ML19270F624

Application to Amend OL Concerning Cycle 4 Refueling.Evaluated Design & Performance During Fourth Fuel Cycle at Full Rated Power of 2,700 Mwt

ML19270F624
Person / Time
Site:	Calvert Cliffs
Issue date:	02/23/1979
From:	BALTIMORE GAS & ELECTRIC CO.
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ML19270F619	List: ML19270F618 ML19270F623 ML19270F624
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NUDOCS 7903020357
Download: ML19270F624 (100)

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{{#Wiki_filter:CALVERT CLIFFS UflIT I CYCLE 4 REFUELIf1G LICEftSE AMEflDMEffT 7003020 357

CALVERT CLIFFS UNIT I CYCLE 4 REFUELING LICENSE AMENDMENT Table of Contents Page Numbers List of Tables ii List of Figures jv 1. Introdr, tion and Summary 1 2. Operating History of the Reference Cycle 2 3. General Description 3 4 Fuel System Design g 5. Nuclear Design 11 6. Thermal-Hydraulic Design 28 7. Accident and Transient Analysis 32 8. ECCS Analysis 67 9. Technical Speci,ications 71 10. Startup Testing ga 11. References 95 i

CALVERT CLIFFS UNIT I CYCLE 4 REFUELING LICENSE AMENDMENT List of Tables Page Number Table 3-1 Cycle 4 Core Loading ?* Table 5-1 Physics Characteristics of Cycle 4 Ccmpared with IE Reference Cycle Ca ta Table 5-2 CEA Reactivity Worths and Allowances 17 Table 5-3 Peactivity Worth of CEA Regulating Groups at Hot 18 Full Power Table 5-4 CEA Ejection Data 19 Table 5-5 Full Length CEA Drop Data 20 Table 5-6 Augmentation Factors and Gap Sizes 21 Table 5-7 Allowances for Case Load Operation 22 Table 6-1 Thermal Hydraulic Parameters at Full Power 3D ii

Lict of Tables (cont.) Pace flunber Table 7-1 Events Considered in Transient and Accident 33 Analysis Table 7-2 Core Parameters Assumed in the Safety Analyses 34 Table 7.1-1 Key Parameters Assumed in the CEA Withdrawal 37 Analysis Table 7.1-2 Sequence of Events For CEA Withdrawal From 38 Zero Power Table 7.3-1 Key Parameters Assumed in the Loss of Coolant 45 Flow Analysis Table 7.3-2 Sequence of Events for Loss of Flow 46 Table 7.4-1 Key Parameters Assumed in the CEA Ejection 56 Analyses Table 7.4-2 CEA Ejection Results 57 Table 7.5-1 Assumption For Seized Rotor Incident 61 Table 7.5-2 Sequence of Events For Seized Rotor 62 Table 8.1 Calvert Cliffs I Cycle IV Core Parameters 70 iii

CALVERT CLIFFS UtlIT I CYCLE 4 REFUELIflG LICEf;5E AMEftDMEtiT List of Fiaures Pace fiumber Figure 3-1 Projected Cycle 4 Loading Pattern 5 Figure 3-2 Fuel and Poison Rod Locations 6 Figure 3-3 Beginning of Cycle 4 Assembly Average Burnup 7 Distribution Figure 5-1 Assembly Relative Power Density at Beginning of 23 Cycle, Equilibrium Xenon Figure 5-2 Assembly Relative Power Density at Middle of 24 Cycle, Equilibrium Xenon Figure 5-3 Assembly Relative Power Density at End of Cycle, 25 Equilibrium Xenon Figure 5-4 Assembly Relative Power Density with CEA Bank 5 26 Inserted at Hot Full Power, Beginning of Cycle Figure 5-5 Assembly Relative Pcwer Densi ty with CEA Bank 5 27 Inserted at Hot Full Power, End of Cycle iv

List of Fiaures (cont.) Pace fiumber Figure 7.1-1 CEA Withdrawal Event Core Power Vs Time 39 Figure 7.1-2 CEA Withdrawal Event Core Heat Flux Vs Time 40 Figure 7.1-3 CEA Withdrawal Event RCS Pressure Vs. Time 41 Figure 7.1-4 CEA Withdrawal Event RCS Temperature Vs Time 42 Figure 7.3-1 Loss of Coolant Flow Event Core Flow Fraction 47 Vs. Tine Figure 7.3-2 Loss of Coolant Flow Event Core Power Vs. Time 48 Figure 7.3-3 Loss of Coolant Flow Event Heat Flux Vs. Time 49 Figure 7.a-4 Loss of Cociant Flow Event RCS Pressure Vs. Time 50 Figure 7.s-5 Loss of Coolant Flow Event RCS Temperature Vs. 51 Time Figure 7.3-6 Loss of Coolant Flow Evei.t Hot Channel OfiBR 52 Vs. Time Figure 7.4-1 CEA Ejection Event Core Power Vs. Time, Full 58 Pcwer Figure 7.4-2 CEA Ejection Event Core Power Vs. Time, Zero 59 Power Figure 7.5-1 Seized Rotor Event Core Power Vs. Time 63 Figure 7.5-2 Seized Rotor Event Core Heat Flux Vs. Time 64 Figure 7.5-3 Seized Rotor Event RCS Pressure Vs. Time 65 Figure 7.5-4 Seized Rotor Event RCS Temperature Vs. Time 66 v

Page 1 1. INTRODUCTION AND SUS

• NARY This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit I during its fourth fuel cycle at full rated power of 2700 MWT.

All planned operating conditions remain the same as those for Cycle 3. The core will consist of presently operating B, D and E assemblies and fresh Batch F assemblies. Plant operating requirements have created a need for flexibility in Cycle 3 termination point ranging from 8950 MWD /T to 10,000 MWD /T. In performing analyses of postulated accidents, determining limiting safety settings and establish:ng limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 4 conditions are enveloped regardless of the Cycle 3 termination point within the above burnup range. Tne evaluations of the reload core characteristics have been examined with respect to the Calvert Cliffs Unit I Cycle 3 safety analysis described in Reference 1, hereafter referred to as the " reference cycle" in this report. This is an appropriate reference cycle because of the similarity in the basic system characteristics of the two reload cores. Specific core differences have been accounted for in the present analysis. In all cases, it has been concluded that either the reference cycle analyses envelope the new conditions, or that the revised analyses presented here continue to show acceptable results. Where dictated by variations from the reference cycle, proposed modifications to the plant Technical Specifications are provided and are justified by the analyses reported herein.

Page 2 2. OPERATII;G HISTORY OF T'I REFir~ f1CE CYCLE Calvert Cliffs Unit I is presently operating in its third fuel cycle utilizing Batch A, B, C, D and E fuel assemblies. Cycle 3 operation at full power began on or about April 3,1978. The Cycle 3 startup testing was reported to the f1RC in Reference 2, It is presently estimated that Cycle 3 will terminate on or about April 21, 1979. However, flexibility in this date is necessary because of uncertainties in the future station capacity factor. The Cycle 3 termination point can vary between 8950 MWD /T and 10,000 MWD /T to accommodate the plant schedule. As of early February, the Cycle 3 burnup had reached 7100 MWD /T. Initial criticality of Cycle 4 is expected to occur on or about May 22, 1979.

Page 3 3. GENERAL DESCRIPTION The Cycle 4 core will consist of the number and types of assemblies from several fuel batches as described in Table 3-1. The primary change to the core in Cycle 4 is the removal of 40 Batch A assemblies, and 32 Batch C assemblies and their replacement by 72 fresh Batch F assemblies. Figure 3-1 shows the fuel management pattern to be employed in Cycle 4. This pattern will accommodate Cycle 3 termination burnups from 8950 MWD /T to 10,000 MWD /T. The Cycle 4 core loading pattern is 90 rotationally symetric. That is, if one quadrant of the core were rotated 90 into its neighboring quadrant, each assembly would be aligned with a similar assembly. This similarity includes batch type, number of fuel rods, initial enrichment and burnup. It does not include guide tube sleeves and demonstration fuel rod locations. Figure 3-2 shows the location of poison rods and stainless steel rods within the lattice of the Batch B irradiation test (EPRI/CE) assembly. Figure 3-3 shows the beginning of Cycle 4 assembly burnup distribution for both the minimum Cycle 3 termination burnup of 8950 MWD /T and the maximum Cycle ; termination burnup cf 10,000 MWD /T. The intial enrichment of the fuel assemblies is also shown in Figure 3-3. At the beginning of Cycle 4, the residual B-10 content of the burnable poison rods in the irradiation test (EPRI/CE) assembly is negligible. The Cycle 4 core will contain a hign burnup demonstration assembly (SCOUT) and a prototype CEA. The locations of the demonstration assembly and the prototype CEA are shown in Figure 3-1.

TABLE 3-1 CALVERT CLIFFS UNIT I CYCLE 4 CORE LOADING Ba tch_ Ba tch-To ta l Total Average Average Initial Number Number Initial Burnup Burnup Poison Poison of of Assembly Number of Enrichment E0C 3 = EOC 3 = Rods Per Loading Poison fuel Designation Assemblies wt% U-235 8950 10,000

Assembly, wt% D_4C Rods Rods I

B I 2.45 34,100 35,000 12 2.9 12 160 D 43 3.03 17,700 18,900 0 0 0 8,448 D/ 24 2.73 20,000 21,200 0 0 0 4,224 E 48 3.03 7,400 8,300 0 0 0 8,448 E/ 24 2.73 10,900 12,100 0 0 0 4,224 F 48 3.03 0 0 0 0 0 8,448 F/ 24 2.73 0 0 0 0 0 4,224 To ta l s 217 12 38,176 Notes I This is the irradiation test assembly located in the center of the core. In addition to the twelve poison rods, it contains four stainless steel rods, one in each corner of the assembly. ? 'S .n

Page 5 F F F* F F E El F Fl E D DI Fl F Fl ** El D Fl D D F Fl El El D E D F E D El D E DI E F D Fl D E DI E DI F E DI D E DI E D E F El Fl D D E D/ E B+ Location of Demonstration Assenbly (SCOUT) Location of Prototype CEA

• • • Location of Irradiation Test Assembly (EPRI/CE)

B A LT LY. ORE Figure GAS & ELECTRIC CO. CALVERT CLIFFS UNIT I CYCLE 4 Calvert Clirfs 3-1 Nuclear Power Plant CORE MAP

UNSHIMhiED ASSEMBLY g i i i I i l II l l l l l i l i i i j i lil l I I l l 1 i I I I Ii l I I I I I i l l l l l l-l I I I i l,I l l l 1 I I i I il l l l l l l I l I i l i i l l I l l l l i l l l I I l i 12 POISON ROD ASSEMBLY (CENTER ASSEMBLY ONLY) ol l l t o XI I l Xi I iXl XI i I l l l l I I l ll I !X' I I X l l l l l i l XI i l I 'XJ l l i I I i i l l 1 X X XI i l Ii XI O l l l 0 FUEL ROD LOCATION Si POISON ROD LOCATION IN TEST ASSEMBLY o STAINLESS STEEL ROD LOCATION IN TEST ASSEMBLY GAS ELE T !C CO. CALVERT CLIFFS UNIT I CYCLE 4 ASSEMBLY FUEL AND OTHER ROD LOCATIONS 3-2 Nuc ont

Page 7 INITI AL ENRICHMENT Wl0 U-235 - 3.03 3.03 BOC 4 BURNUP (W!D/T), EOC3E50 MNDIT ~0 0 BOC 4 BURNUP (MWDIT), EOC3=10000 MWDlT -- O 0 3.03 3.03 3.03 3.03 3.03 0 0 0 8D 9 00 0 0 0 9000 10400 3.03 2.73 3.03 3.03 2.73 2.73 0 0 10Jm .Urm 19Fm 0 0 0 11300 18600 20600 0 3.03 2.73 2.73 3.03 2.73 3.03 3.03 0 0 ll5m ?n3m 0 17703 175m 0 0 12800 21500 0 19000 18800 3.03 2.73 2.73 2.73 2.73 3.03 3.03 3.03 0 0 11500 08m 11100 14700 7900 19701 0 0 12800 12100 12400 15800 8800 26600 3.03 3.03

3. 03 2.73 3.03 3.0?

2.73 3.03 ~ 0 10200 20300 11200 13300 SE00 -2G200 CA00 0 11400 21500 12400 19500 6400 21400 7100 3.03 3.03 2.73 3.03 3.03 2.73 3.03 2.73 0 17;Y) 0 1u800 5700 20200 r@ 20200 3.03 0 18700 0 15900 6400 21400 7200 21400 0 0 3.03 2.73 3.03 I 3.03 2.73 3.03 3.03 3.03 8200 19500 17300 7900 203m EE 17393 6YO 3.03 9100 20700 19000 8800 21500 7200 18400 7100 0 0 2.73 2.73 3.03 3.03 3.03 2.73 3.03 2.45 % 10 0 17500 192m 64m 202m F3m 3I0m 10400 0 18800 20500 7100 21400 7100 35000 CALVERT CLIFFS UNIT I CYCLE 4 Fioure GAS C CO' coivert curr, ASSEMBLY AVERAGE BURNUP AND INITIAL ~ Nuclear Power Plan, ENRICHMENT DISTRIBUTION 3-3

Page 8 4 FUEL SYSTEM DESIGft The mechanical design for Batch F reload fuel is essentially identical to that of the Batch E fuel used in Calvert Cliffs Unit I and described in the reference cycle submittal, Reference 1. Details of the Batch B and D fuel design parameters can be found in References 3 and 4, respectively. C-E has performed analytical predictions of cladding creep-collapse time for all Calvert Cliffs Unit I fuel batches that will be irradiated in Cycle 4 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 4 duration. These analyses utilized the CEPAN computer code (Refer-ence 5) and included as input conservative values of internal pressure, cladding dimensions, cladding temperature and neutron flux. The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch F fuel are identical to those of the Batch B, D and E fuel from Cycle 3. Thus, the chemical or metallurgical performance of the Batch F fuel will remain unchanged from the performance of the Cycle 3 fuel. 4.1 HARDWARE MODIFICATIONS TO MITIGATE GUIDE TU3E WEAR All fuel assemblies presently in Cycle 3 which will be placed in CEA locations in Cycle 4, with the exception of the Batch B test assembly, will have stainless steel sleeves installed in the guide tubes in order to prevent guide tube wear. The sixteen Batch F assemblies which will be placed in dual CEA locations vi,1 also have stainless steel sleeves installed in the guide tubes. A detailed discussion of the design of the sleeves end their effect on reactor operation

Page 9 is contained in Reference 6. The remaining 56 Batch F assemblies will either be modified in the same manr.ar as the sixteen Batch D assemblies in Calvert Cliffs II Cycle 2 or will have stainless steel sleeves. A detailed discussion of the Calvert Cliff II Cycle 2 Batch D modification and its effects on reactor operation is contained in Reference 16, 4.2 BATCH F DEMONSTRATION ASSEMBLY One Batch F assembly consists of 161 standard fuel rods and 15 demonstration rods. A detailed description of these demonstration rods may be found in Reference 7. The mechanical design of the assembly components other than the 15 demonstration rods in this assembly is identical to the design of the other Batch F assemblies. Figure 3-1 displays the location of the demonstration assembly in Cycie 4. No demonstration fuel rod in the demonstration assembly will have a power level within 10% of the maximum radial power peak in the core during Cycle 4. Two types of demonstration fuel rods are being introduced in this assembly (Reference 7). Helium fill pressure differences between the demonstration fuel rod designs were introduced for the following reasons: a. The location of non-fueled regions at grid contact points results in the possibility of somewhat reduced grid / rod contact forces. To offset this possibility, fill pressure in these rods was in-creased to raise beginning of life internal pressure, and thus decrease the magnitude of c lad creepdown. b. A difference in voiJ volume e <ists between the two types of demonstration rods. A higher 'nitial fill pressure will not appreciably increase end of life internal pressure in the test rod type rith greater void volume. s The magnitude of the subject difference ia He fill pressure is 65 psi.

