ML20076J846

Suppl 1 to 830822 Application for Amend to License DPR-53, Changing Tech Specs to Decrease Required Shutdown Margin & Increase Allowed Moderator Temp Coefficient.Supporting Documentation Encl

ML20076J846
Person / Time
Site:	Calvert Cliffs
Issue date:	09/01/1983
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	John Miller Office of Nuclear Reactor Regulation
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ML20076J850	List: ML20076J846 ML20076J860
References
NUDOCS 8309070346
Download: ML20076J846 (121)

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{{#Wiki_filter:U BALTIM ORE GAS AND ELECTRIC CHARi.ES CENTER P. O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL. JR. September 1,1983 v cc ms oc~r SUPPLY Director of Nuclear Reactor Regulation Attention: Mr. 3. R. Miller, Chief Operating Reactors Branch #3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Calvert Cliffs Nuclear Power Plant Unit 1 Dockets Nos. 50-317 Amendment to Operating License DPR 53 Supplement I to Seventh Cycle License Application

Reference:

(A) A. E. Lundvall, Jr., to R. A. Clark letter dated August 22,1983, " Unit i Seventh Cycle License Application" (B) A. E. Lundvall, Jr., to R. A. Clark letter dated June 30,1983, " Unit 1 Planned CEA Inspection Program" Gentlemen: Baltimore Gas and Electric Company hereby supplements an earlier request (Reference (A)) for amendment to Operating License DPR 53 for Calvert Cliffs Unit 1. The enclosed Supplement I presents a detailed description of the required Technical Specification changes with the supporting safety analysis information to ensure conservative operation at a rated thermal power of 2700 MWTh for Unit 1 Cycle 7. In Reference (B) we provided the proposed fuel assembly inspection plan to follow End of Cycle 6. Since then, the core loading has been revised. The new loading pattern now places a total of five modified guide tube assemblies under CEAs for a second cycle. The mechanical performance of the five assemblies is discussed in the supplement, Chapter 4.2. Technical Specification Changes and Justification The proposed changes to the Standard Technical Specifications (STS) required by this Amendment are described in Chapter 9 in the enclosure to this letter and justified in discussions in Section I thrcugh 8 of this enclosure. The majority of the Technical Specifications are duplicating the Technical Specifications as they currently exist for Unit 2. Unit 2 Cycle 5 is the reference cycle for Unit 1 Cycle 7. The changes that are unique to this cycle are STS 3/4.1.1.1, a decrease of the required shutdown margin, and STS 3 /4.3.1.1.4, an increase of the allowed moderator temperature coefficient. Both of these changes are justified in the report through the Dg\\ 8309070346 830901 PDR ADOCK 05000317 p PDR g\\ C I

Mr. 3. R. Miller Sept 2mber 1,1983 reanalysis of the Steam Line Break Design Bases Event. He remaining Technical ' Specification changes are a result of modifications to the Auxiliary Feedwater Actuation System. Le Feed Line Break Analysis justifies these modifications and is provided in the supplement. Determination of Siznificant Hazards Consideration We have determined, based on the analytical information supplied in the supplement, that this amendment request does not involve a significant hazards consideration. His conclusion was derived by applying the Commission's guidance for implementation of 10CFR50.92. The Commission provided this guidance concerning the application of these standards through certain examples in the Federal Register, Volume 48, Number 87, Wednesday, 4/6/83, Rules and Regulations. Example lii of actions involving no significant hazards considerations, on page 14870 of the Federal Register, is quoted below. "For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable." As described in Supplement I no fuel assemblies to be loaded into the Cycle 7 core will be of new or different design than those used previously and found to be acceptable to the NRC. No proposed changes to the Technical Specifications for Cycle 7 involve acceptance criteria that are significantly different from those previously found acceptable to the NRC. The analytical methods used to demonstrate conformance with the Technical Specifications and Regulations are consistent with previous NRC approvals and involve no significant changes. We conclude that the proposed reload license amendment does not involve a significant hazards consideration in that: 1. The probability or consequences of an accident previously evaluated is not significantly increased. The larger part of the reload is enveloped by the previously approved Unit 2 Cycle 5 license amendment. The Steam Line Break Accident was the only Design Bases Event reanalyzed. Its consequences are not significantly increased. 2. This reload application does not create the possibility of a new or different kind of accident from any previously evaluated. 3. %is license reload does not involve a significant reduction in the margins to safety. The small changes in the shutdown margin and increase in moderator temperature coefficients are not a significant reduction in margin of safety as the supplement analysis indicates.

m Mr. 3. R. Miller Sept mber 1,1983 "Ihe Plant Operation and Safety Review Comittee (POSRC) and the Offsite Safety Review Committee (OSSRC) have reviewed this proposed Amendment and these proposed changes to the Standard Technical Specifications and have concluded that implementation of this change will not result in an undue risk to the health and safety of the public. Yours very truly, BALTIMORE GAS A 'D ELECTRIC CO PANY 121 y Enclosure (20 copies) cc: 3. A. Biddison, Jr., Esq. G. F. Trowbridge, Esq. Mr. D. H. Jaffe, NRC Mr. R. E. Architzel, NRC Mr. R. R. Mills, CE Mr. R. E. Cochran, DHMH l l l

C21 vert Cliffs Unit - 1 Cycle 7 Refueling License Amendment Table of Contents Section Pm 1. Introduction.and Summary 1 2. Operating History of the Previous Cycle 2 3 General Description 3 4. Fuel System Design 11 5. Nuclear Design 13 6. Thermal-Hydraulic Design 25 7. Transient Analysis 28 8. ECCS Performance Analysis 73 9 Technical Specifications 76

10. Startup Testing 109

11. References 110 I

1 l \\

SUPPLEMENT 1 CALVERT CLIFFS UNIT 1 CYCLE 7 LICENSE APPLICATION l 1 l

m 1

1.0 INTRODUCTION

AND

SUMMARY

This report provides an evaluation.of design and performance for the operation of Calvert Cliffs Unit 1 during its seventh fuel cycle at full. rated power of 2700 MWt. All planned operating conditions remain the same as those for Cycle 6. The core will consist of 113 presently operating F,- G, and H assemblies, 64 fresh Batch J assemblies, and 40 B, D, E and F assemblies previously discharged from various cycles of both Calvert Cliffs units. Plant operating requirements have created a need for flexibility in the Cycle 6 termination point, ranging from 12,700 MWD /T to 13,700 MWD /T.. In performing analyses of design basis events, determining limiting safety settings and establishing limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 7 conditions would be enveloped, provided the Cycle 6 termination points fall within the above burnup range. The analysis presented herein will accommodate a Cycle 7 length of up to 14,000 MWD /T. The evaluations of the reload core characteristics have been conducted with respect to the Calvert Cliffs Unit 2 Cycle 5 safety analysis described in Reference 1, as supplemented in Reference 2, hereafter referred to as the "re ferenc e cycle" in this report unless otherwise noted. This is an appropriate reference cycle because of the similarity in the basic system characteristics of the two reload cores. Reference 1 was the basic Unit 2 Cycle 5 license submittal. Reference 2 was a supplemental report which provided recommended Technical Specification changes to accommodate the installation of the Safety Grade Auxiliary Feedwater Actuation System (SAFAS) and a transient analysis report covering the three transients that were affected by the SAFAS. 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 the revised analyses presented herein continue to show acceptable results. Where dictated by variations from the previous cycle (Unit 1 Cycle 6, References 3 and 4), proposed modifications to the plant Technical Specifications are provided and are justified by the analyses reported herein. These proposed modifications are similar to those proposed for the reference cycle (References 1 and 2). All Cycle 7 analyses address fuel exposure explicitly. Since the most limiting transient with respect to extended burnup is the steam line break (SLB) event, a generic analysis has been per formed to envelope future cycles and thereby reduce the need for licensing review. This generic analysis which entails a more negative MTC and a reduced shutdown margin requirement is presented herein. The operation of the SAFAS was explicitly accounted for in this SLB analysis and in the analysis of feedline break which is presented herein due to the implementation of the SAFAS for Unit 1 Cycle 7. The performance of Combustion Engineering 14x14 fuel at extended burnup is discussed in Reference 5. Fuel performance for Cycle 7 has been evaluated with the FATES 3 computer code (References 21 and 22) as approved by the NRC in Reference 23.

2-2.0 OPERATING HISTORY OF THE PREVIOUS CYCLE Calvert Cliffs Unit.1 'is presently operating in its sixth fuel' cycle utilizing Batch H, G, F, and D fuel assemblies (including eight D assemblies ' from Unit 2). Calvert Cliffs Unit 1 Cycle 6 began operation on July 5,1982 and reached full power on July 17, 1982. The Cycle 6 startup testing was reported to the NRC'in Reference 6., The reactor has operated _ up to the present time with the core reactivity, power distributions and peaking factors having followed the ' calculated predictions closely. - It is presently estimated that Cycle 6 will terminate on or about September 30, 1983 The Cycle 6 termination point can vary between 12,700 WD/T and 13,700 WD/T to accommodate the plant schedule and still be within the asstaptions of the Cycle _7 analyses. As of August 25, 1983, the Cycle 6 burnup had reached 12,245 WD/T. 1 l l

m 3 3 0 GENERAL-DESCRIPTION The Cycle 7 core will consist of the number and types of assemblies and fuel batches as described in Table 3-1. The primary change to the core in Cycle 7 is the removal of 52 Batch G/ assemblies, 43 Batch F assemblies, 1 Batch D. assembly and 8 Unit-2 Batch D assemblies. These assemblies will be replaced by 48 fresh unshimmed Batch J assemblies (4.05 wt5 U-235 enrichment), 16 fresh unshimmed Batch J' assemblies (3.40 wt5 U-235 enrichment), 4 Batch F assemblies (3.03 wt% U-235 enrichment) discharged from Unit 1 Cycle 5, 12' Batch E/ assemblies (2.73 wt% U-235 enrichment) discharged from. Unit 1 Cycle 4, 12 Batch D/ assemblies (2.73 wt5 U-235) discharged from Unit 2 Cycle 3 and 12 Batch B assemblies (2.45 wt% U-235) discharged from Unit 2 Cycle 1. Figure 3-1 shows the fuel management pattern to be employed in Cycle 7. Figure 3-2 shows the locations of the poison pins within the lattice of once-burned Batch B assemblies and the fuel rod locations in unshimmed assemblies; Figure 3-3 shows the poison pin locations within the lattice of the once-burned Batch H/ fuel. This fuel management pattern will accommodate Cycle 6 termination burnups from 12,700 WD/T to 13,700 WD/T. 'Ihe Cycle 7 core loading pattern is 90 rotationally symmetric. That is, if one quadrant of the core were rotated 90 into its neighboring quadrant, each assembly would be aligned with a simil ar assembly. This similarity includes batch type, number of fuel rods, initial enrichment and burnup. Figure 3-4 shows the beginning of Cycle 7 assembly burnup distribution for a Cycle 6 termination burnup of 13,700 WD/T. The initial enrichment of the fuel assemblies is also shown in Figure 3-4. Figure 3-5 shows the end of Cycle 7 assembly burnup distribution. The end of Cycle 7 core average exposure is approximately 27,400 MWD /T and the average discharge exposure is approximately 36,600 WD/T. Inconel Irradiation Experiment The inconel irradiation experiment, described in Reference 7, began in Cycle 5 of Calvert Cliffs Unit 1. At that time three irradiation specimens consisting of inconel tubes were inserted. They did not contain fuel or poison. Before Cycle 6 operation, one of these specimens was discharged and the other two were reinserted for Cycle 6 exposure. These two elements will be reinserted in Cycle 7. Scout Demonstration Assembly The Scout demonstration assembly was described in Reference 8. It is a Batch F t4st assembly which was originally inserted in Cycle 4.

Changes, similar to those described in Reference 3, were made to this assembly prior to its third cycle of irradiation in Cycle 6.

Further modifications are planned before returning the assembly to the core for its fourtn cycle of irradiation in Cycle 7. The present plans are to remove 2 segmented test rods from' the assembly and replace them with 2 stainless steel rods.

