to low pressurizer pressure, and an MSIV closure time of 12 seconds.# The analysis assumed that the event is initiated by a circumferential rupture of a 34-inch (inside diameter) steam line at the steam generator main steam line nozzle. Bis break size is the most limiting, since it causes the greatest rate of temperature reduction in the reactor core region. The SLB event was analyzed with the assumption of e three minute delay between the time of reactor trip on 14w S.G. pressure and the time when Auxiliary Feedwater (AW) flow is delivered to the affected steam generator. Bis is conservative with re::pect to the expected time of AFW initiation since the generation of the AFW signal actually occurs at the time of the Low Stem Generator Water Level trip signal which occurs later than the Low S.G. Pressure trip. We analysis assumes, therefore, that AFW flow is delivered to the steam generator sooner than the flow is actually available resulting in a conservative prediction of the resulting cooldown. A conservatively high value of the AW flow was calculated assuming that all auxiliary feedwater pmps are operable. An AW flow value of 21% of full power feedwawter flow was used in the analysis. This value accounts for pump run-out due to reduced back pressure. In addition, the analysis conservatively assumed that all the AFW flow is fed only to the damaged steam generator. The analyses assumed that the main feedwater flow is ramped down to 8% of full power feedwater flow in 20 seconds and that the main feedwater isolation valves are completely closed in 80 seconds after a low steam generator pressure or a main steam isolation signal. These assumptions are consistent with Technical Specification limits (see Table 3 3-5). During a Return-To-Power, negabin reactivity credit was assumed in the analysis. This negative reactivity credit is due to the local heatup of the inlet fluid in the hot channel, which occurs near the location of the stuck CEA. Bis credit is based on three-dimensional coupled neutronic-thermal-hydraulic calculations performed with the HERMITE/ TORC code (References 9 and
- 10) for St. Lucie Unit 2, Cycle 1 (Reference 11). It should be noted that only a small fraction of the negative reactivity credit justified for St. Lucie Unit 2 was included in the SLB event analysis for Calvert Cliffs Unit 2, Cycle 5 The manual trip of the RCPs is assumed to result in no flow mixing at the core inlet plenum. Rus , cold edge temperatures were used to calculate the moderator reactivity insertion during the cooldown of the RCS following a SLB.
The two SLB cases censidered in conjunction with the automatic initiation of auxiliary feedwater flow and manual trip of RCPs are:
- Conservative with respect to Technical Specification limit of 6.0 seconds.
E 5
l
- 1. 2 Loop - Full Load (2754 Wt)
- 2. 2 Loop - No Load (1 Wt) i The 1 Loop - Full Load and 1 Loop - No Load cases were not analyzed since l Technical Specifictions prohibit operation in these modes.
The Two Loop-2764 Wt case was initisted at the conditions listed in Table 7 3 2-1. The Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis corresponds to end of life, since this MTC results in the greatest positive reactivity change during the RCS cooldown caused by the Steam Line Rupture. Since the reactivity change associated with moderator feedback varies significantly over the moderator density covered in the analysis, a curve of reactivity insertion versus density rather thari a single value of MTC, is assumed in the analysis. The moderator cooldown curve assumed is given in Figure 7 3 2-1. The moderator cooldown curve given in Figure 7 3 2-1 was + conservatively calculated assuming that on reactor scram, the Control Element Assembly is stuck in the fully withdrawn position which yields the most severe combination of scram worth and reactivity insertion. The reactivity defect associated with the fuel temperature decrease is also t based on an end of life Doppler defect. The Doppler defect' based on an end of life Fuel Temperature Coefficient (FTC), in conjunction with the decreasing fuel temperatures, causes the greatest positive reactivity insertion during the Steam Line Rupture event. The uncertainty on the FTC assumed in the analysis , is given in Table 7 3 2-1. The 8 fraction assumed is the. maximum absolute value including uncertainties for end of life conditions. This too is conservative since it maximizes suberitical multiplication and, thus, enhances the potential for Return-To-Power (R-T-P). The minimum CEA worth assumed to. be available for shutdown at the time of
-), reactor trip at the maximum allowed power level is-6.89% ap . ' This available scram worth was calculated for the stuck rod which produced the moderator cooldown curve in Figure 7 3 2-1.
The analysis conservatively assumed that - on Safety Injection Actuation Signal one High Pressure Safety Injection pump and one Low Pressure Safety Injection - pump fail to start. The analysis also asstuned a conservatively low value of boron reactivity worth at -1.05 ao per 95 PPM. The conservative assumptions on feedwater flow were discussed previously. The feedwater flow and enthalpy as a function of time are presented in Figures 7 3 2-2 and 7 3 2-3, respectively. Table 7 3.2-2 presents the sequence of events for the full power case initiated at the conditions given in Table 7 3 2-1. The reactivity insertion as a function of time is presented in ' Figure 7 3 2-4. The response of the NSSS-during this event is given in Figures 7 3 2-5 through 7 3 2-9 The results of the analyses show that - SIAS is actuated at.16.7 seconds, at which-time the Reactor Coolant Pumps are manually tripped by the operator. The manual trip of RCPs slows down the rate of primary heat. removal and, thus, delays the time when the affected steam generator blows dry. .The affected steam generator blows dry at ' 94.0 seconds and terninates the cooldown of the 8 I
- , , . , _ ,-,,, -. . _ - , . . _ _ . ,_ , -, ,m , , - - - - _- , . -..
RCS. De peak reactivity attained prior to delivery of Auxiliary feedwater flow is .052% to at 123 0 seconds. A peak R-T-P of 9 3%, consisting of 3 5% decay heat and 5.8% fission power is produced at 127.5 seconds. ne continued production of decay heat from the fuel after termination of blowdown, causes the reactor coolant temperatures to increase. Bis in turn reduces the magnitude of the positive moderator reactivity inserted and, thus, the total reactivity becomes more negative. The delivery of auxiliary feedwater flow to the affected steam generator at 183 7 seconds initiates a further cooldown of the RCS which results in more positive reactivity insertion. he positive reactivity insertion causes the core to approach criticality. Be peak criticality attained is +.044% ao at 584 3 seconds. ne reactivity transient is terminated by the boron injected via the High Pressure Safety Injection Pumps. A peak R-T-P of 3 1%, consisting of 2 3% decay heat and 0.8% fission power is produced at 624.0 seconds. The MacBeth correlation (Reference 12) with the Lee non-uniform mixing correlation factor (Reference 13) results in a post-trip minimum DNBR of 135 compared to the DNBR limit of 130 during a SLB event initiated from hot full power conditions. Bus , critical heat fluxes are not exceeded during this event. Two Loop-No Load case was initiated at the conditions given in Table 7 3 2-3 The moderator cooldown curve is given in Figure 7 3 2-10. The cooldown curve corresponds to an end of life MTC. An end of life FTC was also used for the reasons previously discussed in connection with the Two Loop-2754 MWt case. The minimum CEA worth assumed to be available for shutdown at the time of reactor trip at the zero power level is -5.2%Ao . This available scram worth was calculated for the stuck rod which produced the moderator cooldown curve in Figure 7 3 2-10. A maximtra inverse boron worth of 90 PPM /% ap was conservatively assumed for the safety injection during the no load case. Be feedwater flow and the enthalpy used in the analysis are presented in Figures 7 3 2-11 and 7 2 3-12, respectively. Table 7 3 2-4 presents the sequence of events for the Two Loop-No Load case initiated from the conditions given in Table 7 3 2-3 The reactivity insertion as a function of time is presented in Figure 7 3 2-13 He NSSS response during this event are given in Figures 7 3 2-14 to 7 3 2-18. The results of the analysis show that SIAS is actuated at 20 9 seconds, at which time the RCPs are manually tripped by the operator. Auxiliary feedwater flow is initiated at 183 3 seconds which continues the cooldown of the RCS. Thus, the total core reactivity approaches criticality. The peak reactivity attained is +0 34%ao at 450.8 seconds and a peak power of 313% occurs at 472 5 seconds. The addition of boron via High Pressure Safety Injection mitigates the reactivity transient. Be MacBeth correlation with the Lee non-uniform mixing correlation -factor results in a post-trip minimum DNBR of 2.00 compared to the CNBR limit of 1 30, during a SLB event initiated from hot zero power conditions. Thus, critical heat fluxes are not exceeded during this event, t 5
'Ihe Steam Line Rupture event initiated from HFP and H7.P conditions with !
automatic initiation of auxiliary feedwater flow and manual trip of RCPs on SIAS due to low pressurizer pressure shows that the DNBR limits are not exceeded. Since the DNBR limits are not exceeded and no fuel pins are predicted to fail, it is concluded that the consequences of the SLB event are acceptable for Cycle 5 operation. a I
. - , . . . - ~ - __ , _ .
TABLE 7 3 2-1 KEY PARAMETERS ASSUMED IN THE STEAM LINE RUPTURE ANALYSIS 2-LOOP-2754 MWT Reference Unit 2 Parameter Units Cycle
- Cycle 5 Initial Core Power MWt 2754 2754 OF 550 Initial Core Inlet 550 Temperature Initial RCS Pressure psia 2300 2300 Initial Steam Generator psia 871 860**
Pressure low Steam Generator psia 548 548 Pressure Analysis Trip Setpoint Safety Injection psia 1556 1645 Actuation Signal Minimum CEA Worth to -7.02 -6.89 Available at Trip Doppler Multiplier 1.15 1.15 Moderator Cooldown to vs. See Figure 7.3 2-1 See Figure Curve density of Reference 1 7 3 2-1 Inverse Boron Worth PPM /% ao 95 95 Effective MTC x10-So /0F -2.2 -2.2 l S Fraction (including .0060 .0060 uncertainty)
- Unit 1 Cycle 6 (Reference 1)
**Ihe difference in initial steam generator pressure 1:as negliable impact on the results.
8 I
TABLE 7 3 2-2 SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE EVENT WITH AUTOMATIC INITIATION OF AUXILIARY FEEDWATER AND MANUAL TRIP OF REACTOR COOLANT PUMP 2-LOOP-2754 MWT Time (sec) Event Setpoint or Value 0.0 Steam Line Rupture Occurs -- 23 Low Steam Generator Pressure 548.0 psia Analysis Trip Setpoint is Reached; Steam Generator Isolation Analysis Setpoint is Reached 32 Trip Breakers Open; Main Steam - Isolation Valves Begin to Close 37 CEAs Enter Core; - Main Feedwater Rampdown Begins 15 2 Main Steam Isolation Valves - Completely Closed 16.7 Safety Injection Actuation 1645 0 psia Signal 16.8 Reactor Coolant Ptanps - Manually Tripped 17 2 Pressurizer Empties - 23 7 Main Feedwater Rampdown 8% of fbil power Completed feedwater flow 46.7 High Pressure Safety Injection - Pumps at Full Speed 83 2 Main Feedwater Isolation - 94.0 Affected Stean Generator - Blows Dry 123 0 Peak Reactivity, Prior to .052% . Auxiliary Feedwater Flow ( 127.5 Peak Return to Power 9 3% l 183 7 Auxiliary Feedwater Flow -340 lbm/see Initiated to Ruptured Steam Generator i 584 3 Peak Reactivity Post Auxiliary +.044% Feedwater Flow l I * \ t
TABLE 7 3 2-2 (continued) Time (sec) Event Setpoint or Value 624.0 Peak Return to Power Post 3 1% Auxiliary Feedwater Flow 1200.0 Operator Isolates Ruptured - Steam Genrator and Terminates Auxiliary Feedwater Flow i l l l T I l l 8 5
- .-v. - , - ~ - ,, , . , . , , ~ . . . ,.e. , , . . . .