Page 10 CE has performed analytical predictions of c, adding creep-collapse time for the demonstration fuel rods that will be irradiated in Cycle 4 and has concluded that the collapse resistance of these demonstration fuel rods is sufficient to preclude collapse during their design lifetime. This lifetime will not be exceeded by the Cycle 4 duration. 4.3 Prototype CEA Cycle 4 will utilize one prototype CEA as part of regulating Bank 5. The location of this prototype CEA is chown in Figure 3-1. This new CEA design involves a change in cladding material (Inconel to stainless steel) and specially designed reconstitutable poison rods which serve as a lead for all future poison rod designs. The purpose of the poison rod design is to demonstrate that the silver-induim-cadmium which is presently used in the tips of poison rods can be replaced with B C. Both the changes in cladding material 4 and the replacement of Ag-In-Cd with B C are being made for reasons 4 of economics and to improve material availability. A more detailed description of the design of this prototype CEA is given in Reference 8. This prototype CEA meets the same design criteria and has the same design margin as the CEA's used in the reference cycle. There-fore, it has no adverse affect on mechanical integrity, thermal-hydraulics and safety. A detailed description of the affects of the prototype CEA can be found in Reference 8.

Page 11 5. fiUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS 5.1.1 Fuel Manacement The Cycle 4 fuel management employs a mixed central region as described in Section 3. The fresh Batch F is comprised of two sets of assemb_ lies, each having a unique enrichment in order to minimize radial power peaking. There are 48 assemblies with an enrichment of 3.03 wt% U-235 and 24 assemblies with an enrichment of 2.73 wt% U-235. The Cycle 4 burnup capacity for full power operation is expected to be between 10,000 MWD /T and 10,500 MWD /T, depending on the final Cycle 3 termination point. The Cycle 4 performance characteristics have been examined for a Cycle 3 termination between 8950 end 10,000 MWD /MTU and limiting values established. The proposed loading pattern is presented in Figure 3-1. The pattern is applicable to any Cycle 3 termination point between the stated extremes. Physics characteristics including reactivity coefficients for Cycle 4 are listed in Table 5-1 along with the corresponding values from the reference cycle. It is noted that the values of parameters actually employed in safety analyses are typically chosen to conservatively bound predicted values including uncertainties and allowances. Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for Cycle 4 with a comparison to reference cycle data. The power dependent CEA insertion limit and CEA group identification are unchanged from the reference cycle. Table 5-3 shows the reactivity worths of various CEA groups calculated at full power conditions for Cycle 4. 'g, N" '*I'*"IN".4

• * ~ * '

'""*'1W# Pd*******,}

• M' ##,

r#

Page 12 5.1.2 Power Distribution The radial power distributions described in this section are calculated data without uncertainties or other allowances. However, 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 as described in Reference 9. Planar radial power distributions for the unrodded core at beginning of cycle, 6 GWD/T and end of cycle are shown in Figures 5-1 through 5-3 for a Cycle 3 endpoint of 10,000 MWD /T. The planar radial peaks described here are characteristic of the major portion of the active core length between approximately 20 percent and 80 percent of the fuel height. The maximum planar radial power peak for the early shutdown of Cycle 3 is less than that for the later shutdown. The planar radial power distributions for the above regions with CEA Group 5 fully inserted at teginning and end of Cycle 4 are shown in Figures 5-4 and 5-5 for a Cycle 3 endpoint of 10,000 MWD /T. The maximum planar radial pin peak of 1.56 occurs at beginning of cycle and decreases over the cycle. For both DNB and PLHGR 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 4 These conservative values, which are specified in Section 7 and Section 9 of this document, establish the allowable limits for power peaking to be observed during operation. The axial peaking is linited by the 1, ;iting conditions for operation on the axial shape index (ASI). Within these ASI limits, the necessary DNBR and PLHGR margins are maintained for a wide range of possible axial shapes. The maximum three-dimensional or total peaking factor anticipated in Cycle 4 during norral base load

Page 13 all rods out operation at full power is 1.81, not including uncer-tainties and augmentation factors. 5.1.3 Safety Related Data 5.1.3.1 Ejected CEA The maximum reactivity worths and planar radial power peaks associated with an ejected CEA event are shown in Table 5-4 for both beginning of cycle and end of cycle. These values encompass the worst conditions anticipated during Cycle 4 for any expected Cycle 3 termination point. The pointwise Doppler Feedback technique described in Reference 4 was not utilized in the PDQ calculations of ejected CEA worths and associated power peaks. 5.1.3.2 Dropped CEA Table 5-5 contains data for several dropped CEA configurations for both extremes in core life and covers the range of expected values. These data have been calculated with the pointwise Doppler feedback technique described in Reference 4 This treatment is consistent with the safety analysis sirce the time to minimum Of!BR is on the order of one to two minutes, allowing ample time for fuel temperature redistribution following the CEA drop. The power peaking values used in the safety analysis are higher than those expected to occur at any time in Cycle 4 5.1.3.3 Augmentation Factors Augmentation factors have been calculated for the Cycle 4 core using the calculational model described in Reference 10. The input information required for the calculation of augmentation factors that is specific to the core under consideration includes the fuel densification characteristics, the radial pin power distri-bution and the single gap peaking factors. Augmentation factors

Page 14 for the Cycle 4 core have been conservatively calculated by combining for input tne largest single gap peaking factors with the most conservative (flattest) radial pin power distribution. The calculations yield non-collapsed clad augmentation factors showing a maximum value of 1.049 at the top of the core. The augmentation factors for Cycle 4 are compared to the reference cycle values calculated with the same model in Table 5-6. 5.2 Al'ALYTICAL INPUT TO IN-CORE MEASUREMEf1TS In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the nanner described in Reference 11, which is the same method used for the reference cycle. 5.3 NUCLEAP DESIGN "ETEODOLOGY The coarse mesh computer program ROCS (Reference 12) has been used along with the standard fine mesh design program PDQ (Reference 13) in the Cycle 4 safety analysis, a. ROCS was used to survey a variety of core configurations to determine limiting conditions, b. ROCS was used to obtain axial power shapes, to weight the relative importance of fine mesh PDQ planar power and burnup distributions in the determination of three-dinensional effects and to determine the impact of the three-dimensional gross power distributions on reactivity parameters. c. ROCS was used to compute selected safety parameters. The calculation of those limiting parameters which require knowledge of 1-pin peaking factors continues to be based on the fine mesh PDQ program. d. Two-and three-dimensional ROCS calculations were used in conjunction with two-dimensional PDQ calculations to obtain best estimate core parameters such as those shown in Table 5-3.

Page 15 5.4 UNCERTAIrlTIES I:1 MEASURED POWER DISTRIBUTIONS The power distribution measure: rent biases and uncertainties which are applied to reload cycles are: Base Load Operation Load Follow Operation Fq: i + cr 7.0% 10.0% Fr: - +c 6.0% 8.0% These numbers are to be used when the IllCA mod-' described in Reference 11 is used for monitoring power distribution parameters during operation. In the development of LCOs and LSSS data for Cycle 4, allowances for power distribution uncertainties have been applied which are consistent with the measurement uncertainties and biases givcn above. The allowances are presented in Table 5-7.

l$bkL O-l Page 16 CALVERT CLIFFS UNIT I CYCLE 4 PHYSICS CHARACTERISTICS Reload Cycle Reference Calculated Units Cycle Values Dissolved Boron Dissolved boron content for criticality, CEAs wi thcrawn hot (565 F), full power, PPM 9131 10301 equilibrium xenon, BOC 8542 97c 2 Boron worth hot (565*F) BOC PPM /%ap 93 95 hot (565 F) E0C PPM /Sto 81 83 .Peactivity Coefficients (CEAs Withdrawn) Moderator temperature coefficients, hot opera ting hot (565 F), equilibrium 10 4 ao/ F -0.4 -0.4 ~ xenon, BOC hot (565 F), E0C 10 4 to/ F -2.2 -2.1 Doppler coefficient hot (532 F) zero power, 10 5 to/ F -1.50 -1.50 BOC hot (565 F), full power, 10 5 ac/ F -1.14 -1.20 BOC hot (565 F), full power, 10 5 tp/ F -1.24 -1.37 E0C Total Delaved Neutron Fraction, eeff EOC .00610 .00616 ECC .00530 .00522 Neutron Generation Time, t* BOC 10~0sec 25.4 24.8 EOC 10 6sec 29.2 29.1 Notes lEarly previous cycle shutdown. 21ato rrovinue cvcle chutdown.

Page '.7 TABLE 5-2 CALVERT CLIFFS UNIT I CYCLE 4 LIMITING VALUES OF REACTIVITY WORTHS ann ALLOWANCES, i;ac BOC E0C Reload Reload Reference Cycle Cycle Reference Cycle Cycle Worth Available Worth of all 8.1 8.9 8.9 9.4 CEAs insertedl Stuck CEA 1.1 1.8 1.6 1.7 allowance Worth o# all 7.0 7.1 7.3 7.7 CEAs less highest worth CEA stuck out Worth Required (Allowances) Power defect, 2.0 1.6 2.7 2.4 HFP to HZP Moderator voids 0.0 0.0 0.1 0.1 CEA bite, boron 0.5 0.5 0.5 0.c deadband and maneuvering band Shutdown margin 3.42 3.42 3.4 3.4 and safeguards allowance Total reactivity 5.9

5. 3 6.7 6.2 required Available Worth Less Allowances Margin available

>l.1 >l. 8 0.7

1. 5 Notes

'PLRs not included. 2Less than 3.4"2r is recuired at BOC.

Page 18 TABLE 5-3 CALVERT CLIFFS UNIT I CYCLE 4 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %ap Regulatina CEAs Beginnina of Cycle End of Cycle Group 5 .64 .68 Group 4 .29 .39 Group 3 .77 .88 Notes Values shown assume sequential group insertion

Page 19 TABLE 5-4 CALVERT CLIFFS UNIT I CYCLE 4 LIMITING VALUES OF CEA EJECTION DATA Reference Reload Cycle Cycle Post-Ejection Maximum Radial Power Peak Full power with 3.20 3.36 Bank 5 inserted; worst CEA ejected Zero power with 8.05 9.83 Banks 5+4+3+2 inserted; worst CEA ejected Maximum Ejected CEA Worth (%Ac) Full power with 0.38 0.32 Sank 5 inserted; worst CEA ejected Zero power with 0.52 0.60 Ba nks 5+4+3+2 inserted; worst CEA ejected Notes 1 All values are the limiting values used in the transient analyses. 2Locations of CEA types are shown in Figure 5-2 of Reference 4 3 Uncertainties and allowances are included in the above data. 1 i

Page 20 TABLE 5-5 CALVERT CLIFFS UNIT I CYCLE 4 .' LL LENGTH CEA DROP DATA BOC Worth, %ao Maximum Percent Increase in Pin Peak AR0, Drop 12 .11 11.6 ARO, Drop 'l .13 13.3 ARO, Drop 10 .12 8.7 Bank 5 In, Drop 12 .11 11.3 Bank 5 In, Drop 11 .12 12.4 Bank 5 In, Drop 10 .14 9.8 E0C AR0, Drop 12 .11 11.5 AR0, Drop 11 .12 13.1 ARD, Drop 10 .16 11.5 Bank 5 In, Drop 12 .11 10.7 Bank 5 In, Drop 11 .12 11.9 Bank 5 In, Drop 10 .17 12.1 Notes 1No uncertainties are included in the data. 2 Location of CEA types can be found in Figure 5-2 of Reference 4. 3CEAs are either fully inserted or fully withdrawn in radial calculations. 4ARO = All Rods Out.

Page 21 TABLE 5-6 CALVERT CLIFFS UNIT I CYCLE 4 AUGMENTATION FACTORS AND GAP SIZES Reference Cycle Reload Cycle Noncollapsed Honcollapsed Core Core Clad Clad Height Height Augmenta tion Gap Size Augmentation Gap Size (Percent) (inches) Factor (Inches) Factor (Inches) 98.5 134.7 1.069 2.94 1.049 2.94 86.3 118.6 1.065 2.59 1.045 2.59 77.9 106.5 1.061 2.33 1.042 2.33 66.2 90.5 1.056 1.98 1.037 1.98 54.4 74.4 1.049 1.64 1.031 1.64 45.6 62.3 1.043 1.38 1.027 1.38 33.8 46.2 1.034 1.04 1.021 1.04 22.1 30.2 1.024 0.69 1.01 5 0.69 13.2 18.1 1.01 7 0.43 1. 01 0 0.43 1.5 2.0 1.003 0.086 1. 0 01 0.086 Notes Values are based on approved model described in Refererme 10. The conservative reference cycle peak augmentation factor of 1.069 was used in the Cycle 4 safety analyses.