4 Prototype Demonstration Assemblies The Prototype demonstration assemblies were described in Reference 7. These are Batch G assemblies which were originally inserted in Cycle 5. These assemblies will be placed in symmetric locations in the core in Cycle 7 for e third cycle of irradiation. Prior to returning to the core, 2 segmented test rods will be removed from one of these assemblies and replaced with 2 stainless steel rods. U f l l l l l i i t l _.. _ ~.- . - ~,, _ - -

TABLE 3-1 CALVERT CLIFFS UNIT 1 CYCLE 7 CORE LOADING 1 Batch Total Average Initial Number Total Initial Burnup (MWD /T) Poison Poison of Poison Number Assembly Number of Enrichment [E006= Rods per Loading and Non-Fuel of Fuel Designation Assemblies wt% U-235 13,700] Assembly wt% B 0 Rods Rods 4 4 J 48 4.05 0 0 0 0 8448 J* 16 3.40 0 0 0 0 2816 H(l) 40 4.00 11,800 0 0 0 7040 H/(I) 32 3.55 16,400 8 '3.03 256 5376 Il) G 40 3.65 25,900 0 0 2-7038 F(2) 1 3.03 33,800 0 0 3 173 m F(3) 4 3.03 25,600 0 0 0 704 l E/(4) 12 2.73 23,700 0 0 0 2112 D/(5) 12 2.73 23,100 0 0 0 2112 B(6) 12 2.45 17,700 12 3.03 144 1968 TOTAL 217 .405 37,787 (1) Carried over from Unit 1 Cycle 6. (2) SCOUT assembly carried over from Unit 1 Cycle 6. J (3) Twice burned Batch F fuel discharged from Unit 1 Cycle 5. (4) Twice burned Batch E/ fuel discharged from Unit 1 Cycle 4. (5) Twice burned Batch D/ fuel discharged from Unit 2 Cycle 3. (6) Once burned Batch B fuel discharged from Unit 2 Cycle 1. 4

6 1 2 J J-3 4 5 6 7 J J-J F H 8 9 10 11 12 13 J H DI H/ G HI 1 l 14 15 16 17 18 19 20 J J' G H G J* El 21 22 23 24 25 26 27 28 J H G HI G HI G H 29 30 31 32 33 34 35 36 j J DI H G -H B H B 37 38 39 40 41 42 43 44 45 J H/ G HI B J' El H J 46 47 ++ 48 49 50 51 52 53 54 DI G J* G H El HI HI 55 56 57 58 59 60 61 62 + H HI U H B H HI F l + LOCATION OF DEMONSTRATION ASSEMBLY GCOUT) ++ LOCATION OF PROTOTYPE ASSEMBLIES BALTIMORE "8" GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 1 CYCLE 7 Nuc r Plant CORE MAP 3-1

7 UNSHIMMED ASSEMBLY B-12 POISON ROD AS.SEMBLY X X X X X X i X X X X X X l l FUEL R0D LOCATION O POIS0N R0D LOCATION l l I BALTIMORE GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 1 CYCLE 7 Figure

hon, ASSEMBLY FUEL AND OTHER R0D LOCATIONS 3-2 i

Nuc w

8 H/ -8 POIS0N R0D ASSEMBLY i X X X - X = 1 X X l l X X FUEL R0D LOCATION O POIS0N R0D LOCATION R "r* 9 CALVERT CLIFFS UNIT 1 CYCLE 7 GAS ELE T IC CO. 3-3 Nu I r "bont ASSEMBLY FUEL AND OTHER R0D LOCATIONS w

9 INITIAL ENRICHMENT W/0 U-235 BOC 7 BURNUP (MWD /T), E0C 6 - 13,700 MWDIT 1 J Z J 4.05 4.05 0 0 3 J 4 J 5 J 6 F 7 H 4.05 4.05 4.05 3.03 4.00 0 0 0 25,600 13,000 i 8 J 9 H 10 D/ 11 H/ 12 G 13 H) 4.05 4'.00 2.73 3.55 3.65 3.55 0 13,900 23,400 17,50C 27,400 17,700 14 J 15 J' 16 G 17 H 18 G 19 J' 20 El 4.05 3.40 3.65 4.00 3.65 3.40 2.73 0 0 26,200 10,000 25,30C 0 23,800 21 J 22 H 23 G 24 Hl. 25 G 26 H/ 27 G 28 H l 4.05 4.00 3.65 3.55 3.65 3.55 3.65 4.00 0 13,900 26,100 15,400 23,300 17,000 28,000 13,000 29 J 30 D/ 31 H 32 G 33 H 34 8 35 H 36 B 4.05 2.73 4.00 3.65 4.00 2.45 4.00 2.45 0 23,400 10,000 23,300 10,700 17,900 11,500 16,700 37 J 38 H/ 39 -G 40 H/ 41 B 42 J* 43 El 44 H 4.05 3.55 3.65 3.55 2.45 3.40 2.73 4.00 45 J 0 17,400 25,300 16,900 18,400 0 23,600 10,700 4 05 46 D/ 47 G 48 J* 49 G 50 H 51 El 52 H/ 53 H/ 2.73 3.65 3.40 3.65 4.00 2.73 3.55 3.55 54 J 22,600 26,400 0 27,700 11,500 23,700 14,700 14,800 4.05 0 55 H 56 HI 57 El 58 H 59 B 60 H 61 H/ 62 F l 4.00 '3.55 2.73 4.00 2.45 4.00 3.55 3.03 13,000 17,700 23,800 13,000 16,700 10,700 14,800 33,800 GAS E E T IC co* CALVERT CLIFFS UNIT 1 CYCLE 7 Figure coivert clirr, ASSEMBLY AVERAGE BURNUP AT B0C 3_4 Nuclear Power Plant AND INITIAL ENRICHMENT DISTRIBUTION

10 1 J 2 J 10,500 13,400 3 J 4 J 5 J 6 F 7 H 10,500 13,900 15,800 36,600 28,000 8 J 9 H 10 D/ 11 H/ 12 G 13 HI 12,200 28,000 34,800 31,800 39,700 31,600 14 J 15 J' 16 G 17 H 18 G 19 J' 20 E/ 12,200 16,700 39,400 26,400 38,700 17,500 35,700 21 J 22 H 23 G 24 HI 25 G 26 H/ 27 G 28 H 10,500 28,000 39,400 30,700 37,300 31,900 41,200 29,200 29 J 30 D/ 31 H 32 G 33 H 34 B 35 H 36 B 13,900 34,800 26,400 37,300 27,500 30,800 27,800 30,000 l 37 J 38 H/ 39 G 40 HI 41 B 42 J' 43 El 44 H '45 J 15,800 31,800 38,700 31,800 31,200 17,700 35,500 27,200 10,500 46 D/ 47 G 48 J* 49 G 50 H 51 El 52 HI 53 HI 54 j 33,600 38,900 17,500 40,900 27,800 35,700 29,900 30,000 13,400 55 H 56 H/ 57 El 58 H 59 B 60 H 61 HI 62 F 1 28,000 31,600 35,700 29,200 30,000 27,200 30,000 44,400 BALTIMORE GAS & ELECTRIC CO. CALVERT CLIFFS UNIT 1 CYCLE 7 Rsure ASSEMBLY AVERAGE BURNUP AT E0C (MWDlT) 3-5 Nuc r ow

Plan,

11 4.0 FUEL SYSTEM DESIGN 4.1 Mechanical Design The mechanical design for the standard Batch J reload fuel is identical to that of the standard Batch G fuel described in the-reference cycle submittal (Calvert Cliffs Unit 2 Cycle 5, Reference 1). 'It is also identical to that of the standard Batch H fuel used in Calvert Cliffs Unit 1 Cycle 6 (Reference 3). The mechanical designs of the Calvert Cliffs Unit 1 Batch E, F, G, and H fuel-assemblies 'were described in References 9, 10,

7. - and 3,

respectively. Details of the Calvert Cliffs Unit 2 Batch B and D fuel assemblies that will be used in Cycle 7 can be found in References 11 and 12, respectively. C-E has performed analytical predictions of cladding creep collapse time for all -Calvert Cliffs Units 1 and 2 fuel batches that will be irradiated in Cycle 7 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 7 duration (Table 4-1). These analyses utilized the CEPAN computer code (Reference 13) and the analysis procedures described in Reference 14. The analysis procedures described in Reference 14 were approved in Reference 15. TABLE 4-1 Minimum EOC 7 Batch Collapse Time Exposure B (Calvert Cliffs 2) 30,400 EFPH 22,960 EFPH D (Calvert Cliffs 2) >27,000 26,000 E >27,500 26,320 F 35,100; >39,200* 28,860; 39,2005 G 36,000 30,340 H >27,500 20,690 J >27,500 10,340 ' SCOUT Assembly I All batches of fuel (standard rods) were also reviewed for dimensional changes using the SIGREEP model described in Reference 16 (approved in Reference 15). Since the licensing of the reference cycle, some of the input correlations to SIGREEP have been refined to include additional extended burnup data. All clearances were found to be adequate during Cycle 7. The cladding collapse information in Table 4-1 is applicable to the test rods in the SCOUT and PROTOTYPE assemblies. The clearance for fuel rod growth for the test rods will be evaluated during the Cycle 6 outage. The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch J fuel are identical to those of the other t fuel batches to be incluced in Cycle 7.

Thus, the chemical or metallurgical performance of the Batch J fuel will remain unchanged from the performance of the Cycle 6 fuel.

I

12 4.2 Hardware Modifications to Mitigato Guida Tuba Wear All standard fuel assemblies which will be placed in CEA locations in Cycle 7 will have stainless steel sleeves installed in the guide tubes to prevent guide tube wear. A detailed discussion of the design of the sleeves and their effect on reactor operation is contained in Reference 17. Cycle 7 will also utilize five assemblies in CEA locations that were fabricated with modified guide tubes (see References 10 and 18) instead of sleeves to mitigate guide tube wear. Each of these modified assemblies has previously resided in. a CEA position for one cycle and has been examined for guide tube wear after that cycle.. The test results presented [ in References 19 and 20 showed no detectable wear. 4 3 Thermal Design The thermal performance of composite fuel pins that envelope the various fuel assemblies present in Cycle 7 (fuel Batches E, F, G, H and J and Batches B and D from Unit 2) have been evaluated using the FATES 3 version of the fuel evaluation model (References 21 and 22), i.e approved by the NRC (Reference 23). The analyses were performed with histories that modeled the power and burnup levels representative of the peak pins at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at end of Cycle 7. In addition, the SCOUT and PROTOTYPE test pins were analyzed and found not limiting with respect to thermal performance over their respective burnup ranges. The FATES 3 power-to-centerline melt limit was determined for Cycle 7 by taking some credit for the decrease in power peaking which is characteristic of highly burned fuel. Since a gradual decrease in the calculated power-to-melt (due to a decrease in the fuel melt temperature)- also accompanies burnup, the most limiting power-to-centerline melt has been found to occur within an intermediate burnup range. Using conservative estimates of the burnup point at which the power peaking begins to decrease and the rate at which it decreases for Cycle 7, the most limiting power-to-centerline melt has been determined to be in excess of 22 kw/ft and to occur at a rod average burnup of approximately 33,000 MWD /MTU. l

13 5.0 NUCLEAR DESIGN 5.1 Physics Characteristics. 5.1.1 Fuel Management The Cycle 7 fuel regement employs a mixed central region as described in Section 3, Figure a-1. The fresh Batch J fuel is comprised of 'two sets of assemblies, each having a unique enrichment in order to minimize radial power peaking. There are 48. assemblies with an enrichment of 4.05 wt1 U-235 and 16 assemblies with an enrichment of 3.40 wt1 U-235. With this loading, the Cycle 7 burnup capacity for full power operation is expected to be between 12,800 MWD /T and 13,400 MWD /T, depending on the final Cycle 6 termination point. The Cycle 7 core characteristics have been examined for. Cycle 6 terminations between 12,700 and 13,700 MWD /T and limiting values established for the safety analyses. The loading pattern (see Section 3) is applicable to any Cycle 6 termination point between the stated extremes. Physics characteristics including reactivity coefficients for Cycle 7 are listed in Table 5-1 along with the corresponding values from the reference cycle. Please note that the values of parameters actually employed in sa fety analyses are different from those displayed in Table 5-1 and are typically chosen to conservatively bound predicted values with acconmiodation for appropriate uncertainties and allowances. Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for the end of Cycle 7 zero power steam line break accident and a comparison to reference cycle data. The EOC zero power steam line break was selected since it is the most limiting zero power steam line break accident and, thus, provides the basis for. establishing the Technical Specification required shutdown margin. The required shutdown margin for Cycle 7 operation is substantially reduced through the generic steam line break methodology employed (negative reactivity credit due to the local heatup of fluid under the stuck CEA in the hot channel is accounted for). Table 5-3 shows the reactivity worths of various CEA groups calculated at full power conditions for Cycle 7 and the reference cycle. The CEA group identifications remain the same as in the reference cycle. The power dependent insertion limit (PDIL) curve is the same as for the reference cycle. 5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC7, MOC7 and

EOC7, respectively, that are characteristic of the high burnup end of the Cycle 6 shutdown window.