TABLE 7 3 2-3 KEY PARAMETERS ASSUMED IN THE STEAM LINE RUPTURE ANALYSIS 2-LOOP NO LOAD Reference Unit 2 Parameter Units Cycle # Cycle 5 Initial Core Power MWt 1.0 1.0 C Initial Core Inlet F 532 532 Temperature Initial RCS Pressure psia 2300 2300 Initial Steam Generator psia 900 900 Pressure Low Steam Generator psia 548 548 Pressure Analysis Trip Setpoint Safety Injection psia 1556 1645 Actuation Signal Minimum CEA Worth ko -5 3 -5 2 Available at Trip Doppler Multiplier 1.15 1.15-Moderator Cooldown Curve 1 vs. See Figure 7 3 2.10 See Figure density of Reference 1 7 3 2-10 Inverse Boron Worth PPM /5ap 90 90 Effective MTC x10-Nap /0F -2.2 -2.2 8 Fraction (including .0060 .0060 uncertainty)
# Unit 1 Cycle 6 (Reference 1)
I
TABLE 7 3 2-4 SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE EVENT WITH AUTOMATIC INITIATION OF AUXILIARY FEEDWATER AND MANUAL TRIP OF REACTOR COOLANT PUMP 2-LOOP-NO LOAD Time (sec) Event Setpoint or Value 0.0 Steam Line Rupture Occurs 6 305 ft 2 19 Low Ste m Generator Pressure 548.0 psia Analysis Trip Setpoint is Reached; Steam Generator Isolation Analysis Setpoint is Reached 2.8 Trip Breakers Open; Main Steam -- Isolation Valves Completely Closed 33 CEAs Enter Core - 14.8 Main Steam Isolation Valves - Completely Closed 20.9 Safety Injection Actuation 1645 0 psia Signal 21.0 Reactor Coolant Pumps Manually - Tripped 24 3 Pressurizer Empties - 50 9 High Pressure Safety Injection - Ptaps at Full Speed , 82.8 Main Feedwater Isolation - 183 3 Auxiliary Feedwater Flow 340 lbm/see Initiated to Ruptured Steam Generator 450.8 Peak Reactivity +0 34% 472.5 Peak Power 3 13% 1200.0 Operator Isolates Ruptured - Steam Generator and Terminates Auxiliary Feedwater Flow E 5
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, calvert cliffs FEEDWATER FLOW VS TIME 7.3.2-11 ikglear Power Plant
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733 Steam Generator Tube Rupture Event The Steam Generator Tube Rupture (SGTR) event was analyzed for Unit 2, Cycle 5 i to verify that the site boundary doses will not exceed the of 10CFR100 for the event initiated from a pressurizer level of 975 ftgidelines .
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The analysis included the effects of manually tripping the Reactor Coolant ! Ptanps on SIAS due to low pressurizer pressure. l The design basis SGTR is a double ended break of one steam generator U-tube. l Table 7 3 3-1 lists the key transient related parameters used in this analysis. In the analysis, it is assumed that the initial RCS pressure is as high as 2300 psia. This initial RCS pressure mcximiz% the amount of primary coolant transported to the secondary steam system since the leak rate is directly proportional to the difference between the primary and secondary pressure. In addition, the higher pressure delays the lower pressurizer pressure trip which prolongs the transient and, therefore, maximizes the total primary to secondary mass and activities transported. 4 For this event, the acceptable DNBR limit is not exceeded due to the action of the Thermal Margin / Low Pressure (IM/LP) trip which provides a reactor trip to maintain the DNBR above 1.23 Since the SGTR event does not significantly affect the core power distribution, the PLHGR SAFDL is not approached. . The Thermal Margin / Low Pressure trip, with conservative coefficients which account for the limiting radial and axial peaks, maximum inlet temperature, RCS pressure, core power, and conservative CEA scram characteristics, would be the primary RPS trip intervening during the course of the transient. However, to
; maximize the coolant transported from the primary to the secondary and, thus, the radioactive steam releases to the atmosphere, the analysis was performed assuming that the reactor does not trip until the minimum setpoint (floor) of the Thermal Margin / Low Pressure trip is reached. This prolongs the steam releases to the atmosphere and, thus, maximizes the site boundary doses.
The Steam Generator Tube Rupture was analyzed assuming a manual trip of .the reactor coolant ptanps on Safety Injection Actuation Signal (SIAS). The Steam Generator Tube Rupture (SGTR) with RCS trip on SIAS results in higher site boundary doses because: (1) RCP coastdown increases pressure difference between the primary and the secondary, which increases the leak rate, and (2) RCP coastdown decreases the rate of decay heat removal, which increases the steam flow through the atmospheric dump valves. The sequence of events for the SGTR event with manual trip of RCP on SIAS is presented in Table 7 3 3-2. Figures 7 3 3-1 through 7 3 3-5 present the transient behavior of. core power, heat flux, RCS pressure, RCS temperatures, and steam generator pressure. - I-131 activity release is based on the primary to secondary leak and on the steam flow required to reach cold shutdown conditions. This release is calculated as the product of steam flow, the time dependent steam activity and the decontamination factors applicable to each release pathway. The O to 2 hour I-131 site boundary dose is calculated from: DOSE (REM) = Rtotal
- BR # DCF
- x/Q t . - _ - . ._ , .., _
, where: Rtotal = the total activity released to the atmosphere, Ci BR = breathing rate, m 3f3 c X/Q = atmospheric dispersion coefficient, sec/m 3 DCF = Dose Conversion Factor in equivalent I-131, Rem /Ci In determining the whole body dose, the major assumption made is that all noble gases leaked through the ruptured tube will be released to the atmosphere. Therefore, the whole body dose is proportional to the total primary to secondary leak and is calculated using the following equation. Whole Body Dose = [Ky (k + Eg )]
- R
- X/Q
- At where:
Ey = the average y energy (MEV/ dis) for the halogen isotopes of concern Ig = the average S energy (MEV/ dis) for the halogen isotopes of concern R = the activity released to the atmosphere, Ci/sec X/Q = atmospheric dispersion coefficient, sec/m 3 K = .25 Rem x m3 x dis y Mev x see x Ci _ .23 Rem x m3 x dis b - Mev x sec x Ci at = time period (i.e. , O to 2 hours), sec
- The results of the analysis show that 85616 lbs. of primary coolant are transported to the steam generator secondary side. Based on this mass transport and the values in Table 7 3 3-3, the site boundary doses calculated are
Thyroid (DEQ I-131) = 0 34 REM Whole Body (DEQ Xe-133) = 0.18 REM The reactor protective system (TM/LP) is adequate to protect the core from exceeding the DNBR limit. The doses resulting from the activity released as a sensequence of a doub; t-ended rupture of one steam generator tube, assuming the maximum allowable Tech. Spec. activity for the primary concentration at. a core power of 2754 MWt, are significantly below the guidelines of 10CFR100. E h
- e. .
TABLE 7 3 3-1 KEY PARAMETERS ASSUMED IN THE STEAM GENERATOR TUBE RUPTURE EVENT KEY TRANSIENT RELATED PARAMETERS: Reference Unit 2 Parameter Units Cycle
- Cycle 5 Power MWt 2754 2754 MTC x10~4ap / F -2.5 -2.5 Doppler Coefficient 1.15 1.15 Multiplier Scram Worth %ap -4.3 -4.7 F 550 550 Tin RCS Pressure psia 2300 2300' Initial Core Mass x1061bm/hr 133 9 133 9 Flow Rate Initial Secondary psia 815 815 Pressure Tube ID inches .654 .654 Flow Constant 1.17 1.17 ASI (for scram) +.41 +.41
# Reference cycle is Unit 1, Cycle 5 (Reference 14).
1 TABLE 7 3 3-2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE EVENT WITH RCP COASTDOWN ON SIAS Time (sec) Event Setpoint or Value 0.0 Tube Rupture Occurs - 786.5 Low Pressurizer Pressure 1728 psia Analysis Trip Setpiont is Reached 787.6 Dtap Valves Open - 787 9 CEAs Begin to Drop Into Core - 791 3 Pressurizer Empties - 792.6 Safety Injection Actuation 1556 psia Signal Generated, RCPs Manually Tripped 795.5 Bypass Valves Open - 797 1 Maximum Steam Generator 906 psia-Pressure 854.0 Minimum RCS Pressure 1064 psia 1800.0 Operator Isolates Damaged - Steam Generator and Begins Cooldown'to 300 F 12797 0 Operator Initiates Shutdown Cooling (TAV = 300 F) E
TABLE 7 3 3-3 ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR THE STEAM GENERATOR TUBE RUPTURE Paraneter Units Value Reactor Coolant System Maximum pCi/gm 1.0 Allowable Concentration (DEQ I-131)I Steam Generator Maximum A pCi/gm .1 Concentration (DEQI-131)}lowable Reactor Coolant System Maximum pCi/gm 100/E AllowableConcentrationgr Noble Gases (DEQ Xe-133) Steam Generator Partition Factor - .1 Air Ejector Partition Factor -
.0005 Atmospheric Dispersion sec/M3 1.80x10-4 Coefficient 2 Breathing Rate M3 /sec 3 47x10-4 Dose Conversion Factor (I-131) REM /Ci 1.48x104 I Tech. Spec. limits "
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. E-BALTIMORE FIGURE GAS & ELECTRIC CO. STEAM GENERATOR TUBE FAILURE EVENT REACTOR COOLANT SYSTEM PRESSURE VS TIME 7.3.3-3 nucieaNowNSIant
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- C-FIGURE GAS & co. STEAM GENERATOR TUBE FAILURE EVENT calvert cliffs 7.3.3-4 REACTOR COOLANT SYSTEli TE11PERATURE VS TIllE i: Nuclear Power Plant l
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7.4.1 Fuel Handling Event The consequences of a fuel handling incident due to increasing the burnup to 40,000 MWD /T has been investigated for Calvert Cliffs fuel. The results of this investigation demonstrate that the results shown in Reference 15 will not be changed by increasing the burnup. The dose rate at the site boundary will not increase because the gas gap inventory will be less than the gas gap inventory shown in Reference 15 The Reference 15 gas gap activity is based on the hottest fuel assembly in the core, independent of time (burnup) during the fuel cycle. Since the radioactive fission products which contribute significantly to the dose rate at the site boundary reach maximum concentrations at relatively low burnups, the only significant influence of burnup is the increased release from the fuel pellet for a given fuel temperature, commonly called " enhancement". - Be fael temperature, in turn, is principally dependent on linear heat rate. The predicted linear heat rate for Calvert Cliffs Unit 2, Cycle 5 fuel rods has been calculated to determine the radioactive fission product release to the gas gap. The maximum linear heat rate for rods with burnups between 33,000 MWD /T and 40,000 MWD /T is about 7 5 KW/ft; this maximum occurs at 33,000 MWD /T and the linear heat rate decreases to less than 6.5 KW/ft for the maximum fuel rod exposure at EOC5 ne maximum fuel temperature is not high enough to have significant diffusion-type release from the fuel; the method of release will be primarily from knock-out or recoil. Consequently, the radioactive fission product release to the gas gap will be less than 1% of the inventory; this release is based on the ANS 5.4 Standard, " Method for Calculating the i Fractional Release of Volatile Fission Products from Oxide Fuel." The release of less than 1% is over a factor of ten lower than that assumed in the fuel handling accident in Reference 15 i l l i'
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8.0 ECCS Analysis An ECCS performance analysis was performed for Calvert Cliffs Unit 2 Cycle 5 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Reacters (1). The analysis justifies an allowable peak linear heat generation rate (PLHGR) of 15.5 kw/ft which is equal to the existing limit for Unit 2. The ECCS analysis performed for Unit 2 Cycle 5 used input data, including the reload fuel performance parameters, which apply to both Unit 2 Cycle 5 and Unit 1 Cycle 6. Reference 2 contains the details of the method, results and conclusions of the analysis. They are directly applicable to both Unit 2 Cycle 5 and Unit 1 Cycle 6. The results of the analysis identified the peak clad temperature to be 2038 F, ccmpared to the acceptance criteria limit of 2200 F. The peak local clad oxidation was 8.5% and the peak core wide clad oxidation was less than 0.51% versus the acceptance criteria limits of 175 and 1.0%, respectively. Hence, Unit 2 Cycle 5 operation at a peak linear heat generation rate of 15.5 kw/ft and at a power level of 2754 MWt (102% of 2700 MWt) will result in acceptable ECCS performance.