Page 22 TABLE 5-7 CALVERT CLIFFS UtlIT I CYCLE 4 ALLOWAfiCES FOR BASE LOAD OPERATION Cycle 4 Reference Cycle kw/ft LCO 7.0% 5.8% kw/ft LSSS 7.0% 5.8% DNBR LCO 6.0% 5.1% DNBR LSSS 6.0% 5.1%

Page 23 0.72 0.93 0.73 0.99 1.16 1.15 1.10 0.85 1.21 1.15 1.02 0.91 1.27 0.85 1.24 1.08 0.98 1.28 0.98 0.97 X 0.73 1.21 1.08 1.02 0.99

1. 05 1.08 0.89 0.99 1.15 0.97 0.99 0.93 1.11 0.82 1.05 1.16 1.02 1.27 1.04 1.10 0.81 1.01 0.78 0.73 1.16 0.91 0.95 1.06 0.80 1.00 0.85 0.95 0.95 1.13 1.26 0.91 0.85 0.98 0.76 0.98 0.61 NOTE: X = MAXIMUM 1-PIN PEAK = 1.48 GAS EE IC CO.

CALVERT CLIFFS UNIT I CYCLE 4 Figure coivert clins ASSEMBLY RELATIVE POWER DENSITY AT BOC, 5-1 Nxicar Pwier Plant EOUILIBRIU M XENON

Page 24 0.71 0.90 0.71 0.94 1.09 1.08 1.05 0.81 1.12 1.08 0.98 0.90 1.22 6 0.81 1.14 1.03 0.96 1.24

0. 98 0.98 X

0.71 1.12 1.03 1.02 1.01 1.06 1.12 0.95 0.94 1.08 0.96 1.01 0.98 1.17 0.90 1.14 1.09 0.98 1.24 1.06 1.16 0.91 1.13 0.90 0.71 1.09 0.90 0.97 1.11 0.89 1.13 0.98 1.10 0.90 1.07 1.22 0.94 0.92 1.09 0.89 1.13 0.75 NOTE: X= MAXIMUM l-PIN PEAK = 1.39 gas IELE k IC CO-CALVERT CLIFFS UNIT I CYCLE 4 Figu re calvert cliffs ASSEMBLY RELATIVE POWER DENSITY AT 6 GWDIT, Nuclear Power Plant EQUILIBRIUM XENON 5_2

Page 25 0.72 0.90 0.72 0.93 1.07 1.07 1.03 0.81 1.09 1.06

0. 97 0.91 1.20 0.81 1.11 1.01 0.96 1.21 0.98 0.98 X

0.72 1.10 1.01 1.01 1.01 1.05 1.12 0.96 0.93 1.06 0.95 1.01 0.98 1.16 0.93 1.15 1.07 0.97 1.21 1.05 1.16 0.94 1.15 0.93 0.72 1.07 0.91 0.98 1.11 0.92 1.15 1.02 1.13 0.90 1.05 1.20 0.96 0.95 1.12 0.93 1.16 0.81 NOTE: X= MAXIMUM 1-PIN PEAK = 1.35 BA LT Lu CRE CALVERT CLIFFS UNIT I CYCLE 4 Figure GAS & ELECTR C CO. ASSEMBLY RELATIVE POWER DENSITY AT EOC, EQUILIBRIUM XENON 5_3 Nuc r rP nt

Page 26 0.72 0.91 0.72 1.01 1.17 1.10 1.00 NW,/ 0.72 1.14 1.16 1.04 0.85 ' 0. 90,e /// //- ' !/,7 / / '0. 86/,/, 0.72 0.99 1.00 1.34 0.98 0.91 -/// y 0.73 1.14 1.00 1.01 1.05 1.14 1.16

0. 95 1.01 1.16 1.00 1.05 1.02 1.23 0.91 1.16 1.18 1.04 1.33 1.12 1.22 0.91 1.12 0.85

~ 0.73 1.11 0.84 0.96 1.14 0.88 1.10 0.89 0.93

• 0

'/,/ /,/ ,'/,//

1. 03 0.89 / 0.86 0.90 1.09 0.83 0.96 0.39

////) //// NOTE: X =MAXIMU M 1-PIN PEAK =1.56 '/p// CEA BANK 5 LOCATIONS ,p GAS IE E CALVERT CLIFFS UNIT I CYCLE 4 Figure IC CO-calvert curts ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 5-4 Nuclear Power Plant INSERTED, HFP, BOC

Page 27 0.72 0.88 0.71 0.95 1.08 1.02 0.93 '//,// 0.67 1.03 1.07 0.99 0.84 0.8I '// / 0.67 0.75 0.93 0.98 1.27 0.98 0.93 ff/f 0.71 /1. 03.

0. 93 1.01 1.08 1.14 1.20 1.03

'///b

0. 95 1.07 0.98 1.08 1.08 1.29 1.02 1.27 1.08 0.99 1.27 1.14 1,29 1.04 1.25 1.00 X

0.72 1.02 0.84 0.98 1.20 1.02 1.25 1.03 1.06 //,'// '//// 0.94 /0. 82 /, 0.90 1.01 1.23 0.99 1.09 / 0. 47/,'- ','//// ,//,/,/ NOTE: X= MAXIMUM 1-PIN PEAK =1.47 / CEA BANK 5 //'// LOCATIONS /// B A LT LM ORE CALVERT CLIFFS UNIT I CYCLE 4 Fi ure GAS & ELEC IC CO. g ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 5-5 Nuclear Power Plon, INSERTED, HFP, EOC

Page 28 6. THERMAL-HYDRAULIC DESIGil 6.1 DNBR ANALYSIS Steady-state DNBR analyses of Cycle 4 at the rated power of 2700 MWT have been performed using TORC /CE-1. Appropriate adjustments were made to the input of these codes to reflect the Cycle a Power distribution. Table 6-1 contains a list of pertinent thermal-hydraulic design parameters used for both the safety analysis and the generation of reactor protectinn system setroint infnrmation. The analyses were performed in the same manner as for the reference cycle and include the following two considerations. (1) As described in Reference 9 TORC /CE-1 was used in the generation of limiting conditions for operation on DNBR margin in the Technical Specifications. TORC was used for all A00s and postulated accidents which were reanalized for Cycle 4. (2) The engineering factor on local heat flux, the nuclear uncertainty on Fr and ar, additional factor designated as an augmentation factor for fuel rod bowing (although a " bowing augmentation factor" has been used here, the bowina penalty on DNCR is still evaluated within NRC's interim guidelines and presented in Section 6.2) were combined statistically for these ana e?s. (The method by which these factors are combined is described in Section 4 of Reference 14 ) Because the local heat flux factor was applied in this way, credit for its inclusion in the TORC deck was taken in determining the setpoints. This is identical to the procedure used in the revised analysis for Calvert Cliffs Ulit I Cycle 3 (Reference 1).

Page 29 Investigations have been rade to ascertain the effect of the CEA guide tube wear problem and the sleeving repair on DNBR margins i as established by this type of analysis. The findings were reported to the NRC in References 6 andl5 which ccnclude that the wear problem and the sleeving repair do not adversely affect DNBR margin. 6.2 EFFFCTS OF FUEL ROD BOUING ON DUB MARGIN The fuel rod bowing effects on DNB margin for Calvert Cliffs Unit I have been evaluated within the guidelines set forth in Referencel4. as used in the reference cycle analysis (Reference 1). o A total of 81 fuel assemblies will exceed the NRC-specified DNB penalty threshold burnup of 24,000 MWD /T, as established in Reference 14 during Cycle 4. At the end of Cycle 4, the maximum burnup attained by any of these assemblies will be 42,800 MUD /T. From Reference 14, the corresponding DNB penalty for 42,800 MUD /T is 6.30 percent. An examination of power distributions for Cycle 4 shows that the maximum radial peak at hot full power in any of the assemblies that eventually exceed 24,000 MWD /T is at least 10.30 percent less than the maximum radial peak in the entire core. Since the percent increase in DNBR has been confirmed to be never less than the percent decrease in radial peak, there exists at least 10.30 percent DNB margin for assemblies exceeding 24,000 MWD /T relative to the DNB limits established by other assemblies in the core. This margin is considerably greater than the Reference 14 reduction penalty of 6.30 percent imposed upon fuel assemblies exceeding 24,000 MWD /T in Cycle 4. Therefore, no power penalty for fuel rod bowing is required in Cycle 4.

1 TABLE 6-1 CALVERT CLIFFS UllIT I CYCLE 4 THEPfiAL-ilYDRAULIC PARAllETERS AT FULL P0hTR General Characteristics Unit Reference Cycle Reload Cycle Total fleat Output (Core Only) MUT 2700 2700 6 ) 10 BTU / hour 9215 9215 Fraction of Ileat Generated in Fuel Rod 0.975 0.975 q Primary System Pressure flominal PSIA 2250 2250 Minimum in steady state. PSIA 2200 2200 Maximum in steady state PSIA 2300 2300 1 Design Inlet Temperature 'F 549 550 ) Total Reactor Coolant Flow (Minimum Steady State) GPM 370,000 370,000 f 6 10 lb/ hour 139.0 139.0 6 Coolant Flow Through Core (at 2250 psia. 550 F) 10 lb/ hour 133.9 135.3 ** Hydraulic Diameter (Nominal Channel) ft 0.044 0.044 6 2 Core Average Mass Velocity (at 2250 psia, 550 F) 10 lb/ hour-ft 2.51 2.53 ** y Pressure Drop Across Core (Minimum Steady State Flow PSI 10.2 10.6 ** Irreversible ao Over Entire Fuel Assembly) 8 Pressure Drop Across Vessel (Based on flominal Dimensions PSI 32.2 32.6 ** and Minimum Steady Flow)

} TABLE 6-1 -(CONTINUED) General Characteristics Unit Reference Cycle Reload Cycle I Number of Fuel Rods in Core 38,176 38,176 2 Core Average lleat Flux ( Accounts for Above Fraction of BTU / hour-ft 181,200 181,200 lleat Generated in Rod and Axial Densification Factor) Total ifcat Transfer Areas (' Accounts for Axial Densi-ft 49,600 49,600 Tication Factor) Average Linear lleat Rate of Fuel Rod (Cold, Undencified, kw/ft 6.05 6.05 Includes Above fraction of lleat Generated in Rod) 2 h Film Coefficient at Average Conditions BTV/ hour-ft, F 5850 5850 Maximum Clad Surface Temperature F 657 657 3 Average Film Temperature Difference 'F 33 33 i Average Core Enthalpy Rise BTU /lb 69 68** Calculational Factors Engineering heat flux factor 1.03 1.03 Engineering factor on hot channel heat input 1.03 1.03 I Flow factors e, inlet plenum nonuniform distribution 1.05 w rod pitch, bowing and clad diameter 1.065 1.065 l Fuel densification factor (axial shrinkage) 1.01 1.01

• Not used in TORC; inlet flow distribution is input g
  • Theo paramet ere, have chanced due to the decrease in bypass flow

Page 32 7.0 ACCIDENT AND TRM:51ENT AWYSIS OTHER THAN LOCA The purpose of this section is to present the results of the safety analysis (other than LOCA) for Calvert Cliffs Unit 1, Cycle 4 at 2700 fult. Tne events considered for this analysis are listed in Table 7-1. These are the design basis events for the plant. These events can be categorized into tne folloaing groups: 1. Anticipated Operational Occurrences for which the Reactor Protection System prevents the Specified Acceptable Fuel Design Limits (SAFDLs) from being exceeded; 2. Anticipated Operational Occurrences for which the initial steady state overpower margin must be maintained in order to 1;revent the SAFDLs from being exceeded; 3. Postulated Accidents. Each of the events listed in Table 7-1 has been reviewed for Cycle 4 to determine if an explicit reanalysis was required. Table 7-1 indicates the analysis status of each event. Table 7-2 presents the safety parameters used in the cycle 4 analysis in comparison to the reference cycle. The review of each design basis event (DBE) entails a comparison between all the current and reference cycle key transient parameters that significantly impact the results of the event. The reference analysis for each event is the analysis upon which the licensing of Calvert Cliffs Unit 1, Cycle 3 was based except where noted differently. If all the current cycle values of key parameters for a particular event are bounded by (conservative with respect to) the reference cycle, no reanalysis is required. In some instances, a reanalysis is performed if it is deemed beneficial from the stand-point of enchanced operating flexibility or if it is desired to bound parameters which are expected to become more adverse in future cycles. The results of the review show that the key parameters to all the DBEs for Cycle 4 opation are the same as the specified reference cycle input parameters, e:: cept for the folloding: 1. CEA drop time to 90% inserted 2. Integrated Radial Peaki w Factor (F ) r 3. Seized Rotor Pin Census 4. Core Bypass Flow Fraction 5. RTD Response Time For all DBEs other than those reanalyzed, the Calvert Cliffs Unit 1 safety analysis submitted either in the FSAR or in previous reload cycle license submittals bound the results that would be obtained for Unit 1, Cycle 4 and demonstrate safe operation of Calvert Cliffs Unit 1, Cycle 4 at 2700 MWt. Since the CEA drop time to 90% insertion has increased for Cycle 4, the Loss of flow event, CEA ejection event, RCS depressurization event, Seized Rotor event and the CEA withdrawal event were reanalyzed. These events are adversely impacted by the CEA drop time, since a reactor trip is necessary to terminate the event. For Cycle 4 the Integrated Radial Peaking Factor (F ) has decreased in concari-son to Cycle 3. In addition, the net core nass flow has increased for Cycle 4 in 7 comparison to Cycle 3 due to the decrease in the calculated value of the cor.e bypass flow. The decreased peaking factor and the increased core flow will not have any adverse impact on the events listed in Table 7-1 for Cycle 4.