These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80 percent of the fuel height. The high burnup end of the Cycle 6 shutdown window tends to increase the power peaking in this axial central region of the core for Cycle 7. The planar radial power distributions for the above region with CEA Group 5 fully inserted at beginning and end of Cycle 7 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 6 shutdown window. l l I-

14 The radial power distributions described in this section are ' calculated ' data without uncertainties or other allowances. liowever, the single rod power. peaking values. do include the. increased peaking that -is. characteristic of fuel rods adjoining the water holes in the fuel assembly 1 lattice. For both DNB and kw/ft ' safety. and setpoint analyses ' in either rodded or. unrodded configurations, the power peaking values actually. used are higher. than.those expected to occur at any time during Cycle 7. .These conservative values, which are used in Section

7. of this document, establish - the allowable ' limits for power. peaking. to be observed during.

operation. i 1he range of. allowable axial peaking is defined by the_ limiting. conditions for operation covering axial shape index (ASI). Within these ASI ' limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum-three-dimensional or total peaking factor anticipated in Cycle 7 during normal base load, all rods out j operation at full ~ power is 1.92, not including uncertainty allowances and augmentation factors. 5.1.3 Safety Related Data 5.1.3 1 Ejected CEA Data The maximum reactivity worth and planar power peaks associated with an ejected CEA event are shown in T@le 5-4 for - Cycle 7 and the reference cycle. These values encompass the worst conditions anticipated during Cycle 7 for any expected Cycle 6 termination point. The values shown for Cycle 7 are the safety analyses values which are conservative with respect to the. actual calculated values. This section is included herein only to correct a textual error in Reference 1 (see note in Table 5-4). 5.1 3 2/5.1 3.3 Dropped CEA/ Augmentation Factors The Cycle 7 safety related data for these sections are identical to the safety related data used in the reference cycle. 5.2 Analytical Input to In-Core Measurements In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the manner described in Reference 24, which is the same method used for the reference cycle. 5.3 Nuclear Design Methodology Analyses have been performed in the same manner and with the same methodologies used for the reference cycle analyses with the following exceptions: a'. The HERMITE/ TORC methodology discussed in Reference 25 has been applied in the analysis of the steam line break accident, as concurred in by Reference 26. b. All cross sections used in the Cycle 7 design were generated in accordance with the methods described in Reference 27 and approved for i nuclear core design and sa fety-related neutronics calculations in Reference 28. ~.

15 5.4 Uncertainties in Measured Power Distributions The ' power distribution measurement uncertainties which are applied to Cycle 7 are the same as those applied to the reference cycle, as presented in Section 5 4 of Reference 1. l I i -- = ---*--r, +--,---.-e-, --,.-.,-,,_,-%w,, -,,-,.m..,,-,- ,--s ,,,--,.r.. -, -, - -,,we, rv-,*,--------r--,---y--

16 TABLE 5-1 1 CALVERT CLIFFS UNIT 1 CYCLE 7 NOMINAL PHYSICS CHARACTERISTICS Reference ** Units Cycle Cycle 7 Dissolved Boron Dissolved Boron content for Criticality, CEAs Withdrawn Hot Full Power, Equilibrium Xenon, BOC PPM 1032 1070-Boron Worth i Hot Full Power BOC PPM /5ap 105 108 Hot Full Power EOC PPN/5ap 85 84 Reactivity Coefficients (CEAs Withdrawn) Moderator Temperature Coefficients, Hot Full power, Equilibrium Xenon Beginning of Cycle 10-4ap/ F -0.1 -0.2 0 End of Cycle 10-Map / F -2.1 -2.2 0 Doppler Coefficient Hot Zero Power BOC 10-560 / F -1.48 -1.56 Hot Full Power BOC 10-5ap/ F -1.27 -1.28 Hot Full Power EOC 10-5ap/ P -1.47 -1.45 Total Delayed Neutron Fraction, Beff BOC 0.00609 0.00604 EOC 0.00521 0.00522 Neutron Generation Time,18 BOC 10-0 sec 24.0 23 4 EOC 10-6 sec 30.5 29.8

• • Unit 2 Cycle 5.

17 TABLE 5-2 CALVERT CLIFFS UNIT 1-CYCLE 7 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT ZERO POWER STEAM LINE BREAK, %Ap END-OF-CYCLE (EOC) Reference Cycle # Cycle 7 1. Worth of all CEA's Inserted 9.4 9.1

2. ~ Stuck CEA Allowance 1.8 2.6 3.

Worth of all CEA's Less Worth 7.6 6.5' of CEA Stuck Outen 4. Zero Power Dependent Insertion 1.8 1.6 Limit CEA Bite 5. Calculated Scram Worth 5.8 49 6. Physics Uncertainty plus 0.6 0.6 Bias 7. Net Available Screm Worth 5.2 4.3 8. Technical Specification 5.2 4.3*** Shutdown Margin 9. Margin in Excess of Technical 0.0 0.0 Specification Shutdown Margin

1. Unit 2 Cycle 5

• • Stuck CEA is one which yields worst results for HZP S LB,

i.e., worst combination of scram worth and reactivity insertion with cooldown.

1. 1. 1. The generic SLB analysis presented herein supports a shutdown margin of 3.75ap (see Table 7-2).

Presently, the shutdown margin is only being reduced to 4 3%Ap to be consistent with other existin6 safety analyses.

18 TABLE 5-3 CALVERT CLIFFS UNIT 1 CYCLE 7 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, Sap Beginning of Cycle End of Cycle Regulating Re ference* Re ference* CEA's Cycle Cycle 7 Cycle Cycle 7 Group 5 0.49 0.53 0.63 0.64 Group 4 0.27-0.34 0 36 0.44 Group 3 0 91 0.99 1.16 1.07 Note Values shown assume sequential group insertion. ' Unit 2 Cycle 5. l l i I l l l l.

19 TABLE 5-4 CALVERT CLIFFS UNIT - 1 CYCLE 7 CEA EJECTION DATA Limiting Values Reference Cycle Unit 1 Cycle 7 Safety Analysis Value Safety Analysis Value Maximum Radial Power Peak Full Power with Bank 5 inserted; worst CEA ejected 3.6 3.6 Zero power with Banks 5+4+3 inserted; worst CEA ejected 9.4 9.4 Maximum Ejected CEA Worth (%ao ) Full power with Bank 5 inserted; worst CEA ejected 0.28 0.28 Zero Power with Banks 5+4+3 inserted; worst CEA ejected 0.63 0.63 Notes 1. Uncertainties and allowances are included in the above data. 2. The Cycle 7 safety analysis values are conservative with respect to the actual Cycle 7 calculated values. 3 In Reference 1, the correct values used for safety analyses were listed in Table 7 3.1-1. It was erroneously stated in Section 5.1.3 of that reference that the values for the reference cycle were the same as those used in Reference 3 The values listed in the above table were the actual values used in the safety analyses of the reference cycle. This table and the corresponding text are included herein only to correct this textual error.

20 i 0.80 1.05 0.80 1.09 1.25 0.80 1.14 l O.% 1.10 0.81 1.06 0.87 1.00 0.96 1.35 0.98 1.23 0.94 1.28 0.80 1 0.80 1.11 0.98 1.09 0.97 1.02 0.89 1.14 i 1.09 0.81 1.24 0.97 1.19 0.82 1.11 0.84 1.26 1.06 0.95 1.03 0.81 1.22 0.73 1.11 0.80 0.79 0.88 1.30 0.90 1.12 0.75 0.% 0.96 1.05 1.14 1.00 0.80 1.14 0.84 1.11 0.% 0.61 NOTE: X = MAXIMUM 1-PIN PEAK = 1.59 ^ CALVERT CLIFFS UNIT 1 CYCLE 7 Figure GAS EET C CO-ASSEMBLY RELATIVE POWER DENSITY AT BOC, 5_1 coivert clirr, EQUILIBRIUM XENON Nuclear Power Plant

21 0.78 0.99 0.76 1.02 1.17 0.80 1.10 0.87 1.02 0.80 1.03 0.88 1.00 0.87 1.19 0.92 1.18 0.94 1.26 0.83 0.76 1.02 0.92 1.07 0.98 1.06 0.93 1.18 1.02 0.81 1.18 0.98 1.21 0.91 1.19 0.95 i 1.17 1.04 0.95 1.06 0.91 1.30 0.85 1.24 0.78 ~ 0.79 0.88 1.27 0.93 1.19 0.86 1.13 1.14 0.99 1.10 1.00 0.83 1.18 0.95 1.24 1.14 0.78 NOTE: X= MAXIMUM 1-PIN PEAK = 1.47 ^ GAS E E T IC co' CALVERT CLIFFS UNIT 1 CYCLE 7 Figure coivert clifts ASSEMBLY RELATIVE POWER DENSITY AT 7 GWDIT 5-2 Nuclear Power Plant EQUILIBRIUM XENON

22 l 0.82 1.02 0.80 1.04 1.18 0.84 1.12 0.89 1.03 0.84 1.04 0.90 1.02 0.89 1.17 0.93 1.16 0.94 1.23 0.85 i 0.80 1.03 0.93 1.05 0.% 1.04 0.92 1.14 1.04 0.84 1.16 0.96 1.17 0.92 1.15 0.95 1.18 1.04 0.94 1.04 0.92 1.24 0.85 1.18 i 0.83 0.84 0.90 1.23 0.93 1.15 0.86 1.08 1.08 l 1.02 1.12 1.02 0.85 1.14 0.95 1.18 1.08 0.78 NOTE: X MAXIMUM 1-PIN PEAK = 1.45 c ^ CALVERT CLIFFS UNIT 1 CYCLE 7 Figure GAS EET CCO* calvert cliff, ASSEMBLY RELATIVE POWER DENSITY AT E0C, 5-3 Nuclear Power Plant EQUILIBRIUM XENON

23 0.77 1.00 0.78 1.08 1.22 0.76 1.04 0.78 1.02 0.82 1.09 0.85 // A 0.79 0.92 1.27 1.03 1.33 0.80 0.78 1.02 0.92 1.12 1.06 1.13 0.99 1.23 1.08 0.83 1.27 1.06 1.27 0.92 1.22 0.94 1.24 1.10 1.03 1.14 0.91 1.31 0.79 1.14 i X 0.78 0.76 0.86 1.34 1.00 1.22 0.80 b.95 0.88 1.00 / 1.04 0.80 1.23 0.94 1.14 0.88 / '/ // / NOTE: X MAXIMUM 1-PIN PEAK 1.56 7 CEA BANK 5 / LOCATIONS ^ CALVERT CLIFFS UNIT 1 CYCLE 7 sgure GAS EET CCO* coivert ciirr, ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 5-4 Nuclear Power Plant INSERTED, HFP, BOC

24 0.84 1.03 0.80 1.06 1.20 0.83 1.04 0.72 0.% 0.86 1.07 0.86 // 0.72 0.85 1.18 1.00 1.24 0.83 0.80 0.96 0.85 1.05 1.02 1.13 1.00 1.20 1.06 0.86 1.18 1.02 1.25 1.00 1.24 1.02 1.20 1.07 1.00 1.13 1.00 1.34 0.91 1.22 X 0.84 0.83 0.86 1.24 1.00 1.24 0.92 1.07 1.00 1.04 0.83 1.20 1.02 1.22 1.00 g l NOTE: X MAXIMUM 1-PIN PEAK = 1.48 CEA B ANK 5 l f / LOCATIONS CALVERT CLIFFS UNIT 1 CYCLE 7 Figure ' GAS ELE T IC CO-ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 5-5 calvert cliffs l Nuclear Power Plant INSERTED, HFP, EOC

25 6.0 THERMAL HYDRAULIC DESIGN 6.1 DNBR Analysis-Steady state DNBR analyses of Cycle 7 at the. rated power level of 2700 MWt have been performed using the TORC ~ computer code described in Reference 1, the CE-1 critical heat flux correlation described in Reference 2, and the simplified modeling methods described i.1 Reference 3. 'A' variant of TORC called

CETOP, optimited for simplified modeling applications, was used in this cycle to develop the " design thermal margin mod el" described generically in Reference 3 Details of CETOP are discussed in Reference 4.