9 0 Technical Specifications The Technical Specification changes which must be made in order to make the Calvert Cliffs Unit 2 Technical Specifications valid for the operation of Cycle 5 are presented in this section. Table 9-1 presents a sumary of the Technical Specification changes. Table 9-2 presents the explanations for the changes stanarized in Table 9-1. The requested Technical Specification modifications for Unit 2 Cycle 5 (Table 9-1) are very similar to those changes requested for the reference cycle (Unit 1 Cycle 6, References 1 and 2). The differences between the Technical Specification modifications being requested herein for Unit 2 Cycle 5 and those requested for Unit 1 Cycle 6 are explained in Table 9-3 The most noteworthy of these differences are:
- 1. The centerline melt limit is being increased to 22.0 kw/ft, compared to the Unit 1 Cycle 6 value of 213 kw/ft, to increase operating margins and flexibility.
- 2. The KrC surveillance requirements are being modified, compared to no change fcr Unit 1 Cycle 6, to allow the use of MTC determinations made during power ascension startup measurements for the purpose of satisfying surveillance requirements.
3 The minimum pressurizer pressure is being decreased from 2225 psia to 2200 psia, compared to no change for Unit 1 Cycle 6, to improve operating flexibility.
- 4. The FT x Technical Specification is being increased to 1.70, compared to the Unit 1 Cycle 6 value of 1.65, to increase operating margins and flexibility.
5 A relaxed pressurizer level band is being incorporated, compared to no change for Unit 1 Cycle 6, to improve operating flexibility. Following Table 9-3, for each Technical Specification which must be modified, either the existing page with the intended modification or the already modified page with a new figure is provided. l I _ _
Table 9-1 Calvert Cliffs II Cycle 5 Technical Specification Changes Change Tech Spec # - Action 1 Figure 2.1-1, Replace Figure 2.1-1 with Figure 2.1-1 page 2-2 2 Table 2.2-1 Change Steam Generator Low Pressure Trip page 2-9 setting fra 570 psia to 635 psia 3 Table 2.2-1 Change Steam Generator Low Pressure Trip page 2-10 Bypass limit from 685 psia to 710 psia 4 Figure 2.2-1 Replace Figure 2.2-1 with Figure 2.2-1 page 2-11 5' B.2.1.1 Change LHGR to centerline melt limit page B2-1 from 21 kw/ft to 22.0 kw/ft. 6 B.2.1.1, B.2.2.1 Change minimum DNBR value from 1.195 pages B2-1, B2-3 to 1.23 as indicated on noted pages B2-5, B2-6 7 B.2.1.1, B.2.2.1 Change high power level trip and maximum pages B2-1, B2-4 high power level trip actuation from 112% of rated thermal power to 110% 8 B.2.2.1 Change Steam Generator Low Pressure Trip page B2-5 setting from 570 psia to 635 psia, change uncertainty frm 22 psia to 87 psia, and revise description of uncertainty. 9 B.2.2.1 Change %e minimum trip setting (floor) page B 2-6 of the ..ermal Margin / Low Pressure (TM/LP) Trip from 1750 psia to 1875 psia 10 B.2.2.1 Revise description of IM/LP trip, and page B2-7 change allowance from 92 psia to 40 psia 11' Change shutdown margin, T 0 3/4.1.1.1 page 3/4 1-1 from 4 3% ak/k to 5.2% akgg >200 F, 12' 4.1.1.4.2 Change MTC surveillance Item (b) page 3/4 1-6 13 Figure 3 1-2 Replace Figure 3 1-2 with Figure 3 1-2 page 3/41-27 14 4.2.1 3 Change Figure 3 2-3 to Figure 3 2-3b page 3/4 2-2
Table 9-1 (continued) Change Tech Spec # Action 15 Figure 3 2-2 Replace Figure 3 2-2 with Figure 3 2-2 page F 4 2-4 16' Figure 3 2-3b Insert new Figure 3.2-3b after page F4 2-4a (new) Page 3/4 2-4 17' 3/4.2.2.1 (new) Change calculated value of F T from 3/4.2.2 (old) 1.620 to 1.700, change Figure x7 3.2-3 to pages 3/4 2-6, Figure 3.2-3a and change Tech Spec F4 2-7 numbers from y4.2.2 to y4.2.2.1 18* Figure 3 2-3a (new) Replace Figure 3 2-3 with new 3 2-3 (old) Figure 3 2-3a and change page number from page 3/4 2-7a (new) 3/4 2-8 to y4 2-7a 2-8 (old) 19* 3/4.2.2.2 (new) Insert new Page y4 2-8 page y 4 2-8 (new) after Page y 4 2-7a 20* 3 2.3 Change calculated value of F,.T from page y 4 2-9 1.620 to 1.650, insert Action Item (b) and change Figure 3.2-3 to 3 2-3c 215 Figure 3.2-3c Insert new Figure 3 2-3c after page 3/4 2-10a (new) Page 3/4 2-10 22 Figure 3 2-4 Replace Figure 3 2-4 with Figure 3 2-4 page 3/4 211 23' 3 2.5 Add ", Core Power" to Item (d) page F4 2-13 24* Table 3 2-1 Change minimum pressurizer pressure from page y 4 2-14 2225 psia to 2200 psia, change AXIAL SHAPE INDEX limit from Figure 3 2-4 to
"***" and add footnote for "***"
25 Table 3 3-1 Change Steam Generator Low Pressure Trip page 3/4 3-4 Bypass limit from 685 psia to 710 psia 26 Table 3 3-2 Change RTD response time from 8.0 page y 4 3-6 seconds to 12.0 seconds 27 Table 3 3-3 Change Safety Injection (SIAS) page y 4 3-15 Pressurizer Low Pressure Trip Bypass limit from 1700 psia to 1800 psia 28 Table 3 3-3 Change Main Steam Line Isolation (SGIS) page 3/4 3-15 Steam Generator Low Pressure Trip Bypass limit frm 685 psia to 710 psia i I l
P Table 9-1 (continued) Change Tech Spec # Action 29 Table 3 3-4 Change Safety Injection (SIAS) page 3/4 3-17 Pressurizer Low Pressure Trip setting from 1578 psia to 1725 psia 30 Table 3 3-4 Change Main Steam Line Isolation page 3/4 3-17 (SGIS) Steam Generator Low Pressure Trip setting from 570 psia to 635 psia 31* 3/4.4.4 Change the description and limits of the page 3/4 4-5 pressurizer level operating band 32* B 3/4.1.1.1 and Change EOC shutdown margin, T B 3/4.1.1.2 >200F,from4.3%ak/ktoE.3yh/k, page B 3/4 1-1 and change BOC shutdown requirement to 4.5% ak/k 33 B 3/4.2 5 Change minimum DNBR of 1.195 to minimum page B 3/4 2-2 DNBR of 1.23 34* B 3/4.4.4 Change the description concerning page B 3/4 4-2 pressurizer level 35" 531 Increase limit on enrichment in . page 5-4 description of reload fuel assemblies from 3 7 w/o U-235 to 4.1 w/o U-235 I
*Added or modified request relative to Reference 1 or 2. Explanation for addition or modification is contained in Table 9-3
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Table 9-2 Explanations for Cycle 5 Tech Spec Changes Change Tech Spec d Explanation 1 Figure 2.1-1 Thermal Limit Lines have been changed to reflect higher radial peaking factors and the implementation of margin recovery programs. 2 Table 2.2-1 The Steam Generator Low Pressure Trip setting is being raised to accommodate the larger uncertainties associated with the new pressure transmitters. 3 Table 2.2-1 he Steam Generator Low Pressure Trip Bypass limit is being raised to reflect the change in the trip setting (See Change No. 2) 4 Figure 2.2-1 The LHR LSSS has been changed to reflect higher radial peaking factors and the implementation of margin recovery programs 55 B.2.1.1 LHGR to centerline melt is being raised to increase operating margins and flexibility 6 B.2.1.1, he minimum DNBR has been increased to B.2.2.1 1.23 to be consistent with Statistical Combination of Uncertainties 7 B.2.1.1, Statistical Combination of Uncertainties B.2.2.1 has removed the 2% power uncertainty from the transient analyses 8 B.2.2.1 See Change No. 2 9 B.2.2.1 ne minimum trip setting (floor) of the Thermal Margin / Low Pressure (WLP) Trip is being raised to acccanoda*e the larger uncertainties associated with the new pressurizer pressure transmitters. 10 B.2.2.1 The W LP basis has been adjusted to be consistent with Statistical Combination of Uncertainties (SCU) and the bias has been changed as a result of the imple-mentation of SCU and the recatagor-ization of CEAW (A conservative bias value relative to the Transient Analysis results has been incorporated)
Table 9-2 (continued) Change Tech Spec # Explanation 11' 3/4.1.1.1 The shutdown margin has been increased to yield acceptable results for the EOC HZP SLB event. 12# 4.1.1.4.2 The surveillance requirements on MTC are being modified to allow the use of MTC determinations made during power ascension startup .neasurements for the purposa of satisfying surveillance requirements. This change is consistent with the objective of assuring that the most positive MTC at power conditions, which occurs at the highest - boron con-1 centration, meets Tech Spec. 83 1.1.4.b. 13 Figure 3 1-2 The PDIL.is being changed to increase available SCRAM worth and produce accep-table results for the EOC HZP SLB event; the allowable BASSS operating region-is being indicated. 14 4.2.1 3 A new figure is being added to take credit, in terms of maximum allowable fraction of RATED THERPp POWER, for the calculated value of F when moni-toringtheLHRLCOwilEtheex-core detector system. This change (#14) and Change Nos.15 through 19 are being made to avoid unnecessary power level changes resulting from temporary on-line computer outages. 15 Figure 3 2-2 The LHR LCO is being changed as a result-of higher radial peaks, the implemen-- tation of margin recovery programs, and - the addition'of Figure 3 2-3b to take crept for the ' calculated value of F when monitoring the LHR LCO wlEhtheexcoredetectorsystem. 16* Figure 3 2-3b Figu e 3 2-3b is being added to take-cregit for the calculated value of F". when monitoring the LHR LCO wlEhtheexcoredetectcrsystem. 17' 3/4.2.2.1 The plenar radial peaking factor, FxyT, is being raised for Cycle 5 to increase operating margins and flexibility, and the Tech Spec and figure numbers are being changed to facilitate other Tech Spec changes which are being made to take credit for the calculated value of- Fxy T when monitoring the LHR LCO with the ex-core detector system. '
Table 9-2 (continued) Change Tech Spec # Explanation 18' Figure 3 2-3a The planar radial peaking factor, F T is being raised for Cycle 5 to inEre,ase opegating margins and flexibil-ity, the P , curve is being separated !?cm the F ' cupe to datedifferenEFxy acc qF[ and Fx imits, values, based uponancex-core moEitoring of the LHR LCO, are being separated from this figure to facilitate other Tech Spec changes which are being made to take T redit for the calculated value of F when monitoring the LHRLCOwiEEtheexcoredetectorsystem. 19' 3/42.2.2 A new Tech Spec is being added to take credit for the calculated value of F T when monitoring the LHR LCO with theXY ex-core detector system. 20' 323 Thgintegratedradialpeakingfactor, Fr , is being raised for Cycle 5 to increase operating margins and flexibil-it BASSS is being implemented, and the F,.{,curge is being separated from the F diffeknt Fcurvgtoacemmodate and F i values. xy r 21* Figure 3 2-3c Thgintegratedradialpeakingfactor, Fr , is being raised for Cycle 5 to increase operating mar flexibility, and the F,. gins and is curve being separated from the Fxy', curve to.,,accomodate different Exy' and Fr ' values. 22 Figure 3 2-4 The DNB LCO has been changed to reflect higher radial per. king factors and the implementation of margin recovery programs. i 23* 3 2.5 The phrase " Core Power" is being added to reflect the implementation of BASSS 24' Table 3 2-1 The minimum steady state pressurizer pressure has been lowered to increase operating flexibility and the ASI limits have been modified for the implementa-tion of BASSS 25 Table 3 3-1 See Change Ib. 3
Table 9-2 (continued) Change Tech Spec # Explanation 26 Table 3.3-2 he RTD delay time used in the Cycle 5 '
; analysis has been increased from 8 to 12 seconds to increase the acceptance test criterta 27 Table 3 3-3 The Safety Injection (SIAS) Pressurizer Low Pressure Trip Bypass limit is being raised to reflect the change in the trip setting (See Change No. 29) 28 Table 3 3-3 The Main Steam Line Isolation (SGIS).