Page 33 TABLE 7-1 CALVERT Cl.IFFS UNIT 1, CYCLE 4 EVENTS CONSIDERED IN TRANSIE.NT /SD ACCIDE.T AELYSIS \\ Analysis Status Anticipated Operational Ozcurrenzes for which the RPS Assures no Violation of SAFDLs: Control Element Assembly Withdrawal Reanalyzed Boron Dilution Not Reanalyzed Startup of an Inactive Reactor Coolant Pump Not Reanalyzed Excess Load Not Reanalyzed Loss of Load Not Reanalyzed loss of Feedwater Flow Not Reanalyzed Excess licat Removal due to Feedwater t!alfunction Not Reanalyzed Reactor CoolantsSystem Depressurization Reanalyzed l loss of Coolant Flow Reanalyzed Irss of AC Power Not Reanalyzed Anticipated Operationa] Occurrences which are Dependent on Initial Overpower l!argin.for Protection Against Violation of SAFDLs: Loss of Coolant Flow Reanalyzed Loss of AC Power Not Reanalyzed Full Length CEA Drop Not Reanalyzed Part Length CEA Drop Not Reanalyzed Part Length CEA h!a1 positioning Not Reanalyzed Transients Resulting from F!alfunction of One Not Reanalyzed Steam Generator Postulated Accidents: CEA Ejection Reanalyzed Steam Line Rupture Not Reanalyzed Steam Generator Tube Rupture Not Renanalyzed Seized Rotor Reanalyzed 1Requires low Flow Trip ~ .n n.,.n n., a -

Page 34 TABLE 7-2 CALMT CLIFFS UNIT 1. CYCLE 4 CORE PNUNETTRS /tSSUMED IN ' DIE SAFliTY ANALYSES Unit 1 Unit 1 Physics Parameters Units gele3 Values Cycle 4 Valt Planar Radial Peaking Factors For DNB Margin Analyses Unrodded Region 1.65* 1.58 Bank 5 Inserted 1.78* l 71 For lae/ft Limit Analyses (F,) Unrodded Region 1.66* 1.66 Bank 5 Inserted 1.79* 1.79 Peak Augmentation Factor 1.069 1.069 -4 Moderator Temperature Coefficient 10 Ap/ F - 2.,5 -> +. 5 - 2. 5 -> +. 5 Shutdown Margin (Value used in Zero Power -3.4 -3.4 (SLB) Safety Parameters ~^ Power Level % of 2700 Nt 102 102 Maximtra Steady State Core Inlet Temperature F 550 550 Minimum Steady State RCS Pressure psia 2200 2200 Minimum Reactog Coolant Core Flow Ib/hr 134.22 135.24 (2200 psia, 550 F) Full Power 1p .12* .14 Maximum CEA Insertion at Full Power % Insertion 25 25 of Group 5 Maximin Allowable Initial Peak Linear lleat Rate for DBEs other than LOCA kw/ft 16.0 16.0 Steady State Linear Heat Rate to Fuel kw/ft 21.0 21.0 Centerline Melt

• These values are revised limits quoted in the revised Unit 1 Cycle 3 license submittal (Reference 2C),not the values quoted in the original Cycle 3 license submittal (Reference 2A).

. ~.

Page 35 7.1 CEA WITHDRAWAL EVEfiT The CEA withdrawal event was reanalyzed for Cycle 4 due to the increase in the Resistance Temperature Detector (RTD) response. time to envelope future cycles and the increase in the CEA drop time to 900 insertion in comparison to the reference cycle. The reference cycle for this event is the analysis uoon ubich the licensina of Calvert Cliffs Unit 2, Cvcla 2 (sae Rafaranca 12) was based. As stated in CErlPD-199-P (Reference 1), the CEA Withdrawal event initiated at rated thermal power is one of the DBEs analyzed to determine a bias factor used in establishing the TM/LP setpoints. This bias factor, along with conservative temperature, pressure, and power readings assures that the TM/LP trip prevents the D;;BR from dropping below the SAFDL limits (DilBR = 1.19 based on CE-1 correlation) for a CEA Withdrawal event. Hence, this event was analyzed for Cycle 4 to generate the bias term input to the TM/LP trip. The CEA Withdrawal transient may require protection against exceeding both the DNBR and fuel centerline melt (kw/ft) SAFDLs. Depending on the initial conditions and the reactivity insertion rate associated with the CEA withdrawal, either the Variable High Power Level or Thermal Margin / Low Pressure (TM/LP) trip reacts to prevent exceeding the Di;BR SAFDL. An approach to the kw/ft limit is terminated by either the Variable High Power Level trip or the Axial Flux Offset trip. The zero power case was analyzed to demonstrate that SAFDLs are not exceeded. For the zero power case, a reactor trip, initiated by the variable high power trip at 40% of rated thermal power, is assumed in the analysis. The key parameters for the cases analyzed are reactivity insertion rate due to rod motion and mnderator temperature feedback effects, and initial axiai power distribution. The Resistance Temperature Detector (RTD) response time is also important in deternining the pressure bias factor. The range of reactivity insertion rates considered in the analysis is given in Table 7.1-1, along with the values of other key parameters used in the analysis of this event. The maximum reactivity insertion rate for cycle 4 is 1.3x10 to/sec at all power levels. This reactivity withdrawal rate was calculated by combining the maximum CEA differential worth of 2.6x10-' I4:/ inch and the maximum CEA Withdrawal speed of 30 inches per minute. The initial axial power shape and the corresponding scram worth versus insertion used in the analysis of both cases is a bottom peaked shape. This power distribution maximizes the time required to terminate the decrease in DNBR following a trip. The CEA Withdrawal transient initiated at rated thermal power results in the maximum pressure bias factor of 62.0 psia. This bias factor accounts for measurement system processing delays during the CEA Withdrawal event. The pressure bias factor for this cycle has increased from the reference cycle due to the increase in the RTD time constant and the increase in the CEA drop time to 90% insertion. This pressure bias factor is used in generating TM/LP trip setpoints to prevent the SAFDLs from being exceeded during a CEA Withdrawal event. ,v m _ _. ,m.-.

Page 36 The zero power case initiated at the limiting conditions of operation results in a minimum DNBR of 1.71. Also, the analysis shows that the fuel centerline temperatures are well below those corresponding to the fuel centerline melt SAFDL. The Sequence of events for tne zero power case is presented in Table 7.1-2 Figures 7.1-1 to 7,1-4 presents the transient behavior of core power, core average heat flux, the RCS pressure and the RCS temperatures. The analysis of the CEA Withdrawal event presented herein, shows that the DNB and fuel centerline melt SAFDLs will not be exceeded during a CEA Withdrawal transient. m... e .c_ ~

Page 37 TABLE 7.1-1 KEY PARAMETERS ASSUMED Ill THE CEA WITHDRAWAL AttALYSIS Unit 2 Unit 1 Parameter Units Cyclo 2 Cycle 4 MWt 0,102% of 2700 0)l02% of 2700 Initial Core Power Level (H2P HfP) 3 Core Inlet Coolant Temperature (HAP,HFP) F 532)S50 532)550 Reactor Coolant System Pressure psia 2200 2200 -4 U Moderator Temperature Coefficient 10 Ap/ F +.5 +.5 .85 .85 Doppler Coefficient Multiplier -2 CEA Worth at Trip - FP 10 Ap -5.14 -5.14 10-2 Ap -3.4 -3.4 CEA Wei:S at Trip - ZP -4 Reactivity Insertion Rate X10 Ap/sec 2.0 1.3 0.5 0.5 Holding Coil Delay Time sec 2.5* 3.1 CEA Time to 90 Percent Insertion sec (Including Holding Coil Delay) Resistance Temperature sec 5.0 8.0 Detector Response Time (T) Rod Group Withdrawal Speed in/ min 30'.0 30.0 -4 Maximum CEA Differential Worth x10 Ap/ inch 4.0 2,6 i

• The reference cycle analysis assumed a CEA drop time to 90% insertion value of 2.5 seconds (see Reference 12a) but in a subsequent submittal (see Reference 12b) a CEA drop time to 90% insertion value of 3.0 seconds was justified for Unit 2, Cycle 2.

i q .,- c -

Page 38 TABLE 7.1-2 SEQUEtiCE OF EVEf;TS FOR CEA WITHDRAWAL FROM ZERO POWER Time (sec) Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion 37.7 High Power Trip Signal Generated 40% of 2700 MWt 38.1 Trip Breakers Open 38.6 CEAs Begin to Drop Into Core 39.0 Maximum Power Reached 126.4 of 2700 MWt 40.15 Maximum Heat Flux Reached 58.7 of 2700 MWt 40.15 Minimum DilBR Cccurs 1.71 fisximbm' Pressurizer Pressure Reached 2353 psia 42'.3'"'"" M ,g G

Page 39 140 i i i ZERO POWER 120 100

E s

N 80 8 5 60 a_ Mw s:8g 40 8 20 0 i

i 0

20 40 60 80 100 TIUtE, SECONDS BALTIMORE GAS & ELECTRIC CO. CEA WITHDRAWAL EVENT Figure 7 l~1 CORE POWER VS TIME Nuc e v

Plon,

Page 40 120 i i i i ZERO POWER 100 E s Ng 80 bs o-60 t. U y L! = E 40 u oa 20 0 i Ji 0 20 40 60 80 100 TIME, SECONDS BALT L'AORE CEA WITHDRAWAL EVENT Figure GAS & ELECTRIC CO. calvert cliffs CORE HEAT FLUX VS TIME 7 1-2 Nuclear Power Plant

Page 41 2400 i i i i ZERO POWER 2300 5 E ur 2200 5 m !O E 2100 m o <c I: t ce 2000 1900 i i i i 0 20 40 60 80 100 TIME, SECONDS BALTIMORr GAS & ELECTRIC CO. CEA WITHDRAWAL EVENT Figure RCS PRESSURE VS TITE 7.1-3 Coivert Ciirts Nuclear Power Plant

Page 42 600 i i i i ZERO POWER 580 O 5 5 560 Tout E e5 S Tavg W 540 h ,T.. -l[ 1 520 500 i i i 0 20 40 60 80 100 TIME, SECuNDS BALT LYiORE GAS & ELECTRIC CO. CEA WITHDRAWAL "S" RCS TEMPER ATURES VS TIME 7.1-4 Nuc Pov Plant

Page 43 7.1-2. RCS DEPRESSURIZATI0t! EVEriT The RCS Depressurization event was reanalyzed for Cycle 4 to assess the impact of it creasing the CEA drop time to 9D~. insertion from 2.5 seconds for Cycle 3 to 3.1 seconds for Cycle 4 As stated in CET'PD-199-P (Reference 1), this event is one of the DBEs analyzed to determine a bias term input to the Ti4/LP trip. Hence, this event was analyzed for Cycle 4 to obtain a pressure bias factor. This bias factor accounts for measurement system processing ceiays during this event. The trip setpoints incorporating a bias factor at least this large will provide adequate protection to prevent the D*iBR SAFDL from being exceeded during this event. The assumptions used to maximize the rate of pressure decrease and consequently the fastest approach to DriBR SAFDL's are:

1) The event is assumed to occur due to un i:,eavertant opening of both pressurizer relief values while operatir., rt rated tnermal powe.. This results in a rapid drop in the RCS pressure and consequently a rapid decrease in DilBR.

2) The initial axial power shape and the corresponding scram worth versus insertion used in the analysis is a bottom peaked shape.

This power distribution maximizes the time required to terminate the decrease in DilBR following a trip.

3) The charging pumps, the pressurizer heaters and the pressurizer backup heaters are assumed to be inoperable. This maximizes the rate of pressure decrease and consequently maximizes the rate of approach to DNBR SAFDL.

The analysis of this event shows that the pressure bias factor is 35 psia which is less than that required by the CEA Withdrawal event. Hence, the use of the pressure bias factor determined by the CEA Withdrawal event will prevent exceeding the SAFDLs during an RCS Depressurization event. ..._m .. _ - _ n. -n,.

Page 44 7.3 LOSS OF COOLANT FLOW EVENT The loss of Coolant Flow event was reanalyzed for Cycle 4 to detemine the impact on margin requirements that must be built into the Limiting Conditions for Operations (LCOs) due to the increase in the CEA drop time to 90%' insertion. The methodology used to evaluate this event is identical to that employed The nethodoloov in'the Unit 1, Cycle 3 license submittal (Referencautilizes the computer code STRIKIN 11 ( dependent hot channel and core average heat fluxes distributions during the transient. For conservatism credit for the heat flux decay was taken only for axial power distributions for which the initial minimum DNBR was located in an axial region of the core where the scram rods have passed the axial node of mininum DNBR before the time at which minimum DNBR is reached. For those axial ; ower distributions analyzed that did not meet the above criterion, the methodology utilized is consistent with CEUPD-199-P (Reference 1). The computer code TORC (Reference 4) was used for all DNBR calculations. This is consistent with the methods used by C-E to calculate the DNB margin requirements. The TORC code and its application to the analyses were reviewed and apprpved by fiRC in response to the revised Unit 1 Cycle 3 license sub-mittal (Reference 2). The 4-Pump Loss of Coolant Flow produces a rapid approach to the DNBR EAFDL due to the rapid decrease in the core coolant flow. Protection against exceeding the DNBR SAFDL for this transient is provided by the initial steady state themal: margin.which i_s-assured by-maintaining the. technical u specifications' LCOs on DNBR margin and by the response Ef the RPS~which ~~ ~ ~ provides an automatic reactor trip on low reactor coolant flow as measured by the steam generator differential pressure transmitters. The transient is characterized by the flow coastdown curve given in Figure 7.3-1. Table 7.3-1 lists the key transient parameters used in the present analysis. Table 7.3-2 presents the NSSS and RPS responses during a four pump loss of flow initiated at the most negative shape index (Ip) allowed by the LCOs. The low flow trip setpoint is reached at 1.0 seconds and the scram rods start dropping into the core one second later. A minimum CE-1 DNBR of 1.25 is reached at 2.3 seconds. Figures 7.3-2 to 7.3-5 presents the core power, heat flux, RCS pressure, and core coolant temperatures as a function of time. Figures 7.3-6 presents a trace of hot channel DNBR vs time for the limiting case that is characterized by an Ip = .15. The low flow trip, in conjunction with the Initial Overpower Margin maintained by the LCOs in the Technical Specifications assure that the minimum DNBR will be greater than or equal to 1.19 for the Loss of Coolant Flow event. + z