CETOP was approved for use on Calvert Cliffs . Units in Reference 5. CETOP is used only because it reduces computer costs significantly; no margin gain is expected or taken credit for. Table 6-1 contains a list of pertinent thermal-hydraulic design parameters applicable for both safety analyses and for generating reactor protective system setpoint information. The calculational factors (engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in' Table 6-1 have been combined statistically with other uncertainty factors ' at the 95/95 confidence / probability level (Reference 6) to define a design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6 and approved by the NRC in Reference 5. The applicability of this minimum DNBR limit was verified for Cycle 7. Investigations have been made to ascertain the effect of the CEA guide tube wear problem and the sleeving repair on DNBR margins. The findings were reported to the NRC in Reference 7 which concluded that the wear problem and the sleeving repair do not adversely affect DNBR margin. 6.2 Effects of Fuel Bowing on DNBR Margin The effects of fuel rod bowing on DNB margin for Calvert Cliffs Unit 1 i Cycle 7 have been evaluated using the methods described in Reference 8. These methods were approved by NRC in Reference 9. Based upon these methods, a penalty of 0 3% DNBR is required to account for the adverse T-H effects of rod bow at an assembly average burnup of 30 GWD/T. An equivalent penalty of 0.4% in radial peak was applied in the determination of the Tech. Spec. limit on radial peak. A conservative (i.e., maximum over all operating ranges) conversion factor of -1.2% radial peak / $ DNBR was used to determine the equivalent radial peak penalty. For those assemblies with an assembly average burnup in excess of 30 GWD/T, the minimum best estimate margin available relative to more limiting l peaking values present in other assemblies is greater than 15%, l substantially exceeding the corresponding rod bow penalties based upon I Reference 8. Hence, sufficient available margin exists to offset rod bow penalties for assemblies with burnup greater than 30 GWD/T. i i,,m ~,.. -. _,. ,.__m,_ _.,,y_.. _m, ____,..,,__m..._

26 TABLE 6-1 CALVERT CLIFFS UNIT 1 THERMAL HYDRAULIC PARAMETERS AT FULL POWER ** Reference + General Characteristics Unit Unit 2, Cycle 5 Cycle 7 Total Heat Output (core only) MWg 2700 2700 10 BTU /hr 9215 9215 Fraction of Heat Generated 975 .975 In Fuel Rod Primary System Pressure psia 2250 2250 (Nominal) Inlet Temperature F 548 548 Total Reactor Coolant Flow gpg 381,600 381,600 (steady state) 10 lb/hr 143.8 143.8 6 Coolant Flow Through Core 10 lb/hr 138.5 138.5 Hydraulic Diameter ft 0.044 0.044 (nominal channel) 6 2 Average Mass Velocity 10 lb/hr-ft 2.59 2.59 Pressure Drop Across Core psi 11.1 11.1 (steady state flow irreversible 6P over entire fuel assembly) Total Pressure Drop Across psi 34.7 34.7 Vessel (based on steady state flow and nominal dimensions) 2 Core Average Heat Flux BTU /hr-ft 185,532*** 183,000**** (Accounts for above fraction of heat generated in fuel rod and axial densification factor) 2 Total Heat Transfer Area ft 48,415' " 49,100**** (Accounts for axial densification factor) 2 Film Coefficient at Average BTU /hr-ft - F 5930 5930 Conditions

27 TABLE 6-1 (continued) Re ference+ , General Characteristics Unit Unit 2, Cycle 5 Cycle 7 Average Film Temperature 'F 31 31 Difference Average Linear Heat Rate of kw/ft 6.20*** 6.12'#88 Undensified Fuel Rod (accounts for above fraction of heat generated in fuel rod) Average Core Enthalpy Rise BTU /lb 66.5 66.5 C Maximum Clad Surface F 657 657 Temperature Reference Calculational Factors Unit 2, Cycle 5 Cycle 7 Enginering Heat Flux on Hot channel 1.03' 1.03' Engineering Factor on Hot Channel 1.02' 1.02' Heat > Input Rod Pitch and Clad Diameter Factor 1.0658 1.065' Fuel Densification Factor (axial) 1.01 1.01 Notes 'These factors have been combined statistically with other uncertainty factors at 95/95 confidence / probability level (Reference 6) to define a design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 6 and approved by the NRC in Reference 5. This limit was verified to be applicable to Cycle 7. "Due to the statistical combination of uncertainties described in References 6, 10, and 11, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.

• • • Based on a value of 928 shims.
      • Based on a Value of 405 shims and non-fuel rods.

+ Reference cycle (Unit 2, Cycle 5) analysis is contained in Reference 12.

28 7.0 TRANSIENT ANALYSIS This _ section presents the results of the Baltimore Gas & Electric Calvert Cliffs Unit 1, Cycle 7 Non-LOCA safety analysis at 2700 MWt. The Design Bases Events (DBEs) considered in the safety analysis are listed in Table 7-1. These events were categorized in the following groups: 1. Anticipated Operational Occurrences (A00s) for which the intervention of the Reactor Protection System (RPS) is necessary to prevent exceeding acceptable limits. 2. A00s for which the intervention of the RPS trips and/or initial steady state thermal margin, maintained by Limiting Conditions for Operation (LCO), are necessary to prevent exceeding acceptable limits. 3 Postulated Accidents The Design Basis Events (DBEs) considered in the Unit 1 Cycle 7 safety analyses are listed in Table 7-1. Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to melt fuel design limits are presented in Table 7-2. As indicated in Table 7-1, no reanlaysis was performed for the DBEs for which key transient input parameters are within the bounds (conservative with respect to) of the reference cycle values (Unit 2, Cycle 5, Reference 1). For these DBEs the results and conclusions quoted in the reference cycle analysis are valid for Unit 1 Cycle 7. For the event reanalyzed, Table 7-3 shows the reason for the reanalysis, the acceptance criterion to be used in judging the results and a summary of the results obtained. Detailed presentations of the results of the reanalysis are provided in Section 7 3 2. An appendix is included to formally present an analysis of the Feed Line Break event. This event is included to support the installation of a safety grade auxiliary feedwater actuation system for Calvert Cliffs Unit 1.

29 TABLE 7-1 CALVERT CLIFFS UNIT 1, CYCLE 7 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA-SAFETY ANALYSIS Analysis Status 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits: 7.1.1 Boron Dilution Not Reanalyzed 7.1.2 Starjup of an Inactive Reactor Coolant Not Reanalyzed Pump 7.1.3 Loss of Load Not Reanalyzed-7.1.4 Excess Load Not Reanalyzed 7.1.5 Loss of Feedwater Flow Not Reanalyzed 7.1.6 Excess Heat Removal due to Feedwater Not Reanalyzed Malfunction 7.1.7 Reactor Coolant System Depressurization Not Reanalyzed 7.1.8 Excessive Charging Event Not Reanalyzed 7.2 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial steady state thermal margin, maintained by the LCOs, are necessary to prevent exceeding the acceptable limits: 2 7.2.1 Sequential CEA Group Withdrawal Not Reanalyzed 7.2.2 Loss of Coolant Flow Not Reanalyzed 7.2 3 Full Length CEA Drop Not Reanalyzed 7.2.4 Transients Resulting from the Not Reanalyzed MalfunctionofOgeSteamGenerator" Loss of AC Power Not Reanalyzed 7.2.5 7.3 Postulated Accidents 7.3.1 CEA Ejection Not Reanalyzed 7.3.2 Steam Line Rupture Reanalyzed 733 SteamGeneragorTubeRupture Not Reanalyzed 7.3.4 Seized Rotor Not Reanalyzed I Technical Specifications preclude this event during operation. 2Requires High Power and Variable High Power Trip. 3Requires Low Flow Trip. Requires trip on high differential steam generator pressure.

30 .. TABLE 7-2 CALVERT CLIFFS UNIT 1. CYCLE 7 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle Values (Unit 2 Unit 1, Physics Paraneters Unita Cycle 5) Cycle 7 Values Radial Peaking Factors For DNB Margin Analyses T (Fr} Unrodded Region 1.70+'# 1.70+ Bank 5 Inserted 1.87+ 1.87+ ForplanarRadialComponent (F ) of 3-D Peak (CN Limit Analyses) Unrodded Region 1.708 1.70# Bank 5 Inserted 1.878 1.87* Maximum Augmentation -1.055 1.055 Factor' 0 Moderator Temperature 10 A0 / F -2.5++.5 -2.5++.5 Coe fficient Shutdown Margin (Value top -5.2 -3 7+ Assumed in Limiting EOC Zero Power SLB) Tilt Allowance 5 3.0 30

1. For DNBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically.

The procedures used in the Statistical 1 Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c. Ihese procedures have been approved by NRC for the Calvert Cliffs Units in Reference 3

• • The eff initial MTC assumed for the Unit II Cycle 5 SLB event was

-2.2x10 getive 40/ F. +The values assumed are conservative with respect to the Technical Specification limits.

31 TABLE 7-2 (continued) Reference Cycle Values (Unit 2, Unit 1, Safety Parameters Units Cycle 5) Cycle 7 Values Power Level MWt 27008 2700* O Maximum Steady State F 548' 548' Temperature Minimum Steady State psia 2200* 2200* RCS Pressure 6 Reactor Coolant Flow 10 1bm/hr 138.5' 138.5* Negative Axial Shape I .15' .15' p LCO Extreme Assumed at Full Power (Ex-Cores) Maximum CEA Insertion % Insertion 25 25 at Full Power of Bank 5 Ma~imum Initial Linear KW/ft 16.0 16.0 x Heat Rate for Transient Other than LOCA Steady State Linear KW/ft 22.0 22.0 Heat Rate for Fuel CIM Assumed in the Safety Analysis CEA Drop Time from sec 31 31 Removal of Power to Holding Coils to 905 Insertion Minimum DNBR (CE-1) 1.23' 1.23' 'For DNBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically. The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c. These procedures have been approved by NRC for the Calvert Cliffs Units in Reference 3

32 TABLE 7-3 DESIGN BASIS E%ENT REANALYZED FOR UNIT 1, CYCLE 7 Reason for Acceptance Summary Event Reanalysis Criterion of Results (changes relative to reference cycle) Steam Line Changes in moderator Site boundary ' Site boundary doses Break cooldown curve and dose within are a small fraction available scram 10CFR100 limits. of 10CFR100 limits; worth at trip due post trip MDNBR does to Tech. Spec. not violate 1 30 MTC limit change, (MacBeth) SAFDL for resulting from the limiting case. extended burnup. Further details in Section 7 3 2. t

33 7 3 2 S_ team Line Rupture Analysis The steam line rupture (SLB) event has been reanalyzed for Calvert Cliffs Unit I and Unit II. The purpose of the analysis is to provide results based on conservatively enveloping initial conditions and assumptions such that the results will be applicable to a large number of future plant operating cycles. This analysis incorporates the effects of the safety grade auxiliary feedwater actuation system (AFAS). A spectrum of steam line break sizes, both inside and outside containment initiated from hot full power (HFP) and hot zero power (HZP) were analyzed. In addition, the analysis was performed with and without Loss of AC (LOAC) power on turbine trip. For outside containment breaks which credit a high oower trip, a study parametric on moderator temperature coefficient (MTC) was also performed. The acceptance criteria for this postulated accident is that site boundai y doses will be within the 10CFR100 guidelines. The results of the most limiting steam line break event, inside and outside containment are presented herein. Analysis Assumptions and Initial Conditions SLB Inside Containment The SLB event was initiated from the conditions listed in Table 7 3.2-1. The moderator temperature coefficient (MTC) of reactivity assumed in the analysis corresponds to end of life, since this MTC results in the greatest positive reactivity change during the RCS cooldown caused by the steam line rupture. Since the reactivity change associated with mod.erator feedback varies significantly over the moderator density covered in 'the analysis, a curve of reactivity insertion versus density rather than a single value of MTC, is assumed in the analysis. The moderator cooldown curve assumed in the analysis is given in Figure 7.3.2-1. This moderator cooldown curve was conservatively calculated assuming that on reactor trip the control element assembly which yields the most severe combination of scram worth and reactivity insertion is stuck in the fully withdrawn position. The reactiviy' defect associated with the fuel temperature decreases was also based on mi end of life Doppler defect. The Doppler defect based on an end of life fuel temperature coefficient (FTC), in conjunction with the decreasing fuel temperatures, causes the greatest positive reactivity insertion during the steam line rupture event. The Dopper multiplier (uncertainty) on the FTC assumed in the analysis is given in Table 7.3.2-1. The 8 fraction assumed was the maximum absolute value including uncertainties for end of life conditions. This too is conservative since it maximizes suberitical multiplication and thus, enhances the potential for Return-To-Power (R-T-P). The analysis also assumed a conservatively low value of boron reactivity worth of -1.0% Ao per 85 PPM for safety injection flow from the High and Low Pressure Safety Injection pumps. The mini.aum CEA worth assumed to be available for shutdown at the time of reactor trip at the maximum allowed power level is 5.56%Ap. This available scram worth was calculated for the stuck rod which produced the moderator cooldown curve in Figure 7.3.2-1.