Steam Generator Low Pressure Trip Bypass limit is being raised to reflect the change in the trip setting (See Change No. 30) 29 Table 3 3-4 Be Safety Injection (SIAS) Pressurizer Low Pressure Trip setting is being raised to acconmodate the larger uncer-tainties associated with the new pressure transmitters. 30 Table 3 3-4 The Main Steam Line Isolation (SGIS) Steam Generator Lori Pressure Trip setting is being raised to acconunodate the larger uncertainties associated with the new pressure transmitters. 31' 3/4.4.4 A relaxed pressurizer level operating band is being incorporated to improve operating flexibility. 32' B 3/4.1.1.1 and The shutdown margin has been increased. B 3/4.1.1.2 to make it consistent with specification 3/4.1.1.1 ! 33 B 3/4.2.5 The minimum DNBR has been increased to be consistent with Tech Spec B.2.1.1 and B.2.2.1. l 34' B 3/4.4.4 The . bases for. the pressurizer level
- operating band are being modified to support the expanded limits.
l l 35' 5.3 1 The specification of the enrichment limit in reload fuel assemblies has been increased to permit the use of higher i enrichment. fuel.
#Added or modified request relative to Reference - 1 or 2. Explanation . for addition or modification is contained in Table 9-3 l
l
Table 9-3 Explanation for Changes Relative to Rose Reouested for Unit 1 Cycle 6 Change Explanation 5 The centerline melt limit is being increased to 22.0 kw/ft, compared to the Unit 1 Cycle 6 value of 213 kw/ft, to increase operating margins and flexibility.
- 11 The Unit 2 Cycle 5 analyses support a shutdown margin of 5.2% ak/k, compared to the Unit 1 Cycle 6 value of 5 3%
a k/k. 12 The MTC surveillance requirements are being modified, compared to no change for Unit 1 Cycle 6, to allow the use of MTC determinations made during power ascension startup measurements for the purpose of satisfying surveillance requirements. Bis change is consistent with the objective of assuring that the most positive MTC at power conditions, which occurs at the highest boron concentration, meets Tech Spec #3 1.1.4.b. 16 Figye 3 2-3b is being modified to be consistent with an F value of 1.70, compared to the Unit 1 Cycle 6 FXYT value of 1.65; the requested page number has been cNEnged to be consistent with the issued page number in the Unit 1 Cycle 6 SER (Reference 3). 17 The Fx T is being increased to 1 70, compared to the Unit 1 kycle 6 value of 1.65,.to increase operating margins and flexibility. 18 Figye 3 2-3a is being modified to be consistent with an . F x value of 1.70, compared to the pnit 1 Cycle 6 va [ue of 1.65, and to separate the F curve from the F' curve; the requested page number *7has been changed to be consistent with the issued page number in the Unit 1-Cycle 6 SER (Reference 3). 19 Requested page number has been changed to be consistent with the issued page number in the Unit 1 Cycle 6 SER (Reference 3). 20 Figure 3.2-3 is being changed to Figure 3 2-3c, compared to Figure 3 2-3a for pit 1 Cycle 6, to facilitate the separation of the F xy and F T r curves. 21 Figure 3.2-3cp is being added t facilitate the separation of the F and F T , curves between tN . Unit 2 Cycle 5 due Fx T toand the 'Fpifference values, l r compared to Unit 1 Cycle 6 which grduped both parameters into one curve on a single figure due to their similarity.
. . i l
l Table 9-3 (continued) Explanation for Changes Relative to Those Requested for Unit 1 Cycle 6 Change Explanation 23 The Technical Specification No. and page number are being changed to be consistent with existing Technical Specification (Unit 1 Cycle 6 T.S. No. and page number were out of date in request). 24 The minimum pressurizer pressure is being decreased from 2225 psia to 2200 psia, compared to no change for Unit 1 Cycle 6, to improve operating flexibility; the page number has been changed to be consistent with existing Technical Specifications (Unit 1 Cycle 6 page number was 01t of date in request). 31 A relaxed pressurizer level operating band it being incorporated, compared to no change for Unit 1 Cycle 6, to improve operating flexibility. 32 Same as No. 11 34 The bases for the pressurizer level operating band are
' being modified, compared to no change for Unit 1 Cycle 6, to support the expanded limits.
35 The Unit 2 Cycle 5 request has been modified, relative to the Unit 1 Cycle 6 written request, to conform to the verbally agreed upon and subsequently issued change for Unit 1 Cycle 6 (Reference 3). The initial written request asked for the removal of the enrichment limit; the agreed' upon change simply raised the enrichment limit from 3 7 w/o U-235 to 4.1 w/o U-235 I
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. Q O -
g LAJU E 2.2-1 (Cont'd) . I REACTOR ,PROTE;CTIVE,. lilSIRUHC,4IATION IRIP SETP0lNT LlH,l]S,, , w- ,
.i n
C _ FUNCTIONAL UNIT.
.T.R.I,P. SI:! Pol HT A'LLOWA8LE VALUES 'l
,4. Pressurizer- Pressure - liigh 1 2400 psig i 2400 psja l; E' 5.
- Containment Pressure - liigh h 4 psig < 4 psjg -
- 6. Steam Generator Pressure - Low (2) la , -
a l 1
- / ~)
~7. Steam Generator Water Level - Low > 10 inches below top -
> 10 inches beloy top iif feed ring. ,
of feed ring.
- 8. Axial flux offset (3) Trip setpoint adjusted 69 Trip setPoint adjusted to j
. not exceed the limit )lnes not exceed the )(mit lines ~ l of figure 2.2-1. of figure 2.2-1.
m ' '
. l9. Thermal Margin / Low Pressure (1)
~
, a; four Reactor Coolant Pumps Trip setpolet adjusted to': Trip setpoint adjusted to I
. Operating not exceed the limit }l pes. pot exceed the. limit lines t of figures 2.2-2 and 2.2-1 of figureg 2.2-2 and 2.2-1 h ., Steam Generator Pressure < 135 psid 1 135 psjd DiIference - liigh (1) k 10. Loss of Turhine -- llydraulic > 1100 psig 1 100 1 psig S
a , fluid Pressure - Low (3) 5a 11. Rate of Change of Power - liigh (4) 1 2.6 decades per minute l 2.6 decades per minute r, g . .
. TABLE NOTAfl0N
-4 '
(1) Tripmaybebypassedbepw10 1 of RATED TilEltHAL POWER; bypass Iha() be automatlcally removed when Til[RHAL POWER is > 10 1 of RATED TilERilAL P0utR. 3
?
8
' S e .
4
. 4 9 .
i ~, G m - TABLE 2.2-] (Cont'd) - l . . x [ ' TABLE NOTATIONS (Cont'd) gg-r- 930 ' M (2) Trip may be manually bypassed below 685 psla; bypass shall be automatically removed at or above 85 psl4 ul ! i
. (3) Trip may be bypassed below 15% of RATED 111ERilAL POWER; bypass shall' be automatically removed when !
! c: TilERilAL POWER is > 15% of HAIED TilEHilAL POW [R.
- =
~4 (4) Trip may be bypassed below 10 ?. a~nd above 12% of HATED IllERilAL POWER. i, ,
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0 0.2 0.4 0.4 - 4.6 0.4 0.2 PERIPMERA1. AXIAL. SHAPE INDEX.Y, FIGURE 2.21 Peripheral Axial Shape index. Y, Versus Fraction of RATED THERMAL POWER . .
..*.me"-
- 2-11 Amendment No.
CALVERT CUFFS - UNIT r 6.q.nq e,4g,e,=- ,p_ _ ,
-eh . v. 7- - - . -w . ,.-.w-... -e . - - + - - - - -.- - , - - - . = - - - , - - - - - - .ea - - - - - - - - - . - - -- -
Pf T T
.a.._.- ____.. .. . .
2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE g The restrictions this safety limit prevent overheating of the
. fuel cladding and.po ible cladding perforation which would result in the ,
release of fission . cducts.to the reactor coolant. Overheating of the fuel- is prevent maintaining the steady state peak linear heat rate at or less than 1 kw/ft. Centerline fuel melting will not occur < for this peak linear heat rate. Overheating of the fuel cladding is - prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer ccefficient is large and the cladding surface temperatu're is slightly above .the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime
~could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer ccefficient. DNB is not a directly measurable parameter during coeration and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been relat-d to ONS through the CE-1 correlation.
The CE-1 ONS correlation has been teveloped to predict the ONS flux and the location of CNS for axially uniform and non-uniform heat fiux distri-I' butions. The local DNB heat flux ratic, DNBR, defined as tne ration of
- the heat flux that would cause ONB at a particular core location to the local heat flux, is indicative of the margin to ONS.
The minimum value of the ONBR during steady state operation, no al operational transients, and anticipated transients is limited to 1.195 l'13l This value corresponds to a g5 percent pecbability at a 95 percent cen-fidence level that DNB will not occur and is chosen as an appropriate margin to INB for all cperating conditions. The curves of Figures 2.1-1, 2.yAand 2, 2.1-3 t.23 2.1-4 show the loci of points of THERMAL POWER, Re ctor Cociant System pressure and maximum cold leg temperature af y ious pump combinations for which the - minimum DN8R is no less than 0.195 for the family of' axial shapes and l corresponding radial peaks shown in Figure 82.1-1. The limits in Figures 2.1-1, 2.1-2, 2.1-3 and.2.1-4 w'ere calculated for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant inlet temperature is not a safety limit; however, operaticn above 580'F is not possible because of the actuation of the main steam line. safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than 12t of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in
*) t k3 5b -
3 2-1 Amendment No. 78, 31 i
'. CALVERT CLIFFS - UNIT 2 e
-3 - . - - - , + ,ev. .- # v , , , . - , - -v----we--rw- a,- .
p:3 SAFETY LIMITS BASES . Table 2.1-1. The area of safe operation is belcw and to the left of these lines.
- The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shewn en the figures.
The reactor protective system in ecmbination wit'h the Lim's ting Conditions for Operation, is designed :s prevent any anticipa ced comoina-tien of transient condi:icns for reactor coolant system temparature. pressure. and THERMAL POWER level that would result in a CN3R of less than 1.195 and creclude the existence of flow instabilities. l
)1.23 2.1.2 REACTOR CCOLANT SYSTEM PRES 5URE "le restriction of this Safety Limit prc:ects the integrity of the Reac:cr Coolant System frem overpressurizatien anc thereby prevents tne release of radionuclides c:ntained in the reactor coolant from reaching ./*5 the c:ntainmen a tmosphere.
The reactor pressure vessel and pressuri:er are designed to Section
' III,1967 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Ccolant System piping, valves and fittings, are cesigned to ANSI B 3. 7, Class I,1969 Edition, whien permits a maximum transient pressure of 110% (2750 psia) of ecm;onen: design pressure.
