ML20041A644

Application for Amend to License DPR-53 Changing STS to Allow Operation of Sixth Cycle

ML20041A644
Person / Time
Site:	Calvert Cliffs
Issue date:	02/17/1982
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20041A639	List: ML20041A638 ML20041A644 ML20041A648
References
NUDOCS 8202220373
Download: ML20041A644 (176)

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ENCLOSURE CALVERT CLIFFS UNIT 1 CYCLE 6 I

LICENSE APPLICATION 8202220373 820217 PDR ADOCK 05000317 P PDR

Calvert Cliffs Unit 1, Cycle 6 License Application CONTENTS

1. INTRODUCTION AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 f

2. OPERATING HISTORY OF THE REFERENCE CYCLE . . . . . . . . . . . . . . . . . . . . . . 3

3. GENE RAL DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

4. FUEL SYSTEM DESIG N . . . . . . . . . . . . . . . . . .,. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

5. N UCLEA R D ESIG N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 6.

THERMAL-HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27

7. TRANSIENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 l 8.

, ECCS PERFORMANCE ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 151 9.

TECHNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 162 l

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10.

STA RTUP TESTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 196 11.

REFE RENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 197

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1.0 I!!TRODUCTIO!! AllD

SUMMARY

This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit I during its sixth fuel cycle at fbil rated power of 2700 WT. All planned operating conditions remain the same as those for Cycle 5. The core will consist of presently operating F and G assemblies, fresh Batch H assemblies, eight D assemblies discharged from Calvert Cliffs Unit 2 Cycle 3 and one D assembly discharged from Calvert Cliffs Unit I Cycle 4.

Plant operating requirements have created a need for flexibility in the Cycle 5 termination point, ranging frem 12,000 MWD /T to 13,000 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 6 _

conditions are enveloped, provided the Cycle 5 termination points fall within the above burnup range. The analysis presented herein will accomodate a Cycle 6* 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 I Cycle 5 safety analysis described in Reference 1, as modified in Reference 2, hereafter referred to as the " reference 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. 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 frem the reference cycle, proposed modifications to the plant Technical Specification are provided and are justified by the analyses reported herein.

The Cycle 6 average discharge exposure will be approximately 33,700 MWD /T, which is a small increase from the average discharge exposure for the reference cycle which will be approximately 30,800 WWT. 'Where appropriate, additional discussion concerning this increase in average discharge exposure has been included. Since all analyses address fuel

a cxpcsura cxplicitly, this increasa in cveraga discharge cxpcsure his bein

• cxplicitly tcccunt2d for in th2 rcsults of the Cycle 6 analyses presented herein. In all instances it has been concluded that such an increase dces not result in any significant changes with regard to the reference cycle.

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3 2.0 OPERATING HISTORY CF THE REFERENCE CYCLE

. Calvert Cliffs Unit I is prese..tly operating in its fifth fuel cycle utilizing Batch D, E, F, and G fuel assemblies. Calvert Cliffs Unit I Cycle 5 began operatien en January 12, 1981 and reached full ;iewer on January 23, 1981. The Cycle 5 startup testing was reported to the NRC in Reference 3 The reactor has operatw! up to the present time with the core reactivity, reactivity coefficients, pcwer distributions and peaking

, factors having follcwed the calculated predictions very closely.

It is presently estimated that Cycle 5 will terminate on or about April 17, 1982. The Cycle 5 termination point can vary between 12,000 MWD /T and 13,000 MWD /T to accomodate the plant schedule and still be within the assumptions of the Cycle 6 analyses. As of January 9,1982, the Cycle 5 burnup had reached 9940 MWD /T.

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6 30 GE!!E?at DESCRIPTIC!! , ,

Re Cycle 6 core will censist of the nu=ber tmc types of asse=blies and fuel batches as described in Table 3-1. Tne pri=ary change to the cere in Cycle 6 is the re= oval of 52 Batch E asse=blies, 28 Batch F asse=blies and 1 Batch D assembly. Rese assemblies will be replaced by 40 Batch H (4.00 w/o enrichment) asse=blies, 32 Batch H/ (3 55 w/o enrich =ent) asse=blies, 8 Batch D assemblies discharged frem Calvert Cliffs Unit 2 Cycle 3 and 1 Batch D assembly discharged frem Calvert Cliffs Unit 1 Cycle 4. Re 32 low enrichment Batch H/ asse=blies centain 8 burnable poisen pins per assembly. Figure 3-1 shows the fuel manage =ent pattern to be e= ployed in Cycle 6. Figure 3-2 shows the locations of the poisen pins within the lattice of once-burned Batch G/ assemblies and the fuel rod locations in unshi==ed asse=blies; Figure 3-3 shows the poisen pin locations within the lattice of the fresh Batch H/ fuel. Eis fuel cantge=ent pattern will acccc:=cdate Cycle 5 ter=ination burnups freci 12,000 K4D/T to 13,0C0 IND/T.