34 During a ~ return-to-power, - negative reactivity credit was assumed in the analysis. This negative reactivity credit is due to the local heatup of the inlet fluid in the hot channel, which occurs near the location of the stuck ' CEA. This ' eredit is based on three-dimensional coupled neutronic-thermal-hydraulic calculations performed with the HERMITE/ TORC code (References 4 and 5). The actual credit differs from that used in the Calvert Cliffs Unit II Cycle 5 steam line break event (Reference 1) in that the credit in this analysis is Calvert Cliffs specific. The Calvert Cliffs Unit II Cycle 5 analyses used a small fraction of the negative reactivity credit justified for St. Lucie Unit 2 (Flordia Power and Light). The analysis only credited the low steam generator pressure trip. An analysis trip setpoint of 600.0 psia was assumed in the analysis. This represents the Technical Specification setpoint of 685.0 psia and an uncertainty of 85.0 psia. The analysis also assumed that a Steam Generator Isolation Signal (SGIS) is generated when secondary pressure reaches 600.0 psia. This represents the Technical Specification setpoint of 685.0 psia and an uncertainty of 85.0 psia. A Main Steam Isolation Valve (MSIV) closure time of 6.9 seconds (includes valve closure time and signal processi1g delay time) was conservatively assumed in the analysis. The analysis conservatively assumed that following reactor trip, the main feedwater flow is ramped down to 8% of full power feedwater flow in 20 seconds and that the main feedwater isolation valves are completely closed in 80 seconds a fter a low steam generator pressure or a main steam isolation signal. These assumptions are consistent with Technical Specification limits. The analysis assumptions regarding the auxiliary feedwater actuation setpoint, the associated time delays, and the AFW flow through each leg are given below. Tney were conservatively chosen to initiate AFW flow sooner and deliver the maximum AFW flow to the ruptured steam generator, which maximizes the primary cooldown and enhances the potential R-T-P. The auxiliary feedwater Technical Specification actuation setpoint is 45% of steam generator level wide range ir.Jication with an uncertainty of +185. Auxiliary feedwater (AFW) was conservatively assumed to initiate at time of reactor trip, which in all cases resulted in AFW initiation at a level far above the Technical Specification actuation setpoint plus uncertainties. This was done to ensure the analysis results would remain bounding in the event of any future revision of the Technical Specification or uncertainty. Time delays associated with the AFW pumps were conserva-tively set to zero resulting in instantaneous flow, even in loss of AC cases. This is consistent with the enveloping nature of the analysis. All flow from the AFW pumps is conservatively directed to the damaged steam generator until automatic isolation of that steam generator. AFW pump flow is assumed to be at a runout value of 1300 gpm. i The analysis also included isolation of the ruptured steam generator when the steam generator differential pressure reached the analysis setpoint of 365.0 psid. This represents a Technical Specification setpoint of 135.0 psid and an uncertainty of 230.0 psid. In addition, a 20.0 second time l --nw., r- -, -, n-, e,, -,.-,-.m- .-----,,.--,e ,----n an g

L -35 i -d";1cy wn tsumed in tha cntlysib to close th2 AFW icolatien (i.e., block) valves. These assumptions are conservative since it delays the isolation of AFW to the ruptured steam generator. A safety injection actuation analysis setpoint of 1578.0* psia was assumed in. the analysis. The analysis conservatively assumed that on -a Safety Injection. katuation Signal (SIAS), only one High Pressure Safety Injection (HPSI) pumt starts. In addition, a maximum time delay of 30 seconds for HPSI pumps to accelerate to full speed was assumed in the analysis. In case of LOAC power, additional time delays were included in the analysis. It included 10.0 seconds for the diesel generators to start and reach speed following the LOAC and 5.0 seconds for the HPSI pump to be loaded on line i regardless of which sequencer (i.e., shutdown or LOCA) was initiated. The post-trip minimum DNBRs were calculated using the MacBeth correlation (Reference 6) with the Lee non-uniform mixing correlation factor (Reference 7). Analysis Assumptions and Initial Conditions SLB Outside Containment The results of the inside containment break showed that no post-trip Returg-To-Power will occur for break sizes less than approximately 3.0 ft ; consequently, no post-trip return to power will occur for outside containment breaks due to the 2.35 ft in-line fl w restrictor. For 2 break sizes greater than 3.0 ft the inside containment analysis presented above is limiting with respect to R-T-P. Therefore, the assumptions for the outside containment break listed below are designed to maximize the power excursion prior to reactor trip, rather than maximizing the post trip return to power. The limiting SLB event outside containment was initiated from conditions listed in Table 7.3.2-2. The assumptions for the outside containment break which differ from the inside containment event are listed below. The reactivity defect associated with the fuel temperature change was based on a beginning of life Doppler defect. This Doppler defect based on a beginning of life fuel temperature coefficient -(FTC), in conjunction with increasing fuel temperatures, causes the minimum negative reactivity insertion and the maximum power excursion prior to reactor trip. The Doppler Multiplier (uncertainty) on the FTC assumed in the analysis is given in Table 7.3 2-2; a conservatively low value is employed to minimize Doppler feedback during the pre-trip power excursion. The 8 fraction assumed is the end of cycle minimum absolute value including uncertainties. This too is conservative since it maximizes the power excursion prior to trip by minimizing the contribution of delayed neutrons to the rate-of-change of power. A spectrum of MTCs was employed to determine the effect of MTC on power range detector response during an outsid e containment steam line break. Only the High Power Trip and/or the Low Steam Generator Pressure Trip were credited. ' Conservative compared to current Technical Specification and existing uncertainties. i

36 The assumptions made to maximize the site boundary dose are given in Table 7 3 2-3. During the event, two sources _ of radioactivity contribute to the site boundary dose: (1) the initial activity in. the steam generator and (2) the activity associated with pr imary to secondary leakage. The primary activity includes the maximum initial activity allowed by the Technical Specificat!ons and any activity released to the coolant due to fuel failure. 'he analysis conservatively assumed that all fuel pins with minimum DNBR talow the design limit of 1.23 (CE-1 correlation) failed. The minimum DNBR during the transient was calculated using the thermal-hydraulic code CETOP (Reference 8). In calculating the site boundary dose, the analysis conservatively assumed that all activity is released to the atmosphere with a decontamination factor of 1.0. Results SLB Inside Containment The SLB event with Loss of AC (LOAC) power on turbine trip results in the maximum post trip return-to-power (R-T-P) and, thus, the minimum post trip transient DNBR. This occurs because LOAC power causes the Reactor Coolant Pumps (RCPs) to coast down. The decreasing coolant flow is assumed to result in no flow mixing at the core inlet plenum. Thus, cold edge temperatures are used to calculate the moderator reactivity insertion. This resulted in more positive reactivity being inserted and produced the maximum post trip R-T-P. In addition, the lower core flows resulted in - minimizing the transient DNBR. The results of the parametric analysis in break size indicate that the largest break size results in the maximum poet trip R-T-P and, thus, the minimum post trip DNBR. This occurs because the largest break size causes the greatest temperature reduction

and, thus, inserts the greatest magnitude of positive reactivity due to moderator reactivity feedback.

This results in a higher R-T-P and minimum post trip DNBR. Therefore, the results of the largest inside containment SLB with LOAC on turbine trip are presented herein. 2 The sequence of events for the 6.305 ft SLB with LOAC on turbine trip initiated from HFP conditions is given in Table 7 3.2-4. The reactivity insertion as a function of time is presented in Figure 7.3.2-1. The NSSS responses during the transient are given in Figures 7.3.2-3 through 7.3 2-8. The results of the analysis show that the HFP SLB causes the secondary pressure to rapidly decrease until a reactor trip on low stean generator pressure is initiated at 2.5 seconds. The CEAs drop into the core at 3.9 seconds and terminate the power and heat flux increases. Auxiliary feedwater is initiated at runout flow to the damaged side steam generator at time of trip. A LOAC power on turbine trip is assumed to occur at 3.4 seconds. At this time, RCPs begin coasting down and the diesel generators start coming on line. At 13 4 seconds, the diesel generators reach full speed and shutdown sequencer is initiated to load emergency systems. At 23.6 seconds the sa fety injection actuation analysis setpoint is reached and diesel generators switch from shutdown sequencer to LOCA sequencer to load emergency systems. At 28.6 seconds one HPSI pump is loaded on line and at 58.6 seconds the HPSI pump reaches full speed. i

37 The Steam Generator Isolation Analysis Setpoint is reached at 2.5 seconds. At 3.4 seconds, the MSIVs begin to close and are completely closed at - 9 4 seconds. The blowdown - from the intact steam generator is terminated at this time. An AFW isolation signal based on steam generator differential pressure is initiated at 8.8 seconds. At 28.8 seconds, the AFW block valves associated with the steam generator with lowest pressure (i.e., ruptured steam generator) are completely closed. The continued blowdown from the ruptured steam generator causes the core reactivity to approach criticality. The ruptured steam generator blows dry at 104.2 seconds, which terminates the cooldown of the RCS. A peak reactivity of.0775ap at 147.8 seconds is obtained. A peak R-T-P of 8.1% consisting of 4.7% fission power and 3 4% decay power, is produced at 147.8 seconds. The minimum transient DNBR SAFDL of 1.30 (MacBeth) is not violated during this event. SLB Outside Containment The outside containment SLB event initiated from HFP with a LOAC on turbine trip resulted in the maximum site boundary dose. Theresultsofparametrig MTC de that a break size of 0.65 ft -1.08x10ynstrated studies in break size and O a p / F resulted in the maximum number coupled with an MTC of of predicted fuel failures. 2 The sequence of events for a 0.65 ft SLB outside containment with LOAC on turbine trip initiated from HFP conditions is given in Table 7.3 2-5. The NSSS response during the transient are given in Figures 7.3.2-9 and 7.3 2-14. results of the analysis show that the SLB causes a reactor trip signal to be generated on icw steam generator pressure at 33.9 seconds. The trip The breakers open at 34.8 seconds, and the CEAs drop into the core at 35 3 seconds, terminating the power and heat flux increases. A LOAC power on turbine trip is assumed to occur at 34.8 seconds. Rextor power reaches its peak value of 1345 at 35.3 seconds The Steam Generator Isolation Analyses Setpoint is reached at 33.9 seconds. At 34.8 seconds the MSIVs begin to close, and are completely closed at 40.8 seconds. No return-to-power occurs. 1his is due to a slower cocidown rate allowing greater beneficial contribution to negative reactivity by decay heat and safety injection. 2 The 0.65 ft SLB outside containment shows that <15 of the fuel pins experience DNB. To be consistent with the intended enveloping nature of this analysis, 2.0% fuel failure was assumed for site boundary dose calculations. The resultant site boundary dose is: Thyroid (DEQ I-131) = 81 Rem Whole Body (DEQ Xe-133) =.3 Rem