The Safety Limit of 2750 psia is therefore c:nsistent with the design i criteria and associated c0ce requirements. The entire React:r Coolant System is hydretested at 3125 psia to j demonstrate integrity prior to initial cperation. l i - I t l l l
.9 CALVERT CLIFFS - UNIT 2 3 2-3. Amencment No. 78,31 e
- . --- w w -- , vwe,
I . . . , . , 1 2.2 LIMITING SAFETY SYSTEM SETTINGS f
. BASES 2 . 2.1 P.IACfCi TRIP SETPOINTS .
The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Satpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its speci-fied Allowable Value is acceptable en the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the' automatic protective instrumentation channels and provides manual rer: tor trip capability, f
- Power Level-High The Power Level-High trip provides reactor core protection against ~
. reactivity excursions 'which are too rapid'to be protected by a Pressurizer Pressure-High or Thermal Margin / Low Pressure trip. .
The Power Laval-High trip setpoint is operator adjust'able and can be set no higher than 10% above the indicated THERMAL POWER level . Operator action is required to increase the trip setpoint as THERMAL POWER is l increased. The trip setpoint is automatically decreased as THERMAL power decreases. The trip setpoint has a maximum value of 107.0% of RATED THERMAL POWER and a minimum setpoint of 30% of RATED THERMAL POWER. Adding to this maximum.value the possible variation in trip -state. point'due to calibration and instrument errors, the maximum actual ste THERMAL POWER level at which a trip would be actuated is 12 of RATED THERMAL POWER, which is the value used in the safety analyses. 190 *)$ Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core. protection to prevent DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to permit 82-4 - Amendment No. 18 CALVERT CLIFFS-UNIT 2 O
LIMITING SAFETY SYSTDi Sei(INGS BASES
$, 1.23 operation of the reactor at reduced power if one or reactor coolant pumps are taken out of service. The low-flow trip etpoints and Allowable .'
Values for the various. reactor coolant pump combin tions have been -
~
derived in consideration of instrument errors response times of equip:nent involved to maintain the DNBR above 1.19 under normal operation I and expected transients. For reactor operation with only two or three ' reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the power Level-High trip setpoints, and the Themal Margin /Lew
. pressure trip setpoints are automatically changed when ti.e pump condition selector switch is manually set to the desired two- or three-pump )gy position. Changing these trip setpoints during two and. three p operation prevents the minimum value of DNBR from going below 1.195 during l nomal operational transients and anticipated transients when on two or three reactor coolant pumps are operating.
pressurizer pressure-Hich The pressurizer pressure-High trip, backed up by the pressurizer code i' safety valves and mai.n steam line safety valves, provides reactor coolant h system protecticn against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the ! undesirable operation of the pressurizer code safety valves. Containment Pressure-High The Containment pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint . for this trip is identical to the safety injection setpoint. . Steam Generator pressure-Low- - The Steam L.nerator pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and 4 05 subsequent cooldewn of the reactor coolant. The setting of(510) 0 psia l
.is sufficiently below the full-load operating point of 850 psia so .
as not to interfere with normal operation, but still high enough to provide the required protection in the event of , excessively hi steam M " _ $ ! $ *!_ng was used with an un' certainty factor of + psi; EhEEE $$Ebhc6n StacmLino Brock Event. N C CALVERT CLIFFS - UNIT 2 8 2-5 Amendment No. 7),31 k .. e
i LIMITING SAFETY SYSTEM SETTINGS _ BASES e Steam Generator Water Level . The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures
- that the pressure of the reactor ecolant system will not exceed its Safety Limit. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide a margin of more than ,13 minutes before auxiliary feedwater is required.
Axial Flux Offset f,g 3 The axial flux offset trip is provid to ensure that excessive
-axial peaking will not cause fuel damage. The axial flux offset is determined from the axially split excore ectors. The trip setpoints .
ensure that neither a DNBR of less than 1.19 nor a peak linear heat rate which corresponds to the temperature for ue centerline melting will , exist as a consequence of axial power maldistributions. These trip set-points were derived from an analysis of many axial pcwer shapes with ' allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incere axial flux offset relationship. i Therw.a1 Margin / Low pressure f,23 The Thermal Margin / Low P ure trip is provided to prevent operation when the DNBR is less than .195
/975" .
The trip is initiated n er the reactor coolant system pressure signal drops below either 75 psia or a computed value as described below, whichever is higher, she computed value is a function of the i higher of AT power or neutron power, reactor inlet temperature, and the . number of reactor coolant pumps operating. The minimum value of reactor l ' coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA l deviation permitted for continuous operation are assumed in the genera-tion of .this trip function. In addition, CEA group sequencing in accor-Finally, the dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. maximum insertion of CEA banks which can occur during any anticipated l operational occurrence prior to a Power Level-High trip is assumed. B 2-6 Amendment No.78,31 CALVERT CLIFFS - UNIT 2 9
..y._.- _
LIMITING SAFETY SYSTEM SETTJNG_S,
$ BASES i
IW.u. Lost
- The Thermal Margin / Low prwssure trip se points er; d; rind f.e . =;
sr; ;;f;t) li;it; Or;.p. ;;;ii;;ti= ;f ; c;;riet; allowances .for-equipment response time, measurement uncer* Anfe;y.e.gir.i;pr;.-ided.J.iet, incl. des.fntiestedprocessingerrerg en eia.e e..ee ef % ef '
- 7JO T.O".L ".JC ts.;; ;-ensete fer ;;;; gi ;! ;n:;r ;;n;.. r.: ;...r; .
en elis er.;e af 2*f :: ;= ;nnt; f;r ;"t9,i i t_ ;;r;;.r; nn;.. ;nt
;.aee.;ai..;),' and a further allowance afMps;ia to compensate for h 1 prus;.r; ;;ni.r nt e. . e. , tri67 prun;ir.; e. . ;r, ene time delay associated with providing effective termination of the occur ence that exhibits the most rapid decrease in margin to the s4fety limit. The-gg l'
.;;i; elle h;; i; ;-d; u; ef ; . ,.. ..
- y. . .. . -- . . -- - . . 11 :2*ee
= ? ; 70 p;ie ti;; del;j ella ;ne. . l.
Asynnetric Steam Generator Transient Protection Trio Function (ASGTpTF) The ASGTPTF utilizes steam generator pressure inputs to the TM/LP ) calculator, which causes a reactor trip when the difference in pressure between the two steam generators exceeds the trip setpoint. The ASGTPTF is designed to provide a reactor trip for those Anticipated Operational. Occurrences associated with secondary system malfunctions which result I' in asymetric primary loop coolant temperatures. The most ifmiting event ' [J( '. is the loss of load to one steam generator caused by a single Main Steam . Isolation Valve closure. The eiaufpcent trip setpoint and allowable values are calculated to account for instrument uncertainties, and will ensure a trip at or before reaching the analysis setpoint. , Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus' extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability !.. at the specified trip setting is required to en. hance the overall reliability of the Reactor Protectinn System. - Rate of Change of Power High . The Rate of Change of Power-High trlp is provided to protect the core during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit. Its trip setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for , ( '. C operation of this trip. Its functional capability at .the specified trip l i setting is required to enhance the overall reliability of the Reactor i l Protection System. CALVERT CLIFFS '- UNIT 2 B,2-7 Amendment No.f. 78. 31 i .
~
- t 3/4.1 REACTIVITY CONTROL SYSTEMS -- . . w , . - . .
^ " ~ ^ ~
, 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,yg > 200*F
, ,_ [ ,, ,,
.2
- ~ , _ .. .. _ .. ~ _ .... _ . . , . _ , .
LIMITING CONDITION FOR OPERATION 5.R% " 3..l.1.1 The SHUTDOWN MARGIN shall be > h >Ak/k. ,
. APPLICABILITY: MODES 1, 2'", 3 and 4. -
.. T-ACTION:
With the SHUTDOWN MARGIN < ak k anediately initiate and continue boration at > 40 gpm of 2300 ppm boric acid solution or equivalent until
~
the required SHUTDOWN MARGIN is restored .
- - u
' ' ~
SURVEILLANCE REQUIREMENTS 25390 4'.1.1.1.1 The SHUTDOWN MARGIN shalT be detennined to be y,4.3* ak/k: l
- a. Within one hour after detection of an inoperable CEA(s) and at -'
least once per 12 hours thereafter while the CEA(s) is inoperable. . If the inoperable CEA is immovable or untrippable, the above d ' required SHUTOOWN MARGIN shall be increased by an amount at least e
~
CEA(s) qual to the withdrawn worth of the innovable or untrisipable
, b. When in MODES 1 or 2*, at least once per 12 hours by verifying-that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6..
- c. When in MODE 2 ", within 4 hours prior to achieving reactor %
criticality by verifying that the predicted critical CEA # position is within the limits of Specification 3.1.3.6. -
- d. Prior to initial operation above 5:; RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.
o.
- Adherence to Technical Specification 3.1.3.6 as specified in Surveillance l Requirements a.l.l.1.1 assures that there is sufficient available shut-
' ' down margin to match the shutdown margin requirements of the s.fety analyses. .- .
" See Special Test Exceptica 3.10.3. 1 t With K,ff >_l.0. g 4 With X,ff < l.0. ~
l[ ... . . ,. CALVERT CLIFFS - UNIT 2 3/4 1-1 Amendment No. J. JS,3I l l i
REACTIVITY CONTROL SYSTEMS 3
)
SURVEILLANCE REOUIRE"ENTS (Continued) 4.1.1.4.2 The MTC shall be detarmined at the following frequencies and THERMAL POWER conditions during each fuel cycle:
- a. Prior to initial operation above 5% of RATED THERMAL POWE,
"**j'h.f,"'ch,*eh/ATOf#8dMALN8A>
- b. At any THGMAL POWSi within 7 EFF0 aftar c::r' ; . %TL
C"."'. 'O'I'. .;;i' i tr' c i;. ;; ;;;;;.=f,J ;r ;f :00 ;;p.
- c. At any THERMAL PCWG, within 7 EFF0 af er reaching a RATED THERMAL POWS equilibrium boren cene tratien of 300 ppm.
/,'Is rs um sn hially rem.c.kan en egw[%
' ' o$ .
cond s $lon e.h or o.bov t 90 RATE) THERMA L Pow E4.
/
)
, CALVGT CLIFFS-UNIT 2 3/4 1-6
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%e mg.pw h WR __-_ W r 6*N NM".'*P 6 W=M*'WD W
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#'N*.,,"*8", ' *
- _ _ " ,*6 .. , _ . . ._._ .__- ___
POWER DISTRIBUTION LIMITS --
- u . . - :
SURVEILLANCE REOUIREMENTS (Continued)
' ~ ' ~ - ' _ 'c. ~ , Vertfying at least once Ur~ 31 days that the AXIAL SHAPE INDEf is maintained within the limits of Figure 3.2-2, where- 100 - - -
percent of the allowable power represents the maximum THERMAL POWER allowed by the following expression: , MxN . where: . .
- 1. M is the maximum allowable THERMAL POWER ~Tevel for the existing Peactor Coolant Pump combination. _ , L
- 2. N is the maximum allowable fraction of RATED TH P AL POWER as determined by the F*'Y curve shown up Figure 3.2- of -
Specification 3.2.2. ,
>3,3, 3b 4.2.1.4 Incore Detector Monitorino System - The incere detector moni-toring system may be useo for monitoring tne core power distribution by -
verifying that the intore detector Local Power Density alarms: .. g]. r a.--- Are adjusted to satisfy the requirements of the. core-co%er.---- :_.- -- .
- ~ - -
distribution map which shall be updated at leaH once pef 3T - days of accumulated operation in MODE 1. ,,
- b. Have their alann setpoint adjusted to less than or equal 'to the limi-ts shown en Figure 3.2-1 when the following factors are appropriately included in the settino of these alarms: ...
- 1. Flux peaking augmentation facters as shown in Figure .
4.2-1,
- 2. A measurement-calculational uncertainty factor of 1.07, [,
- 3. An engineering uncertainty factor of 1.03, .