Page 45 TABLE 7.3-1 KEY PARAMETERS ASSUi1ED Irl THE LOSS OF C00LAtiT FLOW At:ALYSIS Parameter Uni ts Unit 1, Cycle 3 Unit 1, Cycle 4 Initial Core Power Level (11Wt) 102% of 2700 102% of 2700 Initial Core Inlet Coolant Temperature ( F) 550 550 6 Initial Core Mass Flow Rate (10 lbm/hr) 134.22 135.24 Reactor Coolant System Pressure (psia) 2200 2200 Initial Steam Generator Pressure (psia) 861 861 Moderator Temperature Coefficient (10-4 Ap/F) +.5 +.5 .85 .85 Doppler Coefficient Multiplier LFT Response Time sec 0.5 0.5 CEA Holding Coil Delay sec 0.5 0.5 CEA Time to 90% Insertion sec 2.5 3.1 ' including Holding Coil Delay) ' -2 ('10 ff,{ ' - -5.7 ' t,tA Worth' at Trif -2' -5.7 T Total Radial Peaking Factor (F ) 1.65 1.58 4-Pump RCS Flow Coastdown Figure 7.1-1 of Figure 7.3-1 Reference 2

Page 46 TABLE 7.3-2 SEQUENCE OF EVENTS FOR LOSS OF FLOW Time (sec) Event Setpoint or Value 0.0 Loss of Power to all Four Reactor Coolant Pumps 1.0 Low Flow Trip 38% of 4-Pump Flow 1.5 Trip Breakers Open 2.0 Shutdown, CEAs begin to Drop into Core 2.3 Minimum CE-1 DNBR 1.25 5.5 Maximum RCS Pressure, psia 2276 v. .Y -e,-- 4NT M *7M 6 MM-

• !"84E U S ***

. - - - =

Page 47

1. 0 i

i i i 4-PUMP COASTDOWN

0. 8 zop

0. 6 O

<C E s: o cd 0.4 1---------------- m 1 u.i &O O 0.2 O i i i - i 0 4.0

8. 0 12.0 16.0 20.0 TIME, SECONDS BALTIMORE Fi ure S

GAS & ELECTRIC CO. LOSS OF COOLANT FLOW EVENT calvert ci;rts 7.3-1 CORE FLOW FRACTION VS TIME Nuclect Power Plant

Page 48 120 i i i .i 100 5 8m 80 E5 5 o5 60 a. EI .g ~%, _c ~ a 40 w c:oo 20 0 i i i i 0 4 8 12 16 20 TIME, SECONDS ^ SS OF COOLANT FLOW EVENT Fi 7~ jure GAS EE C O. coivert cliffs CORE POWER VS TIME Nuclear Power Plant -,~...

Page 49 120 i i i i 100 R s B3 80 u_ o W5 o f5 a 60 >l3 u_ W h 40, .w ) eoa 20 0 i i i i 0 4 8 12 16 20 TIME, SECONDS BALTIMORE GAS & ELECTRIC CO. LOSS OF COOLANT FLOW F' 9"

7. 3-3 Calvert Cliffs CORE HEAT FLUX VS TIME Nuclear Power Plant

Page 50 2400 i i i i 2300 sm o 2200 uf Mm ww LA.1 e 2100 o_ wa E 6._. r ~ t 2000 1900 i i i i 0 4 8 12 16 20 TIME, SECONDS BALTLMORE Fi " GAS & ELECTRIC CO. LOSS OF COOLANT FLOW EVENT S RCS PRESSURE VS TIN,E 73 ~4 Calvert Ciirrs Nuclear Power Picnt ~, -, _,. - _ = -

Page 51 620 i i i i 600 Tout o' d 580 E q Tave e5 a a t

E 560 W

w C) 540 520 i i i i 0 4 8 12 16 20 TINE, SECONDS ^ GAS EE C O. LOSS OF COOLANT FLOW EVENT "9"' caiven cliffs Nuclear Power Plant RCS TEMPERATURES VS TIUtE

7. 3-5

. - - = - -

Page 52 2,0 1,9 p 1.8 tiid 1,7 EE 1,6 ci

==5 1.5 5 h 1.4 a: z H2 1.3 E -= 1,2 1,1 1.0 0 2 4 6 8 30 TIME (SECONDS) BALT LMORE GAS & ELECTRIC CO. LOSS OF COOLANT FLOW EVENT Figure Calvert Cliffs Nuclect Power Plant MINIMUM HOT CHANNEL DNBR (CE-1) VS TIME. 7.3-6

Page 53 7.4 CEA EJECTION EVENT The CEA Ejection event was reanalyzed for Cycle 4 to assess the impact of increasing the CEA drop time to 904 insertion and the increase in the augmentation factor in com rison to the reference cycle. In addition, the zero power case was analyzed due to the decrease in axial peak in comparison to the reference cycle. The reference cycle for this event is the analysis upon which the licensine of Calvert Cliffs Unit 2, Cycle 2 (see Reference 12) was based on. Hence, this event was reanalyzed to demonstrate that the criterion for clad damage is not exceeded during Cycle 4 operation. To bound the most adverse conditions during the cyc'e, the most limiting of either the Beginning of Cycle (B0C) or End of Cycle (E0C) value was used in the analysis. A BOC Doppler defect was used since it produces the least amount of negative reactiv'ty feedback to mitigate the transient. A BOC moderator temperature cuefficient of +0.5x10-4 op/0F was used which results in positive reactivity feedback with increasing coolant temperatures. A E0C Beta Fraction was used in the analysis to produce the highest power rise during the event. For the full power and zero power cases, the axial power distributions were selected to yield conservative results. The corresponding shape indices were -0.21 for the full power case and -0.40 for the zero power case. These shapes are conservative with respect to the most negative shape indices allowed by the DNBR monitoring band. This is consistent since the power shifts to the top of the core af ter the CEA ejects. The reactivity-forced power transient was simulated by a digital computer program, CHIC-KIN (Refererce 5), which simultaneously solves the one group neutron point kinetics equations together with the time and space dependent thermal and hydraulics equations for heat generation and transport within a single channel. The kinetics model incorporates the standard six-delay group representation along with explicit reactivity contributions from: (a) CEA motion, (b) Doppler effect, and (c) moderator density variations. By simulating the core average channel, the CHIC-KIN code computes the core average integrated energy output during the course of the transient. In the CEA Ejection event, the principal reactivity feedback mechanism affecting the power transient is the Doppler feedback. In the point kinetics approach, utilized in CHIC-KIN, a spatial Doppler weighting factor (k) accounts for the fact that the Doppler feedback effect is a function of the spatial flux distribution. In order to represent the radial Doppler effect in a conservative manner, a space-time analysis was performed in which point kinetics calculations for various radial slices were compared with time-dependent, two-dimensional diffusion theory results obtained with a C-E modified version of the TWIGL code (Reference 6). The results of the space-time analysis have demonstrated that the use of the static (non-Doppler flattened) radial fuel rod peaking factor, as obtained from two-dimensional diffusion theory calculations, in conjunction with the average hot spot energy releases, yield energy increases that are conservatively large. Radial Doppler weighting factors obtained as a function of the ejected CEA worth are defined such that CHIC-KIN and TWIGL results give the same total core energy release. The average energy rise in the hottest fuel pellet is obtained from the following relationship: g (P/A)H

• Ave HT AE xK

-E =

Page 54 Where /2EAye is the average core energy rise obtained from CHIC-KIN; (P/A)g (the three-dimensional fuel rod peaking factor) is the ratio of the hot spot power density to the core average power density obtained from static, non-Doppler flattened diffusion theory calculations; K is the reduction factor defined above. For the zero power case, it is conservatively assumed that EHT, which accounts for heat transferred out of the fuel rod during the transient, is zero. The average energy in the hottest fuel pellet at the beginning of the transier.t is added to the net average energy rise in the hottest fuel pellet as obtained from Equation (7.4-1) to determine the total average enthalpy in the hottest fuel spot in the core. A similar procedure is used to compute the total centerline enthalpy in the hottest spot. The initial energy is obtained by correlating the initial local fuel temperature with an empirical temperature enthalpy relationship (see Reference 7). The spatial variation of the core local-to-average power ratio results from the convolution of the axial power distribution with radial pin power census distributions for the post-ejection condition, which are based on static core physics calculations. Combining these results with the total average and centerline enthalpies in the hottest fuel spot yields the fractional number of fuel rods with specific total average and centerline enthalpies. The calculated enthalpy values are compared to threshold enthalpy values to determine the amount of fuel experiencing the various degrees of fuel damage. These threshold enthalpy values are (References 8, 9, and 10). Ciad 'D'amage' Thr'ssh6[d: '~ ' Total Average Enthalpy = 200 cal /gm Incipient Centerline Melting Threshold: Total Centerline Enthalpy = 250 cal /gm Fully Molten Centerline Threshold: Total Centerline Enthalpy = 310 cal /gm The criterion for determining the fraction of fuel rods that will release their radioactive fission products during a CEA ejection is the same as the one quoted above for determining clad damage. Thus, it is assumed that any fuel rod that exceeds a total average enthalpy of 200 cal /gm releases all of its gap activity. The gap activity corresponding to the hottest fuel rod during the core cycle is conservatively assumed for each rod that suffers clad damage. The zero power CEA ejection event was analyzed assuming the core is initially operating at 1 MWt for conservatism. At zero power, a Variable Overpower trip is conservatively assumed to initiate at 40% (30% + 10% uncertainty) of 2754 MWt and terminates the event. m ~. - _ ~ ~

Page 55 The full and zero power cases were analyzed, tssuming,a value of 0.05 seconds for the total ejection time, which is consistbnt with the FSAR. Table 7.4-1 lists all the key paraneters used in this analysis. Table 7.4-2 presents the results of the two ejection cases analyzed for Cycle 4 in comparison to the reference cycle. As seen from Table 7.4-2, the average energy deposited for both the full and zero power cases.has increased. The increase for the full power case is due to the increase in CEA drop time to 90% insertion in comparison to the reference cycle analyses; while for the zero oower case, the decrease in the axial peak offsets the increase in the CEA drop time to 90% insertion. The power transient produced by a CEA ejection initiated at the maximum allowed power is shown in Figure 7.4-1, and at zero power is shown in 7.4-2. Since the criterion of clad damage (i.e., less than 200 cal /gm) is not exceeded for either the full or zero power CEA ejection, no fuel pins are predicted to fail. o Ms. .v. o n "b O ,*SMM

TABLE 7.4-1 Page 56 KEY PARAMETERS ASSUT'ED IN THE CEA EJECTIOi ANALYSES Unit 2 Unit 1 Parameter UtiTTS Cycle 2 Cycle 4 Full Power Core Power Level MWt 2754 2754_ Core Average Linear Heat Rate kw/ft 6.23 6.12 of Fuel Rod -4 Moderator Temperature Coefficient 10 Ap/ F +.5 .5 + E.iected CEA Worth % ap .32 +.32 Delayed Neutron Fraction, 8 .0047 .0047 Post-Ejected Radial Power Peak 3.36 3.36 Axial Power Peak 1.39 1.39 CEA Bank Worth at Trip % ap 3.88 3.88 Augmentatior Factor 1.060 1.069 K-Factor .92 .92 Tilt Allowance 1.03 1.03 CEA Drop Time to 90% Inserted sec 2.5* 3.1 Zero Power Core Power Level tiWt 1. 1.0 K-factor .89 .89 Ejected CEA Worth % ap .60 .60 Post-Ejected Radial 9.83 9.83 Power Peak Axial Power Peak 2.0 1.6 CEA Bank Worth at Trip % ap 2.58 2.58 Tilt Allowance 1.10 1.10 CEA Drop Time to 90% Inserted 2.5* 3.1

• The reference cycle analysis assumed a CEA drop time to 90% insertion value of 2.5 seconds (see Reference 12a) but in a subsequent submittal (see Reference 12b) a CEA drop time to 901 insertion value of 3.0 seconds was justified for Unit 2, Cycle 2.

-~~n,g._,.....,.

Page 57 TABLE 7.4-2 CEA EJECTION RESULTS Unit 2 Unit 1 Full Power Cycle 2 Cycle 4 Total Average Enthalpy of Hottest 193 198 Fuel Pellet (cal /gm) Total Centerline Enthalpy of 263 268 Hottest Fuel Pellet (cal /gm) Fraction of Rods that Suffer Clad 0.0 0.0 Damage (Average Enthalpy 1 200 cal /gm) Fraction of Fuel Having at least 0.01 .01 Incipient Centerline Melting (Centerline Enthalpy 1 250 cal /gm) Fraction of Fuel Having a Frily 0.0 0.0 Molten Centerline Condition (Centerline Enthalpy 1 310 cal /gm) t...- Zero Power Total Average Enthalpy of Hottest 173 177 Fuel Pellet (cal /gm) Total Centerline Enthalpy of 173 177 Hottest Fuel *allet (cal /gm) Fraction of Rods that Suffer 0.0 0.0 Clad Damage (Average Enthalpy 1 200 cal /gm) Fraction of Fuel Having at least 0.0 0.0 Incipient Centerline Melting (Centerline Enthalpy 1 250 cal /gm) Fraction of Fuel Having a Fully 0.0 0.0 Molten Centerline Condition (C5nterline Enthalpy 1 310 cal /gm)

Page 58 3.0 i i i i FUI.L POWER H

2. 0 0

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• GAS & ELECTRIC CO.

CEA EJECTION EVENT Ncc e r v Plo r.t CORE POVvt2 VS TINE 74-1 W -,r,4L'---- v-Y~ ~~

Page 59 10.0 i i i ZERO POWER

1. 0 R

8 m uO zOp 0.1 o e u_ i LU . B: -1 I,- .g t o-tu Of Oo 0.01 0.001 i i i i 0 2 4 6 8 10 TIME, SECONDS BALTIM OR' GAS & ELECTRIC CO. CEA EJECTION EVENT F'"' 7.4-2 coivert cliffs CORE POWER VS TITE Nuclear Power Plant

Page 60 7.5 SEIZED ROTOR EVEtiT The Seized Rotor event was reanalyzed for Cycle 4 due to the changes in the following key parameters.

1) The increase in the CEA drop time to 90% insertion

2) The decrease in core bypass flow, which increases the net core flow

3) The decrease in the Radial Peaking Factor

4) A more adverse (flatter) pin census.