38 Conclusions The results of the steam line break inside containment shows that the post-trip' minimum DNBR is within the design limit of 1 3 (MacBeth). The SLB outside containment results in _a site boundary dose which is within 10CFR100 guidelines. There fore, the results of the inside and outside containment SLB events with _ LOAC power on turbine trip is acceptable for Unit 1 Cycle.7. l

i 39 TABLE 7.3.2-1 KEY PARAMETERS ASSUMED IN THE INSIDE CONTAINMENT STEAM LINE BREAK EVENT INITIATED FROM HFP Parameter Units - Value Initial Core Power MWt 2754.0 0 Initial Core Inlet -F 550.0 Temperature Initial' RCS Pressure psia 2300.0 Initial Steam Generator psia 860.0 Pressure Low Steam Generator psia - 600.0 Pressure Trip Setpoint Steam Generator psid 365.0 Differential Pressure Setpoint Safety Injection psia 1578.0 ~ Actuation Signal Minimum CEA Worth % Ap -5.56 Available at Trip Doppler Multiplier ' 1.15 Moderator Cooldown Curve %Ao vs. See Figure density 7.3.2-1 Inverse Boron Worth PPM /5 Ap 85.0

1. p/ F

-2.5 Effective MTC x10 A S Fraction (including .0060 uncertainty)

40 TABLE 7.3 2-2 KEY PARAMETERS ASSUMED IN THE OUTSIDE CONTAINMENT STEAM LINE BREAK EVENT INITIATED FROM HFP ~ Value Par ameter Units Initial Core Power MWt 2754.0 Initial Core ~ Inlet F 550.0 Temperature - Initial RCS Pressure psia 2154.0 Initial Steam Generator. psia 860.0 Pressure Low Steam Generator psia 640.0 Pressure Trip Setpoint Safety Injection psia 1578.0 Actuation Signal 4 Minimum CEA Worth %ap -5.56 . Available at HFP Doppler Multiplier 0.85 Moderator Cooldown Curve 5ap vs. See Figure density 7 3 2-2 Effective MTC x 10-"Ap / F 2.7 to .27 0 6 Fraction (including .0044 uncertainty) g 1 e. .,w, - ~ - - - - ,w-c-- ,,.v..

41 TABLE 7.3.2-3 - ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR THE STEAM LINE BREAK EVENT Parameter Units Value Reactor Coolant System Maximum p Ci/gm 1.0 Allowable Concentration (DEQ I-131)I . Steam Generator Maximum Ajlowable pCi/gm 0.1 Concentration (DEQ I-131) . Partition Factor Assumed for 1.0 All Doses 2 3 1.80x10-4 Atmospheric Dispersion Coefficient sec/M M /sec 3.47x10-4 3 Breathing Rate 0 Dose Conversion Factor (I-131) R EM/Ci 1.48x10 I Tech Spec limits 20-2 hour accident condition L

42 TABLE 7.3.2-4 2 HFP 6.305 FT BREAK WITH LOAC, INSIDE CONTAINMENT Time (sec) Event Setpoint or Value 2 0.0 Steam Line Break Occurs 6.305 ft 2.5 Low Steam Generator Pressure 600.0 psia Analysis Trip Setpoint is Reached; Steam Generator Isolation Analysis Setpoint is Reached; 34 -Trip Breakers Open; Loss of AC On Turbine Trip; RCPs Coastdown Begins; Main Feed Rampdown Begins; MSIVs Begin to Close; Diesel Starting Sequence Begins 3.9 CEAs Enter Core 8.8 Steam Generator Differential 365.0 psid Pressure Setpoint Reached 9.4 Main Steam Isolation valves Fully Closed 13.4 Diesel Generator Up to Speed 23.2 Pressurizer Empties 23.6 Safety Injection Setpoint 1578 psia is Reached 28.6 Safety Injection Pumps Loaded on LOCA Sequencer 28.8 AFW Block Valves Closed Providing Auxiliary Feedwater to Intact S.G. only l 58.6 Safety Injection Pumps up to Full Speed 83 4 Main Feedwater Isolation Valves Completely Closed 7 104.2 Affected Steam Generator Blows Dry 147.5 Peak Power 8.1% of 2700 MWt 147.7 Peak Reactivity .007%Ap

43 TABLE 7 3 2-5 2 HFP, 0.65 FT BREAK OUTSIDE CONTAINMEtrr ~ Time (sec) Event Setpoint or Value 2 0.0 Steam Line Break occurs 0.65 ft 33.9 Reactor Trip Signal Generated 640.0 psia on Low Steam Generator Pressure 33 9 SGIS Signal Generated 640.0 psia 34.8 Trip Breakers Open; Loss of AC Power Begins; RCP Coastdown Begins 34.8 MSIVs Begin to Close 35 3 Maximum Power Reached 1345 35 3 CEAs Enter Core 40.8 MSIVs Fully Closed No Return to Power or Criticality Occurs 9 es n --,c-

---

.e

---4--- ,,--c, - - -. =,,, -w --v y---- wwr,--- ,-,-,--w,

44 8 i-i 7 2 LOOP-FULLP0hER INSIDE CONTAINMENT 6 Q.g5 $4 5A o 3 85 x g 2 25y 1 e 0 -1 -2 y, 40 45 50 55 60 65 MODERATOR DENSITY, LBM/FT3 l l l kAS& STEAM LINE BREAK EVENT FIGURE CO. calvert cliffs MODERATOR REACTIVITY VS MODERATOR DENSITY 7.3.2-1 Nuclear Power Plant l

45 \\ 8 i i i i 2-LOOP-FULL P.0WER OUTSIDE CONTAINMENT 6 q N iE 4 p 4W Eq 2 Ei8 0 2 '60 65 40 45 50 55 3 MODERATOR DENSITY, LEM/FT lastbS$$$!co. STE5MLINEBREAKEVENT FIGURE calvert cliffs OUTSIDE CONTAINMENT 7.3.2-2 Nuclear Power Plant MODERATOR REACTIVITY VS MODERATOR DENSITY ,,..-a ,_,,-..-n-., r,,-

---

.---.,_.e-,..,,..-n

--ym..g.-...-.,

-46 120 i t. i. i. i HOT FULL POWER 100 i-N 80 8% es 60 aWe W 14 0 8 20 (- ~ l 0 l 0. 50 100 150 200 250 300 TIME, SECONDS l f BALTIMORE STEAM LINE BREAK EVENT l GAS INSIDE CONTAINMENT acuaE calve t Cl s. Nuclear Power Plant CORE POWER VS TIME 7 3 2-3 i

47 110 i i i i i J HOT FULL POWER gg 88 77 E 8m 66 es a 55 5 u_ 44 =c d au 33 22 11 0 O. 50 100 .150 200 250 300 TIME, SECONDS BALTIMORE STEAM LINE BREAK EVENT FIMRE ca1[ehtci$$s INSIDE CONTAINMENT cAs 7.3.2-4 Nuclear Power Plant CORE HEAT FLUX VS TIME

48 650 i i i. I. I HOT FULL POWER 600 O E 550 5 500 MM TOUT C b 450 AW S TI eS4. E 400 350 I I I I 0 50 100 150 200 250 300 TIME, SECONDS BALTIMORE STEAM LINE BREAK EVENT 3AS & ELECTRIC CO. INSIDE CONTAINMENT FIGURE nuc$eaNobeNIant REACTOR COOLANT SYSTEM TEMPERATURES VS TIME 7.3.2-5 i

49 2500 i. i i i i 5 E HOT FULL POWER e5 5? 2000 &E 5 h 1500 m CE58 1000 5b5 500 I I I t Q 50 100 150 200 250 300 TIME, SECONDS GAS i T CO. FIGURE ca: vert c11ns INSIDE CONTAINMENT Nuclear Power Plant REACTOR COOLANT SYSTEM PRESSURE VS TIME 7.3.2-6

50 8 i i i i 1. MODERATOR HOT FULL POWER 6 4 2 o. DOPPLER N N 1 BORON 0 e ~ E TOTAL US"2 3-D _q SCRAM I I I I -6 0 50 100 150 200 250 300 TIME, SECONDS GAS & C CO. FIGURE INSIDE CONTAINMENT l calvert cliffs 7.3.2-7 rauclearPowerPlant; REACTIVITIES VS TIME i

51 900. i HOT FULL POWER 800 700 -1 i 5 E 600 ~ J$ h500 C1. 5 LE 400 55 e r 300 3 M 200 .100 Q i i 0 50 100 150 200 250 300 TIME, SECONDS BALTIMORE STEAM LINE BREAK EVENT l Asca1[ehtcNfs INSIDE CONTAINMENT FIGURE C' Nuclear Power Plant STEAM GENERATOR PRESSURE VS TIME 7.3.2-8 i

52 1 150 i i l-HOT FULL POWER 120 E 8h 90 es n 60 1 a_ u8 30 O I i O 60 120 180 240 300 TIME, SECONDS BALTIM0RE STEAM LINE BREAK EVENT GAS & ELECTRIC CO. OUTSIDE CONTAINMENT fant CORE POWER VS TIME ~ Nu ea oe

53 150 .i 3 HOT FULL POWER 120 5 m 8 s 90 m o M Nd 60 LEu u8 30 0 I I i i 0 60 120 180 240 300 ~ TIME, SECONDS l BALTIMORE STEAM LINE BREAK EVENT l GAS gigijag OUTSIDE CONTAINMENT' Calve t Ci s 7.3.2-10 Nuclear Power Plant CORE HEAT FLUX VS TIME-l

54 2500' i i i i i i i i i HOT FULL POWER E 2000 $u ~ ~ 1500 5m M g 1000 caS 8 500 0 0 30 60 90 120 150 180 210 240 270 300 TIME, SECONDS BALTIMORE STEAM LINE BREAK EVENT FIGURE cascalvehtcl$$s OUTSIDE CONTAINMENT 7.3.2-11 Nuclear Power Plant REACTOR COOLANT SYSTEM PRESSURE VS TIME

55 620 i i i i i i i i i 600 V HOT FULL POWER ~ 580 u_ o 560 540 a 520 500 5 h ll80 460 8 440 g TOUT h420 TAV a: 400 TIN 380 360 340 0 30 60 90 120 150 180 210 240 270 300 TIME, SECONDS BALTIMORE STEAM LINE BREAK EVENT GAS & ELECTRIC CO. OUTSIDE CONTAINMENT Calvert Cliffs 7.3.2-12 Nuclear Power Plant REACTOR COOLANT SYSTEM TEMPERATURES VS TIME

56 3 i i i HOT FULL POWER MODERATOR 1 D0PPLER d 3-D FEEDBACK q N N BORON g -1 !dft-TOTAL -3 CEA -5 -7 I I I I O 60 120 180 240 300 TIME, SECONDS STEAM LINE BREAK EVENT FIGURE GAS & E R CO. calvert cliffs OUTSIDE CONTAINMENT 7.3.2-13 Nuclear Power Plant REACTIVITIES VS TIME

57 1000~ i i i i i i i i i HOT FULE ~ POWER ~ 900 800 5 E 700 600 OE 500 8 3 400 u E 300 !!EW 200 </> 100 _1; O' 0 60 120 180 240 270 300 l l TIME, SECONDS BALTIMORE . STEAM LINE BREAK EVENT FIGURE GAS & ELECTRIC CO. OUTSIDE CONTAINMENT 7.3.2-14 STEAM GENERATOR PRESSURE VS TIME Nu r ou r ant L

+ 58 ' APPENDIX TO CHAPTER -7 4 'Feedline Break Event Introduction The :Feedline' Break event (FLB) was analyzed for Calvert Cliffs Unit 1 Cycle

7 to demonstrate: that the RCS pressure limit of 2750 psia is not exceeded'

.and that the site boundary doses do not exceed 10CFR100 guidelines. The break size and plant conditions ~ that were-found ' to~ result in L the worst pressure excursion -in the ' Calvert Cliffs Unit 2 Cycle 5 analysis,were - confirmed as worst for Unit.1 Cycle 7 and the results of this limiting break are presented herein. Discussion The FLB event is initiated by -a break in the main feedwater system (MFS) piping.. Depending on the break size and location and - the response of the MFS, ' the e ffects of a break can vary from a rapid heatup to a rapid cooldown of the Nuclear Steam Supply System (NSSS).

In order to discuss the possible effects, breaks are categorized as small, if the associated discharge flow is within the excess capacity of the MFS, and as large, if otherwise.