. 4. A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and thennal expansion, and
- 5. A THERMAL POWER measurement uncertainty fact [oY 1.02. -
y .
] .
h) CALVERT CLIFFS-UN!T 2 . 3/4 2-2 Amendment No. 5, 9, 76, 78,9.4 3
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4 . . , TO P NA AL NG FACTOR Fh LIMITING CONDITION FOR OPERATION heca1tedvalueof(y,definedasFfy=Fxy(1+Tq ), shall be . limited to < ,
. l APPLICABILITY: MODE 1*. -
N y
> in 6 hours either: ' y _g
. a. Reduca THERMAL POWER to bring the canb' n of THERMAL POWER ,
to within the limits of Figure and withdraw the and full iF'lngth CEAs to or beyond the Long Tenn Steady State Insertion Limits of Specification 3.1.3.6; or
- b. Se in at least HOT STANOBY.
SURVEILLANCE REQUIREMENTS The provisions of'Specificatidn 4.0.4 are not applicable. L.2. . . 4.2.2. F' y shallbecalculatedbytheexpressionFjy=F,y(1+T)andFjy q shall be determined to be within its limit at the following intervals: l a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, i b. At least once per 31 days of accumulated operation in MODE *- 1, l and . l l c. Within four Hours if the AZIMUTHAL POWER TILT (T q ) is > 0.030.
'See Special Test Exception 3.10.2.
- r. .
.4 l 1 f
l CALykRT CLIFFS - UNIT 2 I 3/4 2-6 Amendment No.9, 78,31 .
] -
i . . l
, POWER'0ISTM3UTICN LIMITS .
(, . . , SURVEILLANCE ?,ECUIREMENTS (Continued) t.3 - 4.2.2.'3, F shall be determined each time a calculation of F'xy is required xy . by *Jsing the incere detectors to obtain a pcwer distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This detemina-tien shall be limited to core planes bet.een 157, and 85% of full core height inclusive and shall exclude reg' ions influenced by grid effects.
*t h.2.2.hT.% q shall be determined each time a calculation of F is required
.and the value of Tq used to determine F'y x shall be the measured value of
- i. .
q
~
r . ( T A
.t. .. .
lI l . l CALVERT CLIFF 5-UNIT 1 , hendment M3. 27,'32 CALVERT CLIFF 5-UNIT 2 . 3/42-7. /mendmen: No. 3. IS i . l . . . .
~
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POWER DISTRIBUTION LIMITS , T C TOTAL PLANAR RADIAL PEAKING FACTOR *F xy LIMITING CONDITION FOR OPERATION
.2 T
Th;
._.,een
- 10:10t:d ^ :h: Off,,,cafin:d::
x F xy _ 7 xy ,7m7 q
), ;h 7; 3; ,
jjmit_2 ! Tna, veduc o$ N presen&lg u. sed. 6i spedsdion 4 213 APPLICABILITY: MODE IM DC- " CC" E " E M EPN 3 ACTION: Nith."T -
~ XY ' '52", within 6 hours either$ho./ coba. o% W preszn41s umd ih specSc.hhon a.
- d;= .m3
- nd P' :=. == =th;wios ii=it: =e c 2ioe:ica3.2-2 :nd af :=. th=[-
- within Of Fig;r- withdr; ful' Ength CE.'.: t: Or 5:y:nd th: L:ng T:r St::dy S*:t: lim 4 &qoum +h2.in In::rtica Limits Of 5;;ecific&tica 2.1.0.5, er %n M-3b,
- b. Be in at least HOT STANOBY.
Ocdu.qc. iba. m\co. cyl uod.m
,$p,_e,gica,yee 4 a.g,340 gg;%
s the. %Rs of Sciu.re. 3Q-3b;or SURVEILLANCE REQUIREMENTS g3.1 h2.2.1 The J, D .Q. orovisions of'Specificatidn 4.0.4 are not applicable. y k.2. 2.'lL F'y x shall be calculated by the expression xF'y = Fxy(1+T)and[xy q shall be detemined to be within its limit at the f.ollowing intapals: ibg mont+cr inc3 F;;y
- a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,
. 3
- b. At least once per JT days of accumulated operation in MODE 1 - -
and
. , , ,. r . 3 ,_ _
m.-.
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- u. .. .. - ....
s.q, . . . . ~ . Cee Sp cia' Te;t Ocegti:n .'0.2. l CALVtRT CLIFFS - UNIT 2 3/," 2-0 Amendment No.3, 78,3I b~
-- -, n - ,,, ,
_ ._ .. . . . . . . . _ _ . _ _ . ..__ -- m. .,
.g . PCWER'0!STk!3UTICN LIMIT'S - % I SURVEILLANCE ?.EOUIREMENTS (Continued) 2.2 _
4.2'.2. F xy shall be determined each time a calculation of F[y is required by using the incere detectors to obtain a power distribution map wien alt full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pur.p ccmbination. This determina-tion shall be limited to core planes betaeen 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects. T shall be determined each time a calculation of F,I is required and the value of Tg used to determina Fjyshall be the measured value of
- q. .
'L ' ,, , ,
2 a ~ r o, "_f CALVERT CLIFFS-UNIT 1 . Amendment No. 27, 32 CALVERT CLIFFS-UNIT. 2 / ? ;-7 Amendment No . 3,18 3M a-t (ced.)
' ^ ~
- .: -. . L.., . . . . . . _ . .. . .. . . ~ ~i
~ '
,- POWER OISTRIBUTION LIMITS .-- --
. .'i1.i._ '-..'...i._2_.1
.._. 7 i .l?_ '
T0TAL INTEGRATED RADIAL PEAXING FACTOR - F7 ,
- - . ~
N .- . LIMITING CONDITION FOR OPERATION 3.2.3 The e edvalueofFf~,definedasFf-FII*I),shal[be r q limited to 1 62 g. k %de'THERHE power to 04-- l nbicchen of TtWRHE POWE odd hPPLICABILITY: MODE 1*. cor ye % WW % u,.nns of Fiqm3.Q-3c.f-
- o ACTION:
Mcde ce be. yond Umd5 tho. Lon of S *gTe.rm , P Cib:ANO With FT> r - within 6 hours either: 3.1.'f ;dnd'inM ndw WM i [ of- in GASSSj oc -
- a. or Be in at least HO'T STAN08Y , M 5 3.2-3c.
of THERMAL POWER C% Reducy THERMAL POWER to bring the combinati and F to within the limits of Figure and w'ithdraw the full lengt5 CEAs to or beyond the Long Term Steady State Insertion Limits of Specification- 1 3.6. The THERMAL POWER limit . detennined from Figure .2- shall then be used to establish a revised upper THERMAL POW R 1 vel limit on Figure 3.2-4 (truncate Figure 3.2-4 at the a le action of RATED THERMAL POWER a detennined by Figure .2- and ubsequent operation shall be OJ maintained within the reduced ac ptable operation region of Figure 3.2-4. v ~ 3 ,q. y 3.3-3c. . SURVEILLANCE REQUIREMENTS
. 2. 3 .1 The provisions of Specification 4.014 are not applicable. -
%.2.3.2 ., Ff shall be calculated by the expression Ffr = F q(1+T ) and Fh shall be decennined to be.within its limit at the following intervals:
- a. Prior to operation above' 7.0 percent of RATED THERMAL POWER
, c , after each fuel loading,
- b. At. least once per 31 days of accumulated operation in MODE 1, and
- c. Within four hours if the AZIMUTHAL POWER TILT. q(T ) is > 0.030.
'See special Test Exception 3.10.2.
C
" 3/4 2-9 Amendment No. 9, 76, 78, 31 qALVERT CLIFF 5 - UNIT 2 l
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0 _ . . . _ . . . _. ..... . _._. . _ .... . ._ _ ._. . . . . _ _.. .. .. . . _ . . _ . . a. s _ _. . .. . . . 1 . . . . . . . _ i .......... ......
. . . . . . . . . = . .. .. .. .... . _.. _.. . -. ..
_ _ . . .........._... _ .. _......:.. ..._._...... .... .... . ... . .. _. _. .. . .. ..._. ._._. ..... .. ...........__... ..... :s... ........._._... _.. ..... .._. .
. . . . . . . . . . . ~
. s.
.....s.......
.. .. .. .. . . . . . . . . . . . _ . ,.............s,
._._.._......m.
0A 0.2 0.4 OA
-0.2 0 1 -0.5 '
-0.4
- l MRIPHERAL AX1AI. SHAM lf:DEX (Yy .
FIGurtE 3.2-4
- DNB Axial Flux Offset Control Limits _
- t. d) - .
) ~
~
\,, Amendment tio.
1 ! l CAI.YERT CLIFFS-U!ilT 3/4 2-11 b
~,
z::
- =.--
L' ?$ ..
- E-i~
R , 2W;
.- :.==
~
POWER OISTRIBUTION LIMITS y . 3 -
.E;5
.=
~
DNS PARAMETERS e, : ?? Z1
===
- e- . i c. ';
; ; : .: ~ > - 31 c.+. i LIMITING CONDITION FOR OPERATION ==
u : .=.:
* .M=-
3.2.5- The following CNB related parameters shall be maintained within the. limits shown on Table 3.2-1: E
. "_i.
M.~-
- a. Cold Leg Temperature 7.#
=.+ .
b Pressurizar Pressure - EE
.g
- e. Reactor- Coolant System Total Flow Rata
~fE
~
h
-m
'd . AXIAL SHAPE INDEX j Con 2. Power , p- .M
.=-=
APPLICABILITY: MODE 1. 2; 3~ . N e ACTION: . 55
.=
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than s l(@ 57. of RATED T.HERMAL POWER within the next 4 hours. _ , . _
.: . . i:.:
Ef SURVEILLANCE RECUIREMENT5 M
==-
=c:
f 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be E, within their limits at least once per 12 hours. ._- 4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months. I.=c.
. r
' > 1. ,. ;'_
~
. ~:.:
.3
=-
y=.
-; -- : ' ', - g , ,
. ;e -
- . .t. . m 5 5 . 2 . :, . c.
. .: ==
.. . ;- .: .=: .. L -2 25
_C.: T' CALVERT CLIFFS. UNIT 2- 3[4 2-13! Amendment No. 7, 13
?5
i TABLE 3.2-1 9 g DNS PARAMETERS m p
. LlHITS .
. Il Four Reactor Three Reactor Two Reactor Two Reactor * { Coolant Pumps Coolant Pumps Coolant Pumps Coolant Pumps z ; Parameter Operating Operating Operating-Same Loop Operating-Opposite Loop
-4 m Cold Leg Temperature ** ** **
--< 548*F 2-TO Pressurizer Pressure > 2225 psla* ,
Reactor Coolant System Total Flow Rate > 370,000.gpa
~
AXIAL SilAPE INDEX hure3M +
** ** ** i M44
- Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED TIERNAL POWER E ;.per minute or a TilERMAL POWER step increase of greater than 10% of RATED TilERMAL POWER.
**These values lef t blank pending NRC approval of ECCS analyses for operation with less than four reactor coolant pumps operating. . .
F * " Tho AKI AL SHAPE INDEX Caro Stour :Shali m-be maintainod u.5dhin tho. g U<iits as6blishod % mo. Babc Axial Saru. SalacLo @s6m(805s5) [. 8, b CER insacWons of +ho. laact bank ol 155% bbdn BRSSS is l; Il f opa.cabta. , oc toiWin tho. limes of Ficgura .3.2-9 b CEA y ;[. .
."_, insoc%ns spac.ihd % nqura 3.1-a. -
, i.
e j'. 0 I : ,, "
\ ' l jf e ;
~
r j ; ; g . 3.g ;j . j ; . 8 - l' . ,*
. .'",i ,. *a 6. . C. ii: . in.ua . c.t. i:i t'
n .