The increase in the CEA drop time and the flatter pin census adversely impact the consequences of this event. Increasing the net core flow and decreasing the Radial Peaking Factor will succor the consequences of this event. Hence, a reanalysis was performed for Cycle 4 to ensure that only a small fraction of fuel pins are predicted to fail during a Seized Rotor event. The Seized Rotor event is assumed to be initiated by the complete seizure of a shaft in one of the reactor coolar.t pumps. This reduces the core coolant flow rapidly to the 3 pump flow. In the analysis of the event, it is conser-vatively assumed that the core flow is instantaneously reduced to 3 pump flow. This initiates a reactor trip on low primary coolant flow (93% of 4 pump flow) which terminates the decrease in the DNBR. The methods used to analyze this event are identical to those reported in the Unit 1. Cvele 2 Stretch. Power license submittal (Reference 11). The techniques and methods used to calculate the number of fuel pins experiencing DNS.is discussed in detail in Section 7.1 of the Calvert Cliffs Unit 1 Cycle 3 revised license submittal Nference 2) except TORC /CE-1 instead of COS!10/W-3 was used to calculate the DNBR. The key transient parameters used in this analysis are compared to the reference cycle analysis in Table 7.5-1. The NSSS and RPS response for the Seized Rotor event initiated at an Ip = .15 is shown in Table 7.5-2. Figures 7.5-1 through 7.5.-4 show the core power, core average heat flux, RCS Pressure, and RCS temperatures as a function of time during this event. A conservatively " flat" pin census distribution (a histogram of the number of pins with radial peaks in intervals ' of 0.1 in radial peak normalized to the maximum peak) was used to determine the number of pins that experience DNB. The results indicate that increasing the core flow and decreasing the radial peaking factor offset the increase in the CEA drop time to 90% insertion. It was calculated that for Cycle 4, less than.5 percent of fuel pins will experience DNB for even a short period of time. For the case of the loss of coolant flow arising from a seized rotor shaft, it is assumed that there is an instantaneous reduction to 3 pump flow. The low flow trip assures that less than.5% of fuel pins experience DNB. This is the same as that calculated for the Reference cycle. Hence, the conclu-sions reached for reference cycle remain valid for cycle 4. l

Page 61 TABLE 7.5-1 ASSUP.PTIONS FOR SEIZED ROTOR II;CIDEtiT Unit 1, Unit 1, Parameter Units Cycle 3 Cycle 4 Initial Core Power Level MWt 102% of 2700 102% of 2700 Core Inlet Coolant F 550 550 Temperature 0 Four Pump Core Mass Flow Rate 10 lbm/hr 134.22 135.24 (2200 psia, 550 F) 6 103.4 104.44 10 lbm/hr Three Pump (Core Mass 2200 psia, 550 F) Flow Rate Reactor Coolant System psia 22G3 2200 Pressure Steam Generator Pressure psia 861 861 -4 Moderator Temperature 10 Ap/ F +.5 +.5 Coefficient Doppler Coefficient Mul- .85 .85 Tiplier CEA Wor,th.on Trl,p -. -10d,t;p ._ _. 3; . - _r. - 5 7:,,_" [' ' 1. i ; - 5. 7 5 CEA Drop time to sec 2.5 3.1 90% insertion

Page 62 TABLE 7.5-2 SEQUENCE OF EVEllTS FOR SEIZED ROTOR Time (set) Event ' Setpoint or Value 0.0 Seizure of One Reactor Coolant Pump 0.0 Low Coolant Flow Trip 38% of 4-Pump Flc 0.1 Dump Valve Opens 0.5 Trip Breakers Open 1.0 Shutdown CEAs Begin Dropping into Core 3.5 Maximum RCS Pressure 2277 o w e -.4 mrg ye

Page 63 120 i i i i -3 100

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Es s 80 5 o5 o_ g' 60 E2 N 8 40 U 20 O r i i i 0 4 8 12 16 20 TIME, SECONDS BALT LY, ORE GAS & ELECTRIC CO. SEIZED ROTOR E'!ENT "9" Colvert Cliffs CORE POWER VS TINE

7. 5 -1 Nuclear Power Plant

.~

Page 64 120 ~ i i i 100 h am 80 u_ O $u 5 60 c_ >l . :"u_.,_[ }- Q 40 1 Lu e Oo 20 0 i i i i 0 4 8 12 16 20 TIME, SECONDS BALTIMOP GAS & ELECTRid"CO. SEIZED ROTOR EVENT Fi "' S 7.5-2 coivert cliffs CORE HEAT FLUX VS TIME Nuclear Power Plant .,,w--,.v n

Page 65 2400 i i i i 2300 5 m 2200 uf xaw I __ I. h gr 2100 mox 2000 1900 i i i i 0 4 8 12 16 20 TIME, SECONDS I BALTIMORE Figure GAS & ELECTRIC CO. SEIZED ROTOR EVENT Calvert Cliffs Nuclear Power Plan, RCS PRESSURE VS TIME 7' 5-3

Page 66 620 i i i i 600 O 580 m Tout wx 0F-<m Tavg w' 560

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~ m Tin o e t-k. .u. 540 '~ 520 i 0 4 8 12 16 20 TIME, SECONDS BALTIMORE Fi ** GAS & ELECTRIC CO. SEIZED ROTOR EVENT S bont RCS TEMPERATURES VS TIME

7. 5 -4 Nuc e w

~ Page 67 o 8.0 ECCS Analysis Introduction and Summary

8..1 The ECCS performance evaluation for Calvert Cliffs Unit I, Cycle 4, presented herein demonstrates appropriate confunaance with the Acceptance Criterie for Light-Water-Cooled Reactors as presented in 10CFR50.46(I)

The evaluation demonstrates acceptable ECCS performance for Calvert Cliffs Unit I, during The method cycle 4, at a peak linear heat generation rate of 14.2 kw/ft. of analysis and results are presented in the following sections. 8.2 Method of Analysis The method of analysis consisted cf a comparison of the fuel specific parameters for the limiting fuels in cycles 3 and 4. The comparison demonstrates that the limiting fuel during cycle 4 has a much lower stored energy (i.e. higher gap conductance) than the limiting fuel identified in the cycle 3 analysis of Reference 8 As a result, the peak clad temperature and local clad oxidation (STRIKIff-II)@,5 ) calculations performed for cycle 3(89 are conservative and applicable In addition, the blowdown (CEFLASH-4A)(2), refill for cycle 4. (COMPERC-II)( }, and core wide oxidation (COMZIRC)(, sup.1) analyses frcm the cycle 2 analysis (7 ) remain valid for cycle 4. Therefore, the ECCS performance resuits reported in Reference 8 for cycle 3 are also applicable to cycle 4. The comparison of the fuel specific parameters supporting this method of analysis for cycle 4 is presented below. 8.3 Resul ts The cycle 4 core contains 216 high density fuel assemblies and one low density Batch B assembly. The highest power pin in the low density Batch B I - - ~ ~

Page 68 assembly will not achieve a power level greater than 75% of the highest power pin in the core. Therefore, a Batch B fuel pin will not be limiting in cycle 4. The remaining 216 high density fuel assemblies contain 72 partially depleted Batch D assemblies, 72 partially depleted Batch E assemblies and 72 fresh Batch F assemblies. Burnup dependent calculations were performed for the high density fuel assemblies with the FATESb ) and STRIKIft-II(4,5) codes. The results demonstrate that the most limiting fuel pin aring cycle 4 is located in one of the partially depleted Batch E assemblies. Table A.1 cornnares the fuel specific narareters which correspond to the limiting fuels in cycle 3 and cycle 4. As shown in the table, the 0 limiting high density fuel in cycle 4 has a stored energy 268 F lower than the limiting fuel in cycle 3. Consequently, the ECCS performance results reported for cycle 3 conservatively bound the performance for cycle 4.

82. 41 Conclusion The comparison between the fuel specific parameters for the limiting fuels in cycles 3 and 4 demonstrates that tha cycle 3 ECCS performance analysis conservatively bounds the performance for cycle 4.

Therefore, the peak linear heat generation rate of 14.2 kw/ft which was demonstrated to be acceptable for cycle 3 is also an acceptable limit for cycle 4 opera tion. Conformance of this evaluation is the same as stated in Reference 8. The statements in Reference B demonstrating compliance with Criterion 4 (coolable geometry) and Criterion 5 (long term cooling) remain unchanged. is n"esented in "eferem er C eni ",, the n all Fre w cze not limitine. -, w u u,,. n.- n e n, ~. =. ..; -n ~_.m.,un ~ -.

Page E9 8.5 Computer Code Version Identification Version 77063 of the STRIKIti-II code of Combustion Engineering's ECCS Evaluation Model was used to perform the burnup dependent calculations in evaluating the fuel data. 0 O e i

Table 8 ]_, Calvert Cliffs I Cycle IV Core Parameters Quantity Value Cycle III Cycle IV Batch B Batch E Gap Conductance at PLHGR 859 1426 BTU /hr-ft - F Fuel Centerline Temperature at PLIIGR 3692 3405 F Fuel Average Temperature at PLHGR 2419 2151 F llot Rod Gas Pressure 1221 1026 psia llot Rod Burnup (Minimum IIGAP) 9170 1522 MWD /MTU l, i W

Page 71 9. TECHNICAL SPECIFICATI0t15 In this section all changes that roust be made to the Technical Specifications are provided in order to make the Technical Specifi-cations valid for operation of Cycle 4. Each page from the Technical Specifications which must be modified is shown with the modification included. Ex6MPLE : 3/4 X.X The exS M V M is

1. O(,

l,[D g 4 wd m b-M m

Page 72 INDEX (m LIMITING CONDITIONS FOR OPERATIC'! AMD SURVEll! AUCE REOUIRE!'ENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE..................................... 3/4 2-1 3/4.2.2 TOTAL PL AliAR RADI AL PEAKiliG FACT 0R................... 3/4 2-6 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR............... 3/4 2-9, 3/4.2.4 AZIMUTHAL PONER TILT................................. 3/4 2-12 rn m n xn = m..@.M.5.0.................... me 1 2/4.2.: 3/4.2.6 DNB PARAMETERS....................................... 3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PRO.TECTIVE INSTRUMENTATION................... 3/4 3-1 ("% 3/4.3.2 EliGINEERED SAFETY FEATURE ACTUATION SYSTEM I list RUM E NT AT 10 f t.................................... 3/4 3-10 3/4.3.3 M0111TORING INSTRUMENTATION Radiation Monitoring Instrumentation................. 3/4 3-25 I n c e r t D e t e c t o rs..................................... 3/4 3-29 Seismic Instrumentation.............................. 3/ 4 3-31 Meteorological Instrumentation....................... 3/ 4 3-34 Remote Shutdown Instrumentation...................... 3/4 3-37 Post-Accident Instrumentation........................ 3/4 3-40 Fire Detection Instrumentation....................... 3/4 3-43 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR C00LAtiT L00PS................................ 3/4 4-1 3/4.4.2 S A FE T Y VAL V E S - S HU T D 0',!N............................. 3/4 4-3 3/4.4.3 SAFETY VALVES - 0PERATING............................ 3/4 4-4 CALVERT CLIFFS - UNIT 1 IV Amendment No. 27. /II' . ~n.- -,, .n -n

Page 73 r g ' DEFINITIONS P. CHANNEL CHECK 1.10 behavior during operation by observation.A CHANNEL CHECK s 3 This determination shall include, where possible, comparison of the thannel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be: Analog channels - the injection of a simulated signal into a. the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions. b. Bistable channels - the injection of a siralated signal alarm and/or trip functions.into the channel sensor to verify O CCRE ALTERATION f 4 (gns 1.12 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head ( removed and fuel in the vessel. not preclude completion of movement of a component to a safeSuspen conservative position. l _ SHUT 00WN MARGIN r 1.13 SHUTCC'.lN MARGIN shall be the instantaneous amount of reactivity i which the reactor'subcritical[1s]or would be subcritical from its present condition assuming: i [11 full length control element assemblies (shutdown and 5(. regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdraun g 2 change #r prt length centrol Gement-ascmbl3 pe+i-t4cm-O .R CALVERT CLIFFS-UNIT I 4 l-3 Si ~ 6

Pb e73^ S DEFINITIONS RrACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTo.1 TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism. ENGINEERED SAFETY FEATURE RESPONSE TIME 1.26 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discnarge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the funda-mental nuclear characteristics of the reactor core and related instrumen-tation and 1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission. UNRODDED INTEGRATED RADIAL PEAKING FACTOR - F r 1.28 The UNRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the peak pin power to the average pin power in an unrodded core, excluding tilt. LOAD FOLLOW OPERATION 1.29 LOAD FOLLOW OPERATION shall involve daily power levei changes of more than 10% RATED THERMAL POWER or daily insertion of control rods below the long term insertion limits. =- ent No. )(

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~ Page 75 ( I i [ SArEly Lin115

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v-l f '. c.. c M 1 :... t ;. a. ; ; ; ' '":7 g u t ;. t-hc L._ The conditions f or the Thennal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shown on the figures. 1 The reactor protective system in combination with the Limiting ? Conditions for Operation, is designed to prevent any anticipated combina-t tion of transient conditions f tr reactor coolant system temperature, l pressure, and THERMAL PO'..'ER level that would result in a D!!BR of less f than 1.19 and preclude the existence of flow instabilities. 2.1.2 REACTOR C00LtJiT SYSTEM PRESSURE prr; The restriction of this Safety Limit protects the integrity of the k keactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching i the containment att'osphere. k I The reactor pressure vessel and pressuri:er are designed to Section f I III, 1967 Edition, of the ASME Code for fiuclear Power Plant Components which permits a maximum transient pressure of 1101 (2750 psia) of design j The Reactor Coolant System piping, valves and fittings, are iF pressure. ) designed to A!!51 B 31.7, Class 1, 1969 Edition, which permits a maximum transient pressure of Il0s (2750 psia) of component design pressure. g i i The Safety Limit of 2750. psia is therefore consistent with the desion l criteria and associated code requirements. j w. The entire Reactor Coolant System is hydrotested at 3125 psia to I demonstrate integrity prior to initial operation. 5 2

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! F: CALVERT CLIFFS - UNIT 1 B 2-3 Amendment lio.gg R t i-k. 1