Break locations are identified with respect to the feedwater line reverse flow check valve. The reverse flow check valve of concern is located ' between the steam generator feedwater nozzle and the containment-penetration. Closure of the check valve, to prevent reverse flow from the steam generator, maintains the heat r emoval capability of that steam generator in the presence of a break upstream of the check valve. Feedwater -line breaks upstream of the reverse flow check valve can initiate one of the following transients. A break of any size, with MFS unavailable, will result in a Loss of Feedwater Flow (LOFW) event. A small pipe break with MFS available will result in no reduction in feedwater flow. Depending on the break size, a large break with MFS available will result in either a partial or a total LOFW event. Since FLBs upstream of the reverse flow check valve result in transients no more severe than a LOFW event, these FLBs were not analyzed. In addition to the possibility of partial or total LOIN, FLBs downstream of the check valve have the potential to establish reverse flow from the l affected steam generator back to the break. Reverse flow occurs whenever the MFS is not operating subsequent to a pipe break, or when the MFS is operating, but without sufficient capacity to maintain pressure at the break above the steam generator pressure. FLBs which develop reverse flow through the break are limiting with respect to primary overpressure.

Thus, only these FLBs were considered in the analysis, t

l FLBs downstream of the check valve with reverse flow may result in either a l RCS heatup or a RCS cooldown event, depending on the enthalpy of the reverse flow and the heat transfer characteristics of the a f fected steam ' generator. However, excessive heat removal through the feedwater line break is not considered in the analysis because the cooldown potential is less than that for the Steam Line Break (SLB) event. This occurs because SLEa have a greater potential for discharging high enthalpy fluid due to the location of the steam piping which is located above the feedwater piping within a steam generator. In adgition, the maximum break area for a 2 FLB is 2.2 ft in comparison to 6.305 ft for a SLB. ~

59 Unlike SLBs, FLBs cause a decrease in feedwater flow, resulting in lower steam generator liquid inventory which reduces the heat removal capacity. ~ The reduced heat transfer capability results in a rapid RCS overpressurization and, thus, it is the heatup potential of a FLB which was analyzed. A general description of the FLB event downstream of the check' valves, with the MFS unavailable and with low enthalpy break discharge, is given below. The loss of subcooled feedwater flow to both steam generators causes increasing steam generator temperatures,- decreasing liquid inventories and-decreasing water levels. The rising secondary temperature reduces the primary-to-secondary heat

transfer, which results in a

heatup and pressurization of the RCS. The heatup becomes more severe as the affected steam generator experiences a further reduction in its heat transfer capability due to decreasing liquid inventory. The heatup of the RCS and the depletion of liquid inventory in the steam generator will initiate a reactor trip on either High Pressurizer Pressure or Steam Generator Low Water Level. The RCS heatup can continue even after a reactor trip, due to a total loss of heat transfer-in the affected steam generator as the liquid inventory is completely depleted. The rise in RCS pressure causes the Pressurizer Safety Valves (PSVs) to open. The rise in secondary pressure is limited by the opening of the Main Steam Safety Valves (MSSVs). The opening of the PSVs and the MSSVs, in conjunction with the reactor trip (which reduces core po.ier to decay level), mitigates the RCS overpressurization. The reduction of liquid inventory in the unaffected steam generator in conjunction with low level S.G. signal initiates AFW flow to the unaffected steam generator. Automatic initiation of AFW is sufficient to provide a continued heat sink for the removal of decay heat. Analysis Assumptions and Initial Conditions The following is a discussion of the conservative assumptions and initial conditions chosen to maximize RCS pressure. Blowdown of the steam generator nearest the feedwater line break is modeled assuming frictionless critical flow as calculated by the Henry-Fauske correlation (Reference 9). The Feedwater Line Break location is conservatively modeled to be near the bottom of the steam generator, even though, in reality, the feedwater line nozzle is at a much higher elevation within the steam generator. The analysis assumes that saturated liquid is discharged through the break until the liquid mass reaches 5000 lbm, at which time saturated steam discharge is assumed. This assumption maximizes the liquid inventory discharge through the break, minimizes the energy removal from the primary by the steam generator, and thereby maximizes the RCS overpressurization. The analysis also assumes that the e ffective heat transfer area is decreased linearly as the steam generator liquid mass decreases. The mass interval over which the r ampdown is assumed to occur was conservatively chosen to model a rapid loss of heat transfer in the a ffected steam generator.

60 To maximize RCS pressure, the analysis conservatively credits only the high pressurizer pressure trip. This assumption maximizes the rate of change of pressure at the time of trip, and thus the peak pressure obtained following the trip. The analysis does not credit either the high containment pressure trip or the steam generator low water level trip. Table 7A-1 presents the initial conditions chosen to maximize the RCS pressure. A Moderator Temperature Coefficient curve corresponding to beginning of cycle conditions is assumed. This MTC, in conjunction with increasing coolant temperatures, adds positive. reactivity, and, thus, maximizes the rate of change of heat flux and pressure at the time of trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions is used in the snalysis. This FTC causes the least amount of negative reactivity feedback, allowing higher increases in both the heat flux and RCS pressure. An uncertainty factor of 15% is used in the analysis. An initial RCS pressure of 2154 psia is used in the analysis to maximize the rate of change of pressure at time of trip and, thus, the peak pressure obtained following a reactor trip. An initial steam generator pressure of 815 psia is assumed in the analysis. This pressure delays the opening of the Main Steam Safety Valves (MSSVs) and maximizes the peak RCS pressure. The Steam Dump and Bypass System (SDBS), the Pressurizer Pressure Control System (PPCS), the Pressurizer Level Control System (PLCS) and the Power Operated Relief Valves (PORV) are assumed to be in the manual mode of operation. his assumption enhances the RCS pressure increase, since the automatic operation of these systems mitigates the RCS pressure increase. This analysis conservatively assumed no automatic initiation of auxiliary feedwater. This assumption increases the RCS heatup and pressurization of concern in the FLB event. Credit was taken for manual initiation of auxiliary feedwater 10 minutes after reactor trip. The steam driven pump's auxiliary feedwater reaches the steam generator 58.0 seconds after the manual initiation. This includes 50 seconds required to open steam admission valves to the pump, 4.5 seconds for the pump to accelerate to speed and 3.5 seconds for the water to flood the piping and reach the steam generator. The flow to the steam generator is 434 spm/ leg. l The assumptions made to maximize the boundary site dose are given in Table 7A-2. During the event, two sources of radioactivity contribute to the site j boundary dose: (1) the initial activity in the steam generator and (2) the activity associated with primary to secondary leakage. The leakage through l the steam generator tubes is assumed to be the Technical Specification limit of 1.0 GPM. he initial primary and secondary activities are assumed to be at the Technical Specification limits of 1.0 p Ci/gm and 0.1 p Ci/gm, l respectively. The analysis assumes that all of the initial activity in the steam generators and the primary activity due to the tube leakage are released to the atmosphere with a decontamination factor of 1.0, resulting in the maximum site boundary dose.

61 Results The FLB event with Loss of AC (LOAC) power on -reactor trip results in the maximum RCS pressure. This occurs because LOAC power causes the Reactor Coolant Pumps to coastdown. The reduced core flow decreases the rate of heat removal

and, thus, maximizes the primary heatup-and over-pressurization.

Thus, only the results of the FLB event with LOAC power on reactor trip are presented herein. Figu: e 7A-1 presents the results of the parametric study to determine the break size which leads to the highest RCS peak pressure that was performed in the Calvert Cliffs Unit 2 Cycle 5 FLB event analysis. The trend and results of this parametric study were confirmed to be valid for the Unit 1 Cycle 7 FLB event. Figure 7A-1 shows that, initially, as the break size increases, so does the peak RCS pressure. This is due to faster water drainage out of the ruptured steam generator, which will cause a more rapid primary to secondary heat transfer rampdown. However, as the break size increases further, the greater steam relieving capacity of larger breaks (once the ruptured stemt generator feedwater nozzle uncovers) will ' offset the faster heat transfer rampdown and will result in lower peak essure. The highest peak pressure was obtained for a break size of 0.275 ft 2 The sequence of events for a 0.275 ft Feed Line Break downstream of the reverse flow check valve with LOAC on turbine trip is given in Table 7A-3 Figures 7A-2 through 7A-7 present the transient behavior of core power, core average heat flux, RCS temperatures, RCS pressure, steam generator pressure and steam generator liquid inventory for 1800 seconds of transient. 2 A 0.275 ft break in the main feedwater line is assumed to instantaneously terminate feedwater flow to both steam generators and establish critical flow from the steam generator nearest the break. During the first 24.5 seconds of the event, the absence of subcooled feedwater flow causes the secondary pressure and temperature to increase, which reduces the primary to secondary heat transfer. This causes the primary pressures and temperatures to increase. At 24.5 seconds, the liquid inventory in the ruptured steam generator is sufficiently depleted to cause l a further rampdown in the heat transfer rate. This causes the primary pressure and temperature to rapidly increase and at the same time causes the secondary pressure to decrease. The rapid increase in primary pressure initiates a High Pressurizer Pressure " rip at 27.1 seconds. At 27.9 seconds, the pressure reaches 2525 psia, at which time the Pressurizer Safety Valves (PSVs) open to mitigate the increase in primary pressure. At 28.4 seconds, the turbine stop valves close, increasing the secondary pressure. At 28.5 seconds, the CEAs begin to drop in to the core, inserting negative reactivity which mitigates the primary heatup. However, at this time, the Reactor Coolant Pumps (RCPs) l are assumed to initiate flow coastdown due to LOAC power on turbine trip. The rapid decrease in core flow slows down the rate of heat removal from the primary. At 28.6 seconds, the feed ring is uncovered and steam is discharged through the break, which mitigates the primary heatup. These competing effects result in a peak RCS pressure of 2749 psia at 31.1 seconds. The increase in secondary pressure is mitigated by the opening of the Main Steam Safety Valves in the undamaged and ruptured steam generator at 35.5 and 36.7 seconds, respectively.

62 At 185.6 s:conds a steam Generator Isolation Signallis generated. After appropriate d el ays,' the'~ Main Steam Isolation Valves (MSIVs) are fully closed at 198.5 seconds. This causes the pressure in the undamaged steam generator to increase and the pressure - in the damaged steam generator-to continue to decrease-through blowdown through the rupture. The water level in the undamaged steam generator continues to decrease as a result of boil-off. At about 229.0 seconds the liquid inventory in the undamaged steam generator is sufficiently depleted that there is no heat transfer from primary to secondary. This causes the primary pressure and temperature to increase again. The increase in primary pressure results in the opening of PSVs at 370.0 seconds. The analysis conservatively assumes that auxiliary feedwater is initiated manually at 600 seconds rather than automatically by the low steam generator level instrumentation much earlier in the event. This auxiliary feedwater reaches the. undamaged steam generator at 658 seconds and at a rate of 434 gpm. This auxiliary feedwater slowly reduces the primary heatup and at 779.5 seconds the primary safety valves are fully closed. The resultant site boundary dose calculated with the assureptions given in Table 7A-2 is: Thyroid (DEQ I-131) = 2.2 Rem Whole Body (DEQ Xe-133) < 0.1 Rem Conclusion The results of the FLB event with LOAC power on turbine trip shows that the peak pressure does not exceed the pressure upset limit of 2750 psia and that the site boundary doses are within 10CFR100 guidelines. ' Conservative with respect to current Technical Specifications.

-63 TABLE 7A-1 KEY PARAMETERS ASSUMED IN THE FEEDWATER LINE BREAK ANALT3IS Parameter Units Value Initial Core Power Level MWt 2754.0 Initial Core Coolant Inlet OF 550.0 Temperature Initial RCS Vessel Flow Rate gpm 370,000.0 Initial Reactor Coolant System psia 2154.0 Pressure Initial Steam Generator Pressure psia 815.0 Initial Pressurizer Liquid Volume ft 975.0 Effective Moderator Temperature x10""Ap/ F +0.2 0 Coe fficient Doppler Coefficient Multiplier 0.85 High Pressurizer Pressure psia 2470.0 -Trip Setpoint Auxiliary Feedwater Manual 600 sec. Actuation Initiation Steam Generator Differential psid 10.0 Pressure Setpoint CEA Worth at Trip %Ap -5.2 Reactor Regulating System Operating Mode Manual ** Steam Dump and Bypass System Operating Mode Manual ## Pressurizer Pressure Control Operating Mode Manual ## System Pressurizer Level Control System Operating Mode Manual ##

• • These modes of control system operation maximize the peak RCS pressure.