~
TABLE 3.31 (Continued) TABLE NOTATION ,
~
With the protective system trip breakers in the closed position and l the CEA drive system capable of CEA withdrawal. .
#The provisions of Specification 3.0.4 are not applicable. ,
(a) Trip may be bypassed below 10 ofRATEDTHERMALPOWJR;bypassshall
.be automatically removed when THERMAL POWER is > 10 ~
of RATED THERMAL POWER. 5mo (b) Trip may be manually bypassed bel 8 psia; bypass shall be , automatically removed at or above 685 psia. mo *
. (c) Trip may be bypassed below 155 of RATED THEPy.AL POWER; bypass shall be automatically removed when THERMAL .'0WER is > 155 of RATED THERMAL PO'. ~R.
(d) Trip may be bypassed below 10 % and above 12% of RATED THERMAL POWER.. (e) ~ Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3. - (f) There shall be at least two decades of overlap between the Wide ~ Range Logarithmic Neutron Flux Monitoring Channels and the Power . g Range fleutron Flux Monitoring Channels.
~
ACTION STATEMENTS , ACTION 1 - ~W ith the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within' 48 hours or be in HOT STANOBY within the next 6 hours and/or open the protective system trip breakers. ACTION 2 - With the number of OPERABLE channels one less than the l Total Number of Channels, STARTUP and/or POWER OPERATION l may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.
CALVERT CLIFFS - UNIT 2 3/4 3-4 Amendment No. 31
~~
l 1 . l .
li l! ' l; TABLE 3.3-2 Q .
% REACTORPROTECTIVEINSTRUMENTATIONRESPONS5' TIMES j ,
5 On FUNCTIONAL UNIT :, RESPONSE TINE 5
, 1, Manual Reactor Trip -
Not Appljcable
'," d.O .
g 2. Power Level - High 1 0.40 seconds *( and 1 .0 seconds ## , M 3. y Reactor Coolant flow - Low 1 0 50 seconds
- 4. Pressurizer Pressure - liigh '
< 0.90 seconds *
);
. .- .- f,;
Containment Pressure - liigh
- 5. 1 0.90 seconds ,, ! .
. G :-
p
, H( ,
- 6. Steam Generator Pressure - Low ,
1 0,90 seconds '
<'"'~. P.' '-
I! :, g g ,. '
- 7. Steam Generator Water Level - Low 1 0.90 seconds ; 6 ';,' ;
$ 8 .' Axial Flux Offset .
1 0.40 seconds *# and 1 8,0 seconds ##- [ 9.a. Thermal .1argin/ Low Pressure 1 0.90 seconds *# and 1 .0 seconds ## 4
- b. Steam Generator Pressure Difference - High
> IQ.O
- 1 0.90 seconds '
- 10. Loss of Turbine--Hydraulic Fluid Pressure - Low Not Appilcable
- 11. Wide Range Logarithmic Neutron flux Hoqitor Not Applicable k
- Neutron detectors are exempt fr'e's response time testing. Response time of the peutron flux signal portici
= of the channel shall be measured from detector output or input of first electronic ' component in channel.
0 1 u
# Response time does not include contribution of RIDS.
m .
##RTO response time only. This value is equivalent to the time interval required for the Rios output p to achieve 63.2% of-its total change when subjected to a step change in RTO temperature. -
$\ I
( ,o .
, ) -
s
- TABLE 3.3-3 (Continued)
&. TABLE NO.TATION
~~
' lSCO .
- .'~ -
(a) Trip function a bypassed in this MODE when pressurizar pressure is < 1700 psia; bypas 11 be automatically removed when pressurizer ressure is > - sia. - 1800 > '710 (c) Trip function may be bypassed in this MODE b ow 68 psia; bypass shall be automatically removed at or above 85 psla.
- >'710
- The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS . ACTION 6 - With the number of OPERABLE channel's one less than the
- Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN '
within the following 30 hours. ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied: > a. The inoperable channel is placed in either the bypassed {. or tripped condition within i bour. For- the purposes ._ of testing and maintenance, the inoperable channel may
- be bypassed for up to 48 hours from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. .
- b. Within.one hour, all functional units receiving an input from the inoperable channel are also placed in the same '
condition (either bypassed or tripped. as aoplicable) as that required by a. above for the inoperable channel.
- c. The Minimum Channels OPERABLE r*quirement is-met; however, one additional channel may be bypassed for up to 48 hours while performing tests and maintenance on
that channel provided the other inoperable channel is placed in the tripped condition.
~
- o. .
t. l! C '# 3/4 3-15 Amendment No.31 CALVERT CLIFFS - UNIT 2 l
(h
~. T .
I TABLE 3.3-4 - 99 GG ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES "9
'E 1 i pP ALLOWABLE t
qM FUNCTIONAL UNIT TRIP SETPOINT VALUES _ * !
- ' 35 '
- I
,i 1. SAFETY INJECTION (SIAS)
C cc 'a.. Manual (Trip Buttons) Not Applicable Not Applicable l 53 - b.1 ' Containment Pressure - liigh ,1 4 75 psig
,l
-id 1 4.75 psig , ,.
4
~~. r ;
c' Pressurtzer Pressure - Low -
> 578 sia - 57 psia i
1725 17a5 . m .. - I
' 2. CONTAINHENT SPRAY (CSAS) . . .
- a. Manual (Trip Buttons) Not Appilcable Not Appilcable "-
. r .
- b. . Containment Pressure -- High 1 4.75 psig 1 4.75 psig
- 3, . CONTAINMENT ISOLATION (CIS) # ,
g
,{
- a. Manual CIS (Trip Buttons) Not Applicable Not Applicable 1 4.75 psig
~
- b. . Containment Pressure - liigh . 1 4.75 psig
[
. s-i 4.# MAIN STEAM LINE ISOLATION
- a. Manual (MSIV lland Switches .
. i. -
and Feed llead Isolation Not Applicable Not Appilcable i FF , . Hand Switches) ,-
.g g
, & i .- , 4 , ' b .' . l Steam Generator Pressure - Low > 570 psia >_ 70 psia
> 635 0
'ii..
,,y iW f a Containment isolation of non-essential penetrations is also initiated by SIAS (functional P.
@ units 1.a and 1.c). ,
- e ., :, p
, , 'ro-
.i ..
REACTOR COOLANT SYSTEM l
~
l PRESSURIZER _
- r. .
't. LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a steam bubble and with at least 150 kw of pressurizer heater capacity capable of being supplied by emergency power. .. 7 ..... . . . ..... . . . . . . __ . . . . _ . ,,...... .. its pr:;r:= ;i : & :.- y,, p(gggy,;, q, Jey,l ,gh, ff he ,nsin+a,',& band bebern 133 and W&nniMCA85 I**'I an opern. 8XCIf Ton) Y WheM e14ree id ch APPLICABILITY: MODES 1 and 2. j P " f5 *tt 0 Peta 4tn and Iq+down Sihw hj jou i
# "" * "" I " M ACTION- a maximum ftts4urt her lev' off $e lia.,ifd A
\between I33 and 110 o.el ne ts.
- a. With the pressurizer inoperable du~e to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours or be in at least HOT STANOBY within the next 6 hours and in HOT SHUTDOWN within the following 12 hours.
- b. With the pressurizer otherwise inoperable, be in at least HOT STANOBY with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.
k- - SURVEILLANCE REQUIREMENTS 4.4.4 The pressurizer water level shall be determined to be within r : 7-
- d ::1_: at least once per 12 hours,
- p ;f it; ;r;;r:
he ajove $4Hd CALVERT CLIFFS - UNIT 1 3/4 4-5 Amendment No. 53 CALVERT CLIFFS , UNIT 2 Amendment No. 36
.i I./
. e
- ' 3/4.1 REACTIVITY CONTROL SYSTEMS ,._. ,, _ , , ,
.w. . .. . ~ -
BASES
~
~ ~ ~ ~ - - - - -
SORATION CONTROL , ll3/4.1.1
'I 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN - ~ -
A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditicns, 2) the reactivity transients .
- associated with postulated accident conditions are controllable within .
acceptable limits, and 3) the reactor will be maintained sufficiently . l suberitical to preclude inadvertent criticality in the shutdown condition. g
.:l b 5.M0. -
' SHUTDOWN MARGIN requirecep,ts vary throughout core life as a function of 5 l! fuel depletion, RCS baron conc'entration and RCS TThe minimum availa j ,5HUTDOWN MARGIN for no load erating conditions N9beginning of life is ll:k/k .and at end of life is 4.3". ak/k. . The SHUTDOWN MARGIN is based on the safety analyses performed for a steam line rupture event initiated at no load conditions. The most restrictive steam line rupture event occurs at EOC
; ! conditions. For the steam line rupture event at beginning of cycle conditions
. T. t.k/k is required lo control th. 35.6
'la minimum SHUTCOWN MARGIN of less than I t reactivity transient, and end of cycle conoitions requireQak/k. Accordingly ,
. ,jtheSHUTDOWNMARGINrequirementisbasduponthislimitingconditjonandis l
< 200 F, the
;: consistent with FSAR safety analysis ssumptions. With T
'!reactivitytransientsresultingfrom<nypostulatedaccidMEareminimalanda
.c ,:31 Ak/X snutdown margin provides adeq ; ate protection. With the pressurizer
,g level less than 90 inches, the sources of non-borated water are restricted to j j increa:,e tne time to criticality duri g a boro' dilution event. ~
2/c.l.l.3 BORON OILUTION . 'M.57o . A minimum flow rate of at least 3000 GPM provides adequate mixing,- i:crevents stratification and ensures that reactivity changes will be
- gradual during boron concentration reductions in the Reactor Coolant
. :' System. A flow rate of at least 3000 GPM will circulate an equivalent -
- Reactor Coolant System volume of 9,501 cubic feet in approximately
,24 minutes. The reactivity change rate associated with boron concen-ltration reductions will therefore be within the capability of operator
'i recognition and control. .
,. ! 3 /4.1.1. 4 MODERATOR TEMPERATURE COEFFICIENT (MTC) l The limitations on MTC are provided to ensure that the assumptions
- .used in the accident and transient analyses remain valid through each
- fuel cycle. The surveillance requirements for measurement of the MTC
> I: curing each fuel cycle are adequate to confirm the MTC value since this
{ { coefficient changes slowly due principally to the reduction in RCS boron l concentration associated with fuel burnup. The confirmaticn that the i measured MTC value is within its limit provides ' assurances that the I ncoefficient will be maintained within acceptable values throughout anch'
; !fuelcycle. . .. , , .,
e 'I Amendment No. 78, 31 CALVERT CLIFFS - UNIT 2 - B 3/4 1-1 O .
, g. .
POWER OIS7.I30 TION LIMITS 7: . - . ' - ,x l BASE 5 the analysis establishing the ON8 Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valid duringioperytion at the various ~~ allowable CEA group insertion limits. If F F or T exceed their basic limitations, operation may continue u5de,r Ehe ad81tional restric-tions, imposed by the ACTION statements since these additional restric- - tions provide adequate provisions to assure that the assumotions used' in estabitshing tne Linear Heat Rate, Thermal Margin / Low Pressure and
' Local Pcwer Censity - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL 90WE3 TILT > 0.10 is not expected and,1f it should occur, sub-sequent operation would be ~ restricted to only those operations
- required ~
to identify the cause of this unexpected tilt. .- - ;. I le T value of T that must be used in the equation F*Y = F*Y (1 + T9)- and F =F 7 r (1+Tq ) is the i.easured tilt. _ The surveillance recuirements for verifying that F I ~ FfandI Tq ' within and T do their limits provide not exceed assurance the assumed values. that Verifying the Factual T and vRu,es F 9f7 FafSr, e each ?uel loading prior to exceeding 757. of RATED THEh,,tAL PCWER provides additional assurance that the core was properly loaded. .