He Cycle 6 ccre leading pattern is 900 rotationally sy:m:etric. Tnat is, if one quadranY, of the core were retated 900 into its neighboring quadrant, each asse=bly would be aligned with a similar asse=bly. This si=ilarity includes batch type, nu=ber of fuel reds, initial enrichment and burnup. It does not include guide tube modifications, de=enstratien red 1ccations and an incenel irradiation experi=ent.

Figure 3-4 shows the beginning of Cycle 6 asse=bly burnup distributien fer a C /cle 5 ter=inatien burnup of 12,500 M4D/T. The initial enrich =ent of the fuel assemblies is also shown in Figure 3-4 Figure 3-5 shcws the end of Cycle 6 asse=bly burnup distribution. The end of Cycle 6 ccre average exposure is approxi=ately 25,300 K4D/T, and the average discharge expcsure is approxi=ately 33,700 MWD /T.

I 31 C-E EPRI Test Asse=bly The Batch D assembly centaining 14 C-E EPRI test pins was described in the Calvert Cliffs Unit 1 Cycle 5 Refueling Licensing A==endment, Reference 1. Tnis asse=bly will be discharged at the end of Cycle 5.

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5 32 Inconal Irradittion Experiment The inconel irradiation experiment was also described in Reference 1.

The t'hree assemblies containing experimental inconel tubes will be inserted into high flux locations in Cycle 6 for a second cycle of irradiation.

3 3 SCOUT Demonstration Assembly The SCOUT demonstration assembly was described both in Reference 4 and Reference 5. Cycle 5 is the second cycle of irradiation for this Batch F test assembly. Prior to returning to the core for its third cycle of irradiation in Cycle 6, 4 test pins containing annular fbel pellets will

.be removed from the assembly for testing. These pins will be replaced by fuel pins containing standard Batch D fuel pellets which have undergone 3 previous cycles of irradiation.

34 PROTOTYPE Demonstration Assemblies The PROT 01TPE der:cnstration assemblies were described in Reference 1.

. These assemblies will be placed in symmetric locations in the core in Cycle 6 for a second cycle of ir;adiation.

TABLE 3-1 Calvert Cliffs Unit 1 Cycle 6

_ Core Loading 4

Batch Total

! Avera9e Initial Number Total Initial Poison Poison of Number Assembly Number of Enrichment (Burnup E0C5 = Rods Per Poison of Fuel Loadin9 Designation Assemblies wt% U-235 12,500} Assembly (3) wt% 4B C (3) Rods Rods i

0042 III 1 3.03 32,200 0 0 0 176 D

I 8 3.03 19,900 0 0 0 1,408 i

F 44 3.03 23,000 0 0 0 7,744 G 40 3.65 10,200 0 0 0 7,040 G/ 52 3.03 14,000 8 3.03 416 8,736 11 40 4.00 0 0 0 0 7,040 11/ 32 3.55 0 8 3.03 256 5,376 TOTALS 217 672 37,'520 NOTES :

(1) Discharged from Calvert Cliffs-! Cycle 4, quarter-core location 40 (2) Discharged from Calvert Cliffs-II Cycle 3, quarter core location 18 and 39 10 concentration equals 0.2685 gms 810/ inch (3) Shim B

7' 1 2 .

H H-3 4 5 6 7 ,,

H H HI F G

. 8 9 10 11 12 13 H G D G Gl Gl 14 15 16 17 18 19 20 H HI F HI F HI F 21 22 23 24 25 26 27 28 H G F Gl F G F G 29 30 31 32 . 33 34 35 36 H D HI F G GI GI Gl 37 38 39 40 41 42 43 44 HI G F G Gl G Gl GI 45

" 46 47 48 49 50 51 52 53 H I GI HI GI 54 H

55 " 56 57 58 59 60 61 62 G GI F G GI Gl Gl D

• L0 CATION OF DEMONSTRATION ASSEMBLY (SCOUT)

" LOCATION OF PROTOTYPE ASSEMBLIES BALTIMORE Figure GAS & ELECTPIC CO. CALTERT CLIFFS UNITI CYCLE 6 Ccivert Cliffs CORE MAP 3-1 Nuclecr Power Plant

UNSHIMMED ASSEMBLY

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l GI-8 POISON R0D ASSEMBLY X X X '

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X IX

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__ FUEL R0D LOCATION 2 POISON ROD LOCATION BALTIMORE R"S GAS & ELECTRIC CO, CALVERT CLIFFS UNIT I CYCLE 6 calvert clins ASSEMBLY FUEL AND OTHER R0D LOCATIONS 3-2 Nuclecr Power Plant

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Hl-8 POISON R0D ASSEMBLY U X X IX -X X X f