64 TABLE 7A-2 ' ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR THE FEED LINE BREAK EVENT Par ameter Units Value Reactor Coolant System Maximum pCi/gm 1.0 Allowable Concentration (DEQ I-131)' Steam Cenerator Maximum Ajlowable pCi/gm 0.1 Concentration (DEQ I-131) Partition Factor Assumed for 1.0 All. Doses Atmospheric Dispersion Coefficient sec/M3 1.80x10-4 2 M /sec 3.47x10-4 3 Breathing Rate 0 Dose Conversion Factor (I-131) REM /Ci 1.48x10 I Tech Spec limits 20-2 hour accident condition

65 . TABLE 7A-3 ' Time'(sec) Event Setpoint or Value 2 0.0 Break in Main Feedwater Line' .275 ft 24.5' - Heat Transfer Area Rampdown 19691 lbs in LHSG Begins 27.1 High. Pressurizer Pressure Trip '2470 psia Setpoint is Reached 27.9 Primary Safety Valves Begin 2525 psia to Open CEAs Begin to Enter Core; 28.4-LOAC on Turbine Trip; RCS Pumps Begin to Coast Down 31.1 Peak RCS Pressure 2749 psia 35.5 Undamaged Steam Generator 1000 psia Safety Valves Begin to Open 36.7 Damaged Steam Generator Safety 1000 psia Valves Begin to Open 40.0 . Maximum Steam Generator 1020.65/1007.46 psia Pressure Undamaged / Damaged 43.4 Primary Safety Valves are anzg psia Closed 65.3 Damaged Steam Generator 960 psia Safety Valves are Closed 68.5 Undamaged Steam Generator 960 psia Safety Valves are Closed 185.6 Main Steam Isolation Signal 600 psia 198.5 Main Steam Isolation Valves are Fully Closed 370.0 Primary Safety Valves Begin 2525 psia to Open 658.0 Auxiliary Feedwater Flow 434 spm Established to Undamaged Steam Generator 750.8 Undamaged Steam Generator Safety 1000 psi Valves Begin to Open 779.5 Primary Safety Valves are 2424 psia Fully Closed 1i..,_.. _. -

66 I 2800-s 12 .o', W LOAC g 2750 W !E u6 a. 5 2700 s2 m 5 8 2650 v 8b& 2600 O.2 0.4 ~'.6 0.8 1.0 1.2 l'.4 0 2 BREAK SIZE, FT

• PRESSURE INCLUDES ELEVATION HEAD

AS &

E R CO. FEED LINE BREAK EVENT FIGURE calvert cliffs REACTOR COOLANT SYSTEM PEAK PRESSURE VS BREAK SIZE 7A-1 Nuclear Power Plant

i 67 \\ 125 i i i i .i 2- - 100 E 85 75 r-E M c2 g 50 w 5 a 25 r i s i k 0 0 300 600 900 2200 1500 1800 TmE', SECONDS 1 l l l BALTIMORE FEEDWATER LINE BREAK EVENT GAS & ELECTRIC CO. FIGURE calvert cliffs WITH LOAC FOLLOWING REACTOR TRIP 7A-2 fluclear Power Plant CORE POWER VS TIME i l . - ~ -, -_.-.-.--...,~...,. ,...,,-.._,~....,_...,_.... -._ -- ----._,,-._ _ _--_____ _ _._ _,,-,_- - -,

6@ .125 i i i i I ~ 100 r-8% 75 M $d 50 W8 25 i 0 I I I l 0 300 600 900 1200 1500 1800 l TIME, SECONDS BALTIMORE FEEDWATER LINE BREAK EVENT FiebRE GAS & ELECTRIC CO. WITH LOAC-FOLLOWING REACTOR TRIP Nu!$eaNOerS$ ant CORE HEAT FLUX VS TIME 7g_3 I

1 69 100. i i i i I d 640 TOUT E N ) AVG T E h 580 D mMU IN W 520 5 e Eb 460 400 I I I I i 0 300 600 900 1200 1500 1800 TIME, SECONDS. BALTIMORE FEEDWATER LINE BREAK EVENT cAs calvehtc1Nfs WITH LOAC FOLLOWING REACTOR TRIP FIGURE Nuclear Power Plant REACTOR COOLANT SYSTEM TEMPERATURES VS TIME 7A-4 ,---mm

70 2850 i i i i I 5E 2600 .,l -uE NE 2350 EM M \\ $2100 -{ a \\ = N 1850 W 1600 0 300 600 900 1200 1500 1800 TIME, SECONDS l FEEDWATER LINE BREAK EVENT FIGURE GAS & E R CO. l calvert cliffs WITH LOAC FOLL0illNG REACTOR TRIP 7A-5 Nuclear Power Plant REACTOR COOLANT SYSTEM PRESSURE VS TIME D -r --m-ea-c,-e-----newn .w w ,.r-e-.~,-------------r,,-*- ,->----e

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71 1250 i i i i l ~ , 1000 -A G o_ 750 - he c. k 500 ~ = E 250 p ~~ i i i 0 0 300 600 900 1200 1500 1800 TIME, SECONDS BALTIMORE .FEEDWATER LINE BREAK EVENT GAS & ELECTRIC CO. WITH LOAC FOLLOWING REACTOR TRIP FIGURE nu!i ErlobeNIant STEAM GENERATOR PRESSURE VS TIME 7A-6

72 .140000 i ,i i j ) 120000 4 i 'I 100000 d-a E l 3 us l 80000 F

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I 60000 '1 8 \\ G .\\ , 40000 L \\ \\ \\ \\ \\ 20000 T \\ l I: \\ 0 I 4 0 300 600 900 1200 1500 1800 TIME, SECONDS BALTIMORE FEEDWATER LINE BREAK EVENT GAS & ELECTRIC CO, FIG 0RE calvert cliffs WITH LOAC FOLLOWING REACTOR TRIP 7A~7 Nuclear Power Plant STEAM GENERATOR MASS VS TIME i _-,___,,_.,_,-._-,____.____,,.-_.m.

73 8.0 ECCS-ANALYSIS 8.1 Introduction An ECCS performance analysis was performed for Calvert Cliffs Unit 1 Cycle 7 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Criteria for Emergency. Core Cooling Systems for Light Water. Reactors (Reference 1). The analysis justifies an allowable Peak Linear Heat Generation Rate (PLHGR) of 15.5 kw/ft. Ihis 'LHGR is equal to the existing limit for Calvert Cliffs Unit 1. The method of analysis _ and detailed results which support this value are presented in the following sections. 8.2 Method of Analysis The ECCS performance analysis for Calvert Cliffs 1 Cycle 7 consisted of an evaluation of the differences in the fuel rod conditions between Cycle 7 and Cycle 6, the reference cycle analysis. Acceptable ECCS performance was demonstrated for Cycle 6 in Reference 2 and approved by NRC in Reference 3 The blowdown and refill-reflood hydraulic calculations employed in the Cycle 6 evaluation apply to Cycle 7 since the hydraulic ~ parameters of the RCS remain unchanged. Therefore, only the fuel rod operating conditions for Cycle 7 were evaluated. The evaluation was performed using the NRC-approved FATES-3 (Reference 4 and 5) and the NRC-approved STRIKIN-II (Reference 6) computer codes to determine the limiting fuel rod conditions for ECCS performance for Cycle 7 for use in the comparison to the Cycle 6 fuel rod conditions. 8 3 Results Table 8-1 presents a comparison of the significant parameters for Cycles 6 and 7. The fuel rod conditions for the limiting case, i.e., maximum initial stored energy, for Cycle 7 are bounded by those of Cycle 6 for the following reasons. First, the initial fuel average stored energy as indicated by the fuel average temperature calculated by STRIKIN-II is greater for Cycle 6 than for Cycle 7 by 18 F. Secondly, the hot assembly average channel l PLHGR for Cycle 7 is less than that of Cycle 6. Also, the limiting hot rod radiation heat transfer enclosure for Cycle 7 was found to be letss [ severe than that for Cycle 6. l r The fuel rod conditions at extended burnup for Cycle 7 are bounded by those l of Cycle 6. The hot rod gas pressure at extended burnup for Cycle 7 is 76 psia lower than the corresponding pressure for Cycle 6 as shown in Table 8-l 1. The initial fuel average temperature at extended burnup for Cycle 7 is 66 F lower than at the limiting (maximum stored energy) burnup and is 0 nearly identical (2 F higher) to the value at extended burnup in Cycle 0 6. STRIKIN-II transient calculations confirmed that the 2 F difference is more than compensated for by the less severe radiation enclosure and lower hot assembly average channel PLHGR for Cycle 7 For these reasons it is concluded that the peak clad temperature and oxidation percentages for Cycle 6 conservatively apply to Cycle 7. r

74 8.4 Conclusions For the reasons presented in Section 8.3, the results of the Calvert Cliffs Unit 1 Cycle 6 ECCS performance analysis conservatively apply to Cycle 7. In the Cycle 6 analysis the peak clad temperature was calculated to be 2038 F as compared to the acceptance criteria limit of 2200 F. The _ peak local and core wide clad oxidation percentages were calculated to be. 8.55 and < 0.515, respectively, as compared to the acceptance criteria limits of.175 and 15, respectively. Therefore, operation at a PLHGR of 15.5 ' kw/ft and a power level of 2754 MWt (1025 of 2700 MWt) will result in acceptable ECCS performance for Calvert Cliffs Unit 1 Cycle 7. 1 ..e_--,.---,,,y 1-,

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75 TABLE 8-1 CALVERT CLIFFS UNIT 1 CYCLE 7 ECCS ANALYSIS COMPARISON OF SIGNIFICANT PARAMETERS WITH CYCLE 6 Parameter Cycle 6 Cycle 7 Power Level (102% of Nominal), MWt 2754 2754 Peak Linear Heat Generation Rate, Hot Assembly, Hot Channel, kw/ft 15.5 15.5 Peak Linear Heat Generation Rate, Hot Assembly, Average Channel, kw/ft 13 14 12.52 Limiting Burnup Case Fuel Conditions Fuel Average Temperature at PLHGR, OF# 2213 2195 Fuel Centerline Temperature at PLHGg, O F# 3634 3578 Gap Conductance at PLHGR, BTU /hr-ft, F# 2025 1928 Hot Rod Gas Pressure, psia

• 1251 1198 Hot Rod Burnup, MWD /MTU 3000 1000 Extended Burnup Case Fuel Conditions Fuel Average Temperature at PLHGR, OF#

2127 2129 Fuel Centerline Temperature at PLrlGg, O

• F 3551 3566 Gap Conductance at PLHGR, BTU /hr-ft, OF#

2370 2441 Hot Rod Gas Pressure, psia

• 2191 2115 Hot Rod Burnup, MWD /MTU 34000 52500**
• Values are those at the indicated hot rod burnup as calculated by STRIKIN-II at 15.5 kw/ft.
  • The limiting conditions for extended burnup have changed to the end of cycle maxitaum burnup due to the NRC mandated grain size restriction imposed on the FATES 3 code (Reference 5).

76-9.0 TECHNICAL SPECIFICATIONS The Technical ' Specification changes which must be made in order to make the Calvert Cliffs Unit 1 Technical Specifications valid for the operation of Cycle 7 are presented in this section. Table 9-1 presents a summary of the Technical Specification changes. Table 9-2 presents the explanations for the changes summarized in Table 9-1. The requested Technical Specification codifications for Unit 1 Cycle 7 (Table 9-1) are very similar to those changes requested for the reference cycle (Unit 2 Cycle 5, References 1 and 2). 'Ihere are two significant differences compared to the requested changes for Unit 2 Cycle 5 both of which are supported by the generic Steam Line Rupture analysis presented in Section 7. These differences are: 1. The shutdown margin is being lowered to 4.3% A k/k to reduce operating requirements with regard to shutdown boron levels. This change is consistent with the generic SLB analysis presented herein (see Table 7-

2) and existing safety analyses (see note to Table 5-2).

2. The moderator temperature coefficient negative limit is being increased to accommodate the effects of extended burnup. Following Table 9-2, for each Technical Specification which must be modified, either: 1. the existing page with the intended modification, 2. the already modified page with a now figure, or 3 a sample page (area to be modified or added identified in enclosed area with "#") is provided. l _.. - _ _ _ -.}}