... ,~
1 - 3/t.2.5 ONS PARAMETERS . The limits on the DNS related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are' consistent with the safety analyses assumptions and been analytically demonstrated adequate to maintain a minimum CN8R of 1.19 throughout each analyzed
- transient. g3 The 12 hour periodic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restored .
- within their limits following load changes and other ' expected transient operation. The 18 month periodic measurement of- the RCS total flow rate ' '
is adequate to detect flow degradation and ensure correlation of the
, flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour basis. - .
.-- n d
CALVERT' CLIFFS - UNIT'l B-3/4 2-2 Amencment~No. 33, pf, 55 CALVERT CLIFFS - U:'IT 2 - ,:cendment ,*:c. Is , 37, 33 -
- . _ _ , - ., , --y. . - - - - - - - . -
REACTOR COOLANT SYSTEM 3ASES
)
, limi: the Reactor Coolant System pressure to wi:Si- its Safety Linit of 2750 psia following a comolaca loss a f turoine genera:ce load while operat-ing at RATED THERMAL POWER and assuming no reac:ce trip untii the first Reactor Protactive System trip satooint (Pressuri:er Pressure-High) is reached (i.e. , no credit is taken for a direct reactor trip on tne loss of turbine) and also assuming no operation of the pressuri:er Scwer operated relief valve or steam dump valves.
Cemonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordanca wi:h :ne provisions of Sac:icn XI of the ASME Boiler and Pressure Yessel Code. 3/4.1.3 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressuri:er code safety valves. These relief valves have remotaly operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for botn the relief valves and the' block valves is cacable of being sucolied fr:m an emergency power sourca ,to ensure the ability to seal this assible RCS leak- - age pach. bend +ar pressurs ttr levtl boun.ds
- Me pre 3rennmed level,)
ACS re.ssute renam,s us%on : nit bound's af on anslyttd 3/4.4.A The opern+
. e mi ensure's PRESSURIZER co ndi n)hf.j r ne $le awgessjye cher3sey nxc=f as we// as durinjes2 th /om! t ir s e*> vt $s v r s As+1om ev(nf , fvt
/,+a, shns bEa A steam bubble in tne pressuriger ni:n the level as programmed ensures tnat the RCS is not a hydraulically solid sys, and is capable of ac:ce=c-dating pressure surges during oceration. ~~c :; :- ;d ': r' also protects the pressuri:er code safety valves and ;cwer operated relief valve against wa:er relief. The ;ower operated relief valves function to relieve RCS pressure during all design transients. Oceration of :he cower operated relief valve in conjunction with a reactor :ric on a Pressuri:er--?ressure-Hign signal, minimi:es' the undesirable opening of the spring-loaded pressuri:er coce safety valves. '
l The reouf rement that 150 kw of pressuri:er heaters and their asscciated controls be cacaole of being supplied electrical ;cwer from an emergency :us orovides assurance that these heaters can be energi:ed during a loss of off-site ;ower condi: ion to main:ain nacyrd circulation a HOT STANOSY. 3/4.1.5 STEAM GENERATCRS The Surveill ance Requi r; tn.u , , ins;ection of the steam generator tubes ensure that the structural integr'. y of :his portion of :ne RCS will be maintained. The program for inservice f rspection of staam genera:ce tubes is based on a modification of Regulatory Guide 1.33, Revision 1. Inservice inspection of steam generator tubing is essential in order to CALVERT CLIFF 5 - UNIT 1 3 3/4 t-2 Amendment No. 34, 53 l I CALVERT CLIFFS - UNIT 2 Amendment No. 73, 25 1 1 1
~ = _ . _ . _ _ _ _ _ . ....e. ,
DESIGN FEATURES
.m DESIGN PRESSURE AND TEMPERATURE ,
5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 50 psig and a temperattire of . 276*F.
, . -~~.-.._
5.3 REACTOR CORE . FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zircaloy-4. Each fuel rod shall hcve a nominal active fuel length of 136.7 inches and contain a maximum total weight of 3000 grams uranium. The initial core loading shall have a maximum enrichment of 2.9g weight percent U-235. Reload fuel shall be similar in physical design
- the initial core loading and shall have a maximum enrichment of 3.7 weight percent U-235.
> 4.1 5.3.2 Except for special test as authorized by the NRC, all fuel assemblies under control element assemblies shall be sleeved with a sleeve design previously approved by the NRC.
CONTROL ELEMEf!T ASSEMBLIES . 5.3.3 'The reactor core shall contain 77 full length and no part length - l cor. trol element assemblies.. 5.4 REACTOR COOLANT SYSTSM . DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained: i a. In accordance with the code requirements specified in Section
, 4.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements,
, b. For a pressure of 2500 psia, and
- c. For a temperature of 650*F, except for the'pressurizar which, i s 700* F . .. . .
j - 7.__ q. _ , . .
??*. . . . . .
$ Ei%
... b 5-4 Amendment No. 78,31 CALhERT CLIFFS - UNIT 2
10.0 Startup Testing The startup testing program proposed for Cycle 5 is identical to the 'j program proposed for the reference cycle in Reference 1. i l
11.0 References Chapters 1 through 5 4 1.. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Sixth Cycle ' . License Application," dated February 17, 1982.
- 2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Supplement 1 to Sixth Cycle License Application," dated April 29, 1982. l 3 Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Ctark (NRC), " Fourth Cycle q License Application," dated December 4, 1980. -
- 4. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Report of Startup Testing for Cycle Four," dated May 15, 1981.
5 CEN-105(B)-P, "Reconstitutable Bg C Type CEA Design for Use in the BG&E Reactor," dated February 1979
- 6. Letter, A. E. Lundvall (BG&E) to R. A. Clark (NRC), " Unit 1 Docket 50-317, Report of Prototype CEA Performanct An ing Cycle 4 (Misnomer, Should have been: ' Report of Prototype CEA Performance during Unit 2 Cycle 3')," dated October 5, 1981.
7 Letter, A. E. Lundvall, Jr. (BG&E) to R. W.Reid (NRC), " Unit '2 Cycle 2 License Application," dated July 26, 1978.
- 8. CENPD-187, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding,"
dated June 1975 9 CEN-182(B)-P, " Statistical Approach to Analyzing Creep Collapse of Oval Fuel Rod Cladding Using CEPAN," dated September 1981.
- 10. Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), "Regarding Unit 1 Cycle 6 License Approval (Amendment #71 to DPR-53 and SER)," dated June 24, 1982.
- 11. CEN-183(B)-P, " Application of Ld".PD-198 to Zircalloy Component Dimensional Changes," dated Septemer, 1981.
- 12. CEN-83(B)-P, "Calvert Cliffs Unit 1 Reactor Operation with Modified CEA Guide Tubes," dated February 8, 1978, and Letter, A. E. Lundvall, Jr.
(BG&E) to V. Stello, Jr. (NRC), " Reactor Operation with Mcdified CEA Guide Tuber " dated February 17, 1978. 13 CENPD-139, "C-E Fuel Evaluation Model Topical Report," dated July,1974.
- 14. CEN-161(B)-P, " Improvements to Fuel Evaluation Model," dated July,1981.
15 CENPD-153-P, Revision 1, " Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered Fixed In-Core Detector System," dated May 1980.
i e o '
~
~~
References (Chapter 6)
- 1. CEffD-161-P, " TORC Code, A Comouter Code for Determining the Thermal Margin of a Reactor Core", July 1975
- 2. CEl90-162-P-A (Proprietary) and CEM'0-162-A (Nonproprietary),
" Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard S9acer Grids Part 1, Uniform Axial Power Distribution", April 1975
- 3. CENPD-206-P, " TORC Code, Verification and Simplified Modeling Methods", January 1977
- 4. Letter, P. W. Kruse to W. J. Lippold, " Responses to First Round Questions on the SCU Procram: CETOP.-Q . Code Striscttire and Mndelino Mathnds, (CEN-124 (B)-P, Part 2)", May 1981 and letter, P.W. Kruse to W.J. Lippold (above document),BGE-9676-576,May1,1981.
i
- 5. Letter D.H. Jaffe (NRC) to A.E. Lundvall,Jr. (BG&Ei, "Regarding Unit 1 Cycle 6 License Approval (Amendment #71 to DPR-53 and SER),',' June- 24, 1982.
- 6. CEN-124(8)-P, " Statistical Combination of Uncertainties, Part 2",
January 1980
- 7. CEN-83(8)-P, "Calvert Cliffs Unit 1 Reactor Operation With Modified CEA Guide Tubes", February 8, 1978 and letter, A. E. Lundvall, Jr. to V. Stello, Jr., "Rea.: tor Operation With Modified CEA Guide Tubes",
February 17, 1978
- 8. Letter, D. F. Ross and D. 'G. Eisenhut (NRC) to D. B. Vassallo and K.
R. Galler (NRC), " Revised Interim Safety Evaluation Report on .the Effects of Fuel Rod Bowing in Thermal Margin Calculation for Light Water Reactors", February 16, 1977
- 9. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 1",
December, 1979.
- 10. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 3",
Maren 1980
- 11. Letter, A. E. Lundvall. Jr. (BG&E) to R. A. Clark (NRC), " Sixth Cycle License Application," dated February 17, 1982.
l , .~ l l
O e References for Chapter 7
- 1. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Sixth Cycle License Application," Docket No. 50-317, February 17, 1982.
2a. " Statistical Combination of Uncertainties Methodology; Part 1; C-E Calculated Local Power Density and Thermal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," CEN-124(B)-P, December, 1979 2b. " Statistical Combination of Uncertainties Methodology; Part 2; Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units I and II," CEN-124(B)-P, January 1980. 2c. " Statistical Combination of Uncertainties Methodology; Part 3; C-E Calculated Local Power Density and Departure from Nucleate Boiling Limiting Conditions for Cperation for Calvert Cliffs Units I and II," CEN-124(B)-P, March 1980. 3 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), Regarding Unit 1 Cycle 6 License Approval (Amendments #71 to DPR-53 and SER), June 24, 1982.
- 4. CENPD-190A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July, 1976.
5 GEMP 482, H. C. Brassfield, et al., "Recomended Property and Reactor Kinetics Data for Use in Evaluating a Light Water-Cooled Reactor Loss-of-Coolant Incident Involving Zircaloy-4 or 304-SS, Clad 00 ," April, 1968. 2
- 6. Idaho Nuclear Corporation, Monthly Report, Ny-123-69, October,1969 7 Idaho Nuclear Corporation, Monthly Report, Hai-127-70, March,1970.
- 3. Letter, A. E. Lundvall, Jr., to R. W. Reid, " Fourth Cycle License Application," February 23, 1979 9 CENPD-188-A, "HERMITE Space-Time Kinetics," July, 1976.
- 10. CENPD-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin for a Reactor Core," July,1975 1v. St. Lucie Unit 2 FSAR 4
- 12. R. V. MacBeth, "An Appraisal of Forced Convection Burn-Cut Data," Proc.
Instn. Mech. Engrs., 1965-66, Vol. 180,.Pt. 3C, pp. 37-50.
- 13. D. M. Lee, "An Experimental Investigation of Forced Convection Burnout in High Pressure Water; Part IV, Large Diameter Tubes at About 1600 Psia,"
AEEW-R, November, 1966.
- 14. Letter, A. E. Lundvall, Jr., to R. A. Clark, "Fifth Cycle License Application," September 22, 1980.
- 15. Calvert Cliffs Nuclear Power Plant FSAR, Section 14.5, Fuel Handling Incident.
Chapter 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.
- 2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Sixth Cycle License Application," dated February 17, 1982.
Chapter 9
- 1. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Sixth Cycle License Application," dated February 17, 1982.
- 2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Supplement 1 to Sixth Cycle License Application," dated April 29, 1982.
3 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), "Regarding Unit 1 Cycle 6 License Approval (Amendment #71 to DPR-53 and SER)," dated June 24, 1982. Chapter 10
- 1. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Sixth Cycle License Application," dated February 17, 1982.}}