1 I Page 76 (' ' 2.2 LIMIT 1HG SAFETY SYSTEM SETTINGS BASES 1 U 2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values The Trip Setpoints at which the Reactor Trips are set for each parameter. have been selected to ensure that the reactor core and reactor coolant Operation with system are prevented frce exceeding their safety limits. a trip set less conservative than its Trip Setpoint but within its speci-i fied Allowable Value is acceptable on the basis that y Allowable Value isequaltoorlessthanthedriftallowanceassumed for each trip in the 4 safety analyses. the M ererce between the trip set,_Jint and the Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. (O Power Level-Hiah 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 Level-High trip setpoint is operator adjustable and can be set no higher than 10'; above the indicated THERMAL POWER level. Operator action is required to increase the trip setpoint as THER.".AL PO'.lER is high increased. The trip setpoint is automaticall decreased as THERMAL cower. powe i

kpximum, decreases. The trip setpoint has a maximum lue of/106.E of RATED 107.05l THERMAL PC'!ER and eg =

lue the possible variation in trip point due to l' setpoint of 30'; of RATED THERMAL POWER. ! Low u Adding to this maximum va Powe calibration and instrument errors, the maximum actual steady-state THERitAL POWER level at which a trip would be actuated is 112*', of RATED g THERMAL POWER, which is the value used in the safety analyses. 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 I Provisions have been made in the reactor protective system to permit flow. k CALVERT CLIFFS - UNIT 1 B 2-4 sumuseup _ w P g

i Page 77 ), L.. LIMITING SAFETY SYSTEM SETTINGS W BASES i~ Steam Generator Water Level I The Stear' Generator Water Level-Low trip provides core. protection l' by preventino operation with the steam generator w.'ter level below the f~ minimum volume reen red for adequate heat removal capacity and assures that the i _ ^ pressuregof the reactor coolant system will not be exceeded. The specifiec setpoint provices allowance that there will safgty i be sufficient water inventory in the steam generators at the time of lWt trip to provide a margin of more than 13 minutes before auxiliary feedwater is required. {

a.. -

Axial Flux Offset j ff., The axial flux offset trip is provided to ensure that excessive axial peaking will not cause fuel damage. The axial flux offset is detennined from the axially split excore detectors. The trip setpoints ensure that neither a DiGR of less than 1.19 nor a peak linear heat rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip set- -( [ points were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated 't ; with the excore to incore axial flux offset relationship. b Thermal Marcin/ Low Pressure The Thennal Margin / Low Pressure trip is provided to prevent operation I t when the DIG? is less than 1.19, c- ~'d ~' _,4 9, _'d m-.'+ '::;' e': . n ; ' t y, __ The trip is initiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, and the number of' reactor coolant ptrups operating. The minimu value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT an; the maximu-CEA i deviation pennitted for continuous operation are assumed in the genera-I tion of this trip function. In addition, CEA group sequencing in accor-dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the j maximum insertion of CEA banks which can occur during any anticipated 0 operational occurrence prior to a Power Level-High trip is assumed. I .~ CALVEF.T CLIFFS - U:,ii 1 B E-6 Amendment No. g ' ~ ~ M 9 3 r ,$ a W ~~

Page 78 LIMITING SAFETY SYSTi" SETilNGS BASES The Thermal Marcin/Lov. Pressure trip setpoints are derived from the core safety limits through application of aporopriate allowanccs for equipment response time, measurement unccrtainties anc proccssin'g error. A safety margin is proviced which includes: an allowance of 5" of RATED THEP."AL POWER to compensate for potential power measurement error; an allowance of 2 F to compensate for potential temperature measurement uncertainty' and a further allowance of b psia to compensate for l84 pressure measurement error, trip system processino error, and tima delay associated with providing effective termination of the occurrence that exhibits the most rapid decreas,e.in margin to the safety limit. ThelllR 84 psia allowance is made up of a 22 psia pressure measurement allowance and a L12Jpsia time delay allowance. 62 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 (PN of the main steam line safety valve., 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 et the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. Rate ' Chance of Power-High The Rate of Change of Power-High trip is provided to protect the core during startuo operations and its use serses as a backup to the administra-tively enforced startup rate limit. Its trip setpoint does not correspond to a Safety Limit and 60 credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is recuired to enhance the overall reliability of the Reactor Protection Syste.a. s CALVERT CLIFFS - UNIT 1 B 2-7 Amendment No. 2J, 32 -. - ~ ~

Fage 79 f REi,CTIVITY CD:; TROL SYSTE"5 CE A DROP TIME L IMIT1MG CD';DIT10'i FOR OPERAT10' 3.1.3.4 The individual full !cngth (thutdwin and control) CEA drop time, fro:n a fully withdrawn position, shall be 1 17.5] seconds from when the l 3.1 electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with: a. T 3 515 F, and avg b. All reactor coolant pumps-operating. APPLICABILITY: MDDES 1 and 2. ACTIO :: With the drop time of any full length CEA determined to exceed a. the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2. D L'ith the CEA drop times within limits but determined at less than b. full rcacter coolant flow, operation may proceed provided THERMAL P05lER is restricted to less than or equal to the maximum THER"AL PC. ER level allowable for the reactor coolant pump combinatio i operating at the _ time of CEA drop time d et ermina tion. SURVE!LL A';CE RE001REMD:TS ' The CEA drop time of full lenoth CEAs shall be demonstrated 4.1.3.4 through measurement prior to reactor criticality: For all CEAs following each removal of the reactor vessel head, a. For specifically affected individual CEAs following any main-tenance on or r.odification to the CEA drive system which could b. affect the drop time of those specific CEAs, and At least once per 18 months. c. m CALVERT CLIFFS-U:UT 1 3/4 1-23 Amendment ?!o. 32 ~.__

Page 80 POWER DISTRIGUT10l1 LIMITS - e f V _SURVEILLA!iCL REQUIREMEriT5 (Continued) 'L._ t c. Verifying at least once per 31 days that the AX1AL SHAPE liiDEX is maintained within the limits of Figure 3.2-2, where 100 j. percent of the allowable power represents the maximum THER'iAL POWER allowed by the following expression: I j, 1_ h t. 'l I MxN g {- where: rg 1. M is the maximum allowable THERMAL POWER level for the = existing Reactor Coolant Pump combination, a ~ui 2. U is the maximum allorable fyaction of RATED THER'iAL 'n POWER as determined by the F curve of Ficure 3.2-3. ^ xy 4.2.1.4 Incore Detector Monitorint System - The incore detector moni-toring system may be used for monitoring the core power distribution by verifying that the incore detector local Power Density alams: In:= Are adjusted to satisfy the requirements of the core power a. e p distribution map which shall be updated at least once per 31 days of accumulated operation in MODE 1. i E b. Have their alarm setpoint adjusted to less than or ecual to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms: 1. Fit.x peaking augmentation f actors as shown in Figure r 4.2-1, _4 2. A measurement-calculational uncertainty factor of 1.058), 1.070 3. An engineering uncertainty f actor of 1.03, 4 A linear beat rate uncertainty factor of 1.01 due to axial fuel densification and themal expansion, and i t 5. A THERMAL POWER measurement uncertainty factor of 1.02. I t t h h ()n M 1 5"Y l* l0 N A [ V== l CALVERT CLIFF 5 - UNii 1 3/4 2-2 AmendmentNo.2J,2/,22,/g t [- q,- ww _ _ _

Page 81 [ I j._ ._ l. _ l 4.- .. ! _. '.i_. l 1 3,g . _ j _.. _ _. _l. _ _4__ .j. q. .t. ..l,_. .o. 5~- I l ~ ~ ._,i_.. _-)._. t_, ...f. _. { .l. .j.. .l.. .. j _.. _.,I __ L_ .l. 3,o 5 iiI .._1. ~ I i i _)._ ..t_ i '._..J. i l I i l l -l i _2 - } 'l I i fi- _ f. ~ U-UNACCEPTADLE I l l l i l l i I r-_ c i I 0.9 w Uf. A CCEr1 ABL E g - -j.. OPE A ATION OPE R ATION . j _.. _ ]. a. REGION l ._ l... I REGION l a

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Page 82 1.050 o (134.7, 1.049) (118.6, 1.045) a (106.5, 1. 042 ) (90.5, 1.037) (74.4, 1.031) 5 h (62.3, 1.027) u_ 5 C h o (46.2, 1.021) E 8< o (30.2, 1.015) (18.1,1.010) (2.0, 1.001) I t f i 1 1 0 20 40 60 80 100 120 140 DISTANCE FROM BOTTOM 0F CORE, INCHES B ALTIM ORE GAS & ELECTPJC CO. AUGUENTATION FACTORS vs DISTANCE " 9" coivert cuffs FROM BOTTOM 0F CORE 4.2-1 Nuclear Power Plant

Page 83 j l O[ rJ4ER DISTRIBUT10'; LIMITS T TOTAL PLA::AP RADI AL PEAKlfi5 F ACTOR - F, 7;;, e. LIMITlf;S CD:;DIT10!: FOR OPERATIO!: t 1 T

• 1. (1+T ), shall be

[ =F 3.2.2 The calculated value of F*#, defined as F 9 limited to _< l.660. i 1 APPLICAElLITY: M3DE 1*. r,,- ACT IO::: With F > 1.660, within 6 hours either: F. T xy I Reducg THERMAL POWER to bring the ctr,bination of THERMAL POWER and F to within the limits of Ficure 3.2-3 and withdraw the a. full Mngth CEAs to or beyond the [.ong Tena Steady State fs Insertion Limits of Specification 3.1.3.6; or 6 b. Be in at least HOT STA!;DBY. c== i + 8 SUR'.'E l_'.L_A!;CE RE021REME!:T 5 ? 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. T T T xy(1+T)AandF 4.2.2.2 F shall be calculated by the expression F =F 4 xy ry xy shall be detenained to be within its limit at the following intervals: sat i Prior to operation above 70 percent of RATED THERKAL POWER T a. i after each fuel loeding, \\ b. At least once per 31 days tf accumulated operation in MODE 1, and Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.030. c. q

• See Special lest Exception 3.10.2.

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Page 85 POWER DISTRIC'JTIO LIMITS TOTAL IUTEGRATED RADIAL PEAKING FACTOR - FT r LIMITING CON 31T10N FOR OPERATION T 3.2.3 The calculated value of F, defined as F = F (1+T ), shall be ( limited to < f 63 @ r r r q 1.571 l g APPLICAElllTY: MODE 1*. 1 ACTION: ,b T > h, within 6 hours either: With F 'd 1,571 r 2 a. Be in at least HOT STANDBY, or Q, b. 1 Reduce THERMA'. P0ilEP to bring the combination of THERHAL POWER and T' to witnin the limits of Fiqure 3.2-3 and withdraw the full Sength CEAs to or beyond the Long Term Steady State ~ i Insertion Linits of Specification 3.l.3.6. The THERMAL POWER limit determined from Ficure 3.2-3 shall then be used to estab-i lish a revised upoer THEkMAL POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction of RATED THERMAL POWER deterrined by Figure 3.2-3) and subseauent opera-tion shall be maintained within the reduced acceptable opera-tion region of Figure 3.2-4 W SURVE1LL ANCE RE0'J1REMENTE 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 i shall be calculated by the expression F =F 1+T ) band F T T shall be determined to be within 'its limit at the fbilow5n(g ifitervals[ Prior to operation above 70 percent of RATED THERMAL POWER a. after each fuel loading, b. At least once per 31 days of accumulated operation in MODE 1, and [ c. Within four hours if the AZIMUTH;L POWER TILT (T ) is > 0.030.j 6en in non-LCAb fba oPGAnoQ and T by the exnression Fr = 1.02 Fr (1 + Tq) ff 'See Special itst Exception 3.10.2. when in LDAh Folto.J oPEMno.J. ~- T ~. ' ['.; i-CALVERT CLIFF 5 - UNIT I 3/4 2-9 Amendmenthlo.2),2/,22 9y o_

Page 86 i I i: p_ 1.2 .. !....... k.... j, }.. ....t.' j . j j.- - l l ! l : i = r _ - -.'.,.. 1... l.i. ~ ;l. 3 ..j.. i l l I_i_. i .j i !i 3.3 j.. - __ __ ;. -j I t ,j t -: UN ACCEPT ABLE ^^ ~ ~ "~ onE n A1ios '*{4_'_ '_ j- ':UN ACCEPT ABLE.

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Page 87 l PO'..'En D15TRI BUT10': L 1'1I TS i D FUEL RESIDE!;CE TIME { LIMITING C0"D1TIO!: FOR OPERATION / 3.2.5 The additional core aver?ne fuel burnon necur. slated duririg fuel j cycle 3 shall be limited to 1 310 Ef fective Full Pmier Days. [ELICABILITY: MODE 1.~~ ACTION: With the additional core average fuel burnu accumulated during fuel cycle 3 determined to exceed 310 Effective Full Po.ler Days, be in at least HOT STANDBY viithin the next 6 hour. 5 / l e SURVE1LLANCE REQUIREMENTS 4.2.5 The core average vel burnup, based on gross thermal energy generation, shall be de ermined by calculation at least once per 31 days. D CALVERT CLIFFS - Utili 1 3/t. 2-13 Amendment lio. 27, 22'

..M 0 e e Page 88 b Ci-( i Adtevv<lv bl[e c D f 9 t 5 / s

e t }' } TABLE 3.2-1 i n DilB PAPAf1ETERS r< l I LililT S n U Four Reactor Three Reactor Two Reactor Two Reactor Coolant Pumps Coolant Pumps Coolant Pumps Coolant Pumps E Parameter Operating O_p_e ra t i ng O_perating-Same Loop Operating-Opposite Loop, 8 rY/r M

1. dF

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54\\^f W R f ' "^r ~ Col' leg Temperature 1 f"d Presst.cizer Pressure 1 2225 psia

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.u r l t r Reactor Coolant System Tntal flow Rate 1 370.000 gpm M :- r:g,. 2. AXI AL SHAPE illDEX Figure 3.2-4 F:j

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• Limit not applicable during either a THERMAL POWER ramp increase in excess of.% of RAT

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.e ,8 i Page 90 3/4.2 PD',lER DISTP.1BUT101 LIMITS /b x BASES 1 1 3/4.2.1 Lif;E AR HEAT RATE The limitation on linear heat rate ensures that in the even LOCA, the peak temperature of the fuel cladding will not exceed 2200 Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distributio l are capable of verifying that the linear heat rate doe q continuously monitoring the AXI AL SHAPE INDEX with the OPERABL limits. netric excore neutron flux detectors and verifying that the AXI AL SHAPE INDEX is maintained within the allowable limits of Figure 3.2 sy. In conjunction with the use of the excore monitoring system and in lishing the AX1 AL SHAPE INDEX limits, the folicwing assumptio 1