X X FUEL R0D LOCATION 2

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POIS0N R0D LOCATION i

GAS ELE T IC CO-CALVERT CLIFFS UNIT I CYCLE 6 Rsure l

coivere ciirr, ASSEMBLY FUEL AND 0THER R0D LOCATIONS 3-3 Nuclear Pcwer Picnt

INITIAL ENRICHMENT. W L U-235 1 2 BOC6 BURNUP (MWb/T), EOC5 s 12,500 MWu/T 4.00 4.00 0 0 3 4 5 6 4.00 4.00 3.55 3.03 7 3.65 0 0 0 23,000 11,100

" 8 9 10 11 12 13

~ 4.00 3.65 3.03 3.65 3.03 3.03 8,800 19,900 10,000 13,700 12,700 14 15 16 17 18 19 20 4.00 3.55 3.03 3.55 3.03 3.55 3.03 0 0 23,700 0 24,700 0 23,900 21 22 23 24 25 26 27 L28 4.00 3.65 3.03 3.03 3.03 3.65 3.03 3.65 0 .8, 800 23,500 12,700 22,300 9,100 21,800 12,000 29 30 31 32 33 34 35 36 4.00 3.03 3.55 3.03 3.65 3.03 3.03 3.03 0 19,900 0 20,700 12,100 15,200 15,000 15,200 37 38 39 40 41 42 43 44 3.55 3.65 3.03 3.65 3.03 3.65 3.03 3.03 45 0 10,000 24,700 9,100 15,200 11,400 15,500 14,800

4. 0 46 47 43 4Y 30 31 52 53 3.03 3.03 3.55 3.03 3.03 3.03 3.55 3.03 54 23,000 13,700 0 21,800 15,000 15,500 0 15,300 4"f 55 3.65 56 3.03 57 3.03

'58 3.65 59 60 61 '62 3.03 3.03 3.03 3.03 l 11,100 12,700 23,900 12,000 15,200 14,800 15,300 32,200 1

BALTIM ORE GAS & ELECTRIC co. CALVERT CLIFFS UNIT I CYCLE 6 ": S'"

• c=lvert c!iffs ASSEMBLY AVERAGE BURNUP AT BOC -

Nuclecr ?:.ver .:! nt 3-4 AND INITIAL ENRICHMENT uISTRIBUTION-

11 1 2 10,200 13, d00 3 4 5 6 7 10,900 14,200 15,100 35,200 26,700 8 9 10 11 12 13 11,700 23,600 32,900 26,500 28,800 27,700 14 15 16 17 18 19 20 11,700 15,800 36,100 17,300 37,800 17, 800 36,900 21 22 23 24 25 26 27 28 10,900 23.,600 36,000 26,900 35,400 25,600 35,100 27,700 29 30 31 32 33 34 35 36 14,200 33,000 17,400 34,100 28,300 30,500 29,500 29,500 37 38 39 40 al a2 43 da 15,100 26,500 37,800 25,600 30,400 28,000 30,500 29,400 45 10,200 48 49 50 51 52 53 46 a7 54 35,200 28,900 17,900 35,100 29,500 30,500 18,100 29,900

~

13,400 55 56 57 58 59 60 61 62 26,700 27,700 36,900 27,700 29,500 29,400 29,900 42,800 -

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^ CALVERT CLIFFS UNIT I CYCLE 6 Figure GAS E E T IC CO' '

calvert cliff: ASSEMBLY AVERAGE BURNUP AT E0C (MWblT) 3-5 Nuclect Pew er Plent j

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4.0 FUEL S'fSTEM DESIGN 4.1 Mechanical Cesi:;n The mechanical cesign fcr the standard Batch H reload fuel is identical to that of the standard Batch G fuel used in the reference cycle with the exception of a .200 inch reduction in the overall length of the fuel rods. The length reduction will provide additional clearance for fuel rod length increase during the lifetime of the Batch H fuel.

The mechanical designs of the Batch D, F, and G fuel asse=blies were described in References 6, 4, and 1, respectively. Details of the Calvert Cliffs Unit 2 Batch D fuel assemblies that will be used in Cycle 6 can be fcund in Reference 7.

C-E has perfor=ed analytical predictions of cladding creep collapse time for all Calvert Cliffs Units 1 and 2 fuel batches that will be irradiated in Cycle 6 and has Ioncluded that the collapse resistance cf all standard fuel rods is sufficient to preclude collapse during their design lifetime. This lifetime will not be exceeded by the Cycle 6 duration (Table 4-1). These analyses utilized the CEPAN cc=puter code (Reference

8) and the analysis procedures described in Reference 9 TABLE 4-1 Batch Minimum ECC 6 Collapse Time Excesure D (Calvert Cliffs-1) >35,000 EFFH 32,800 EFFH D (Calvert Cliffs-2) >35,000 26,000 F > 35,000 23,900 G 2 27,500 20,000 H > 27,500 10,400

13 All batches of fuel were also reviewed for dimensional changes using the SIGREEP model described in Reference 10. All clearances were found to be adequate during Cycle 6.

The clad collapse information and conclusions with