1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the flux peakino auamentation factors are as shown Figure 4.2-1, 3) the AZIMUTHAL POSER TILT not exceed the limits of Specification 3.2.2.

i The Incore Detector Monitoring System continuously provides a f~% direct measure of the peaking f actors and the alarms unich have bee i established for the individual incore detector secme j f peak linear heat rates will be maintained within the a j Figure 3.2-1.the conservative directions, for 1) flux peaking augmenta l shown in Ficure 4.2-1, 2) a measurement-calculational uncertainty factor of /11,QN) an engineering uncertainty f a r ~ d 1 l

5) a THERMAL POWER measurement uncertainty factor of 1.02.

i 4 3/4.2.2, 3/4.2.3 and 3/4.2.4 TOTAL PLANAR Att0 INTEGRATE FACTORS - F AND F AHD AZIMUTHAL PO'JER T-T T~ q y y are provided to ensure that the assump-I l The linitations on F and T tions used in the analysi57 or es9ablishing the Linear Heat Rate a f Power Density - High LCOs and LSSS setpoints remain valid during cpera-The li'aita-tion at th9 various allowable CEA group insertion lirits.are p tions on F and T r q l .pi-fln UncW]ao%}ylacb'r lJ0glic in L Drth fr u e u o g o Q l AmendmentNo.Al, p B 3/4 2-1 CALVERT CLIFFS - UMIT 1

I Pa3e 91 t i T R DISTRIBUTION LIMITS BASES ?* w the analysis establishing the DNB Margin LCO, and Themal Margin / Low Pressure LSSS setpoints remain valid durina. operation at the various allowable CEA group insertion limits. If I' F ' or T exceed their basiclimitations,operationmaycentinueufthe,r[ head 3itionalrestric-II[ tions imposed by the ACTION statements since these additional restric-i tions provide adequate provisions to assure that the assumptions used in establisMng the Linear Heat Rate, lhemal Margin / Lor Pressure and local Power Density - High LCOs and L555 setpoints remain valid. An AZlMUTHAL F0WER TILT > 0.10 is.not expected and if it should occur, sub-sequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt. l T The value of T that must be used in the equation F =FU (1 + T ) T r (1+T ) is the measured tilt. and F =F 9 r q T T age I'l The surveillance requirements for verifying that F ,F and.T t-within their limits provide assurance that the actual v ues"9fF'3F T I and T do not exceed the assumed values. Verifyinc F and F afGr each iuel loading prior to exceeding 75' of RATED THE F.AL POWER provides l additional assurance that the core was properly loaded. j h.,.,= 3/4.2.4 FUEL RESIDENCE TIME l nfulburupdu fel c"tle ir;s/ D / The imi tio ng t e th' d u ures that fu c ddi coll ew ~ l no' occu. P fo -ance ata f om simil e fu rot and a alysps of he i tall d f el rr s sho that 0 cla sing ellhewi . not ccur nt lim' tin batc until well 4 yond l th pr pose third ycle irdcplepel fop;nup atic. H wev ,o catic eyondAhe eci iec mit ion 'ill equi furthpranalyses. 3/4.2.5 DNB PARAMETERS - 1 I The limits on the DNB related parameters assure that each of the parameters are maintained within the nomal steady state envelope of j operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses cssumptions and have been enalytically i demonstrated adequate to maintain a minimum DNER of 1.19 throuchout each I analyzed transient. ~ li h i The 12 hour periodic surveillance of these parameters through instru-i ment readout is sufficient to ensure that the parameters are restored l within their limits following load changes and other expected transient i operation The IR month periodic measuremer+ of the RCS total flow rate l: E is adequate to detect fles degradation and ensure correlation of the flow indication channels with measured flow such that the indicated fl percent flow will provide sufficient verification of flow rate on a I 12 hour basis. p--- b ALVERT CLIFFS - UNIT 1 C B 3/4 2-2 Amendment No.~21, 22, g f F. =.. e ---e __ _ _ ae. ~ w-

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3 i Page 93 3 h/-- l i 1NT EN'~l D/d A LLx/ 1LAMic I i 3 I l 5 l

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Page 94 10.0 startup Testing The following discussions represent the major startup tests proposed for Calvert Cliffs 1, Cycle 4. Sufficient data is obtained to verify that the plant operates in a safe condition within the bounds of the applicable acceptance criteria and there':-a, the safety analysis. Hot Functional Testing CEDM Perf - ance Testing During this testing, the proper functioning of the CEAs, CEDMs, and CEA position indication will be verified through the insertion and withdrawal of the CEAs. Rod drop times will be measured and evaluated. Any irregu-larities shall be analyzed. RCS Flow Verification RCS flow rates will be verified based on differential pressure measurerents obtained across the RCPs and RV. These values will be compared to those obtained during previous testing for consistency. Initial Criticality Approach to criticality will commence with the withdrawal of the Shutdown CEA Groups, followed by the withdrawal, in sequence, of the Regulating CEA Groups resulting with Group 5 at mid-core. Criticality will be established through boron dilution. The plant will be allowed to stabilize following Critical Boron Concentration, and then proceed to the Low Power Physics Tests to verify physics design parameters. Low Power Physics Testina CEA Symmetry Check CEAs will be inserted into the core, and withdrawn from the core to confirm proper latching to their respective CEA extension shafts. A qualitative reactivity change will be apparent for single CEAs; and a quantitative reactivity, for dual CEAs. Critical Boron Concentration Critical Boron Concentrations will be determined for AR0, and Groups 1 iircugh 5 inserted.

Page 95 Isothermal Temperature Coefficient By varying the RCS temperature, the Isothermal Temperature Ccefficient will be determined. CEA Regulating Group 5 will be used to control and maintain flux and reactivity within a defined operating band. CEA Group Worth Measurements The RCS will be diluted / borated while the CEAs are inserted / withdrawn to compensate for a change in reactivity. These changes will be monitored via the reactivity computer. Non-overlapped worths will be determined. Power Ascension Test Two major plateaus for testing will be the 50% plateau and 100% plateau. The following specific tests will be performed as shown in order to compare and verify as-built characteristics of the core with their respective predictions. In addition to the 50% and 100% plateaus, the core power distributions will be determined and verified with predictions. 50% and 100% Plateau Testing Upon reaching the 50% power level, Xenon Equi' .m will be established with all roads out. Preparation for the Var:... T test will commence avg by diluting CEA Group 5' to approximately 105 inches. Following Xenon Equilibrium, T will be varied, thereby yielding data for the' isothermal c temperature coefficient determination. The Power Coefficient will be determined by maintaining T constant and varying the power level. avg In both cases, CEA 5-1 will be used for reactivity control and F.aintaining power.

Page 96 Acceptance Criteria Acceptance criteria for the above startup testing will be developed consistent with those presented during previcus startups. Acceptance Limits: CEA Groups Worth 1 15% on each group i 10% on sum of all groups r:easured. Critical Boron Measurements i 10% Temperature Coefficient i.3 x 10-4 A P/aF Power Coefficient i.2 x 10-4 4 g/% Rod Drops < 3.1 seconds Power Distribution F T, F T within Technical xy, and Tq r Specification Limits. If any acceptance criteria limits listed above are exceeded, an evaluation shall be made to determine first, the applicability of the prediction to the precise plant conditions under which the test was performed: second, the accuracy of the measurement; finally, the validity of the physics data input to the safety analysis for the entire cycle. Specifi-cally, if any regulating bank worth measurement falls outside of its acceptance criteria or if the total worth of the regulating banks falls outside of its acceptance criteria, shutdown bank C shall be measured and compared with its acceptance criteria. If shutdown Bank C worth fails outside of its acceptance criteria or if the accumulated total worth of all the banks measured f alls below their total worth acceptance creiterion (after appropriate corrections and adjustments) then an evaluation shall be made of the validity of the safety analyses for the entire cycle. A surrmary report of the results of these tests will be submitted to the NRC within 45 Days of completion of the startup program. ~

Page 97 11. REFERErlCES A. Chapters 1 throuah 6 1. Letter, J. W. Gore, Jr. to E. G. Case, " Third Cycle License Applicatien" dated December 1.1977, as modified by letter, A. E. Lundvall. -]r. to R. W. Reid, " Request for Amendment to Operating License", dated May 8, 1978 2. Letter, A. E. Lundvall, Jr. to R. W. Reid, " Report of Startup Testing for Third Cycle", Calvert Cliffs fluclear Power Plant Unit tio.1, Docket tio. 50-317, dated September 8,1978 3. " Baltimore Gas and Electric Company Calvert Cliffs fluclear Power Plant Units 1 and 2 Final Safety Analysis Report", dated January 4,1971, ar a en.ec. s 4. Letter, A. E. Lundvall, Jr. to B. C. Rusche, "Second Cycle License Application", dated October 1,1976 5. "CEPAll Method of Analyzing Creep Collapse of Oval Cladding", CEtiPD-187, dated June 1975 6. "Calvert Cliffs Unit 1 Reactor Operation with Modified CEA Guide Tubes", CE!l-83(B)-P, dated February 8,1978 and letter, A. E. Lundvall, Jr. to V. Stello, Jr., " Reactor Operation with Modified CEA Guide Tubes", dated February 17, 1978 7, "BG&E Calvert Cliffs I Slides Depicting SCOUT-I High Burnup Demonstration Program", m r iet,,, enteg j,f7p,. g Lurf.n:1. . %r. r eis, (,-; 8. Report of a Reconstitutable - B C Type CEA Design For Use in 4 the Ei&E Reactor CEri-105(B)-P, dated February 1,1979 cc:-u b ) :. "vn e-.-ele p,,m _ ir _,, a %. _,....,.,,..,. 7 / r '-[ _ [.,, tO 'T fr O*" 27f? *-i 2 le' tere 4 e +.~ } / ^ /-- O -1 letter date! 3/^a/~7, ,- uc,.-_, m _ s n san.mme.am na em?.. ~..w-

s., a - _ a r

Page 98 10. "C-E Fuel Evaluation Model Topical Report", CENPD-139, dated July 1, 1974 11. " INCA, Method of Analyzing In-Core Detector Data in Pcwer Reactors", CENPD-145-P, datea April 1975 12. Letter from M. R. Paradis to T. E. Short, " Omaha Public Power District Ft. Calhoun Station Unit No.1", dated August 8,1978 13. W. R. Cadwell, "PDQ-7 Reference Model", WAPD-TM-678, dated January 1968 14 " Fuel and Poison Rod Bowing", CENPD-225, dated October 1976 15. " Millstone Unit 2 Reactor Operation with Modified CEA Guide Tubes", CEN-80(N)-P, dated February 8, 1978 16 Letter, A. E. Lundvall to R. W. Reid, CEN-101 (B)-P "Calvert Cliffs II Cycle 2 Reload Submitta'l Update", dated August 28, 1978

Page 99

11. REFERENCES (COMT'b) 6, CFacter 7 1.

C-E Topical Report #CEMPD-199-P, "C-E Setpoint Methodology," April,1976. 2. A) Letter from G. W. Gore to E. G. Case, December 1,1977. B) Two letters from A. E. Lundvall to E. G. Case, March 20, 1978. C) Letter from A. E. Lundvall to E. G. Ccse, fiarch 17, 1978. D) Letter from A. E. Lundvall to R. W. Reid, March 16, 1978 E) Letter from A. E. Lundvall to R. W. Reid, March 29, 1978. 3. C-E Topical Report #CENPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," August, 1974. 4. C-E Topical Report #CEMPD-161-P, " TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core, uuly,1975. 5. WAPD-TM-479, J. A. Redfield, " CHIC-KIN - A Fortran Program for Intermediate and Fast Transients in a Water Moderator Reactor," January, 1965. 6. WAPD-TM-743, J. B. Yasinsky, M. Natelson, and L. A. Hageman, "TWIGL - A Program to Solve the Two-Dimensional, Two Group, Space-Time Neutron Diffusion Equations with Temperature Feedback," February, 1968. 7. CENPD-190A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July, 1976. 8. GEMP-482, H. C. Brassfield, et. al., "Recocrended 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 UD," April, 1968. 2 9. Idaho Nuclear Corporation, Monthly Report, Ny-123-69, October, 1969. 10. Idaho Nuclear Corporation, Monthly Report, Hai-127-70, March,1970. 11. Letter from A. E. Lundvall to D. Davis, July 13, 1977.

12. A) Letter from A. E. Lundvall to R. W. Reid, July 26, 1978, B) Letter from A. E. Lundvall to R. W. Reid, CEN-101(B)-P "Calvert Cliffs II Cycle 2 Reload Submittal Update", Dated Auguust 28, 1978

Page 100 l[, References (CCtst) C, Chapter 8 1. Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled fluclear Power Reactors, Federal Register, Vol. 39, flo. 3 - Friday, January 4, 1974.

2. CEt!PD-133, "CEFt.^.SH-%, A FORTRA!1 IV Digital Computer Prtgram for Reactor Blowdown Analysis", April 1974 (Proprietary).

CEtiPD-133, Supplement 2, "CEFLASH-4A, A FORTPAft IV Digital Computer Program for Reactor Blowdown Analysis (Modification)", December 1974 (Proprietary). ' 3. CEllPD-134, "CO iPERC-II, A Program for Emergency Refill-Reflood of the Core", April 1974 (Proprietary). CE!1PD-134, Supplement 1, "COliPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)", December 1974 (Proprietary). "4, CEllPD-135, "STRIKIti-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, April 1974 (Proprietary). CErlPD-135, Supplement 2, "STRIKIff-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (flodification)', February 1975 (Proprietary). CEtiPD-135, Supplement 4, "STRIKIf1-II, A Cylindrical Geometry Fuel Rod Heat Transfer Progrc.m", August 1976, (Proprietary). 5. CErlPD-135, Supplement 5, "STRIKIll-II, A Cylindrical Gecmetry Fuel Rod Heat Transfer Program", 6. CEllPD-139, "CE Fuel Evaluation tiodel", July 1974 (Proprietary). '7, Letter from A. E. Lundvall (BG&E) to B. Rusche (flRC) transnitting Cycle II ECCS Analysis, riovember 5,1976. o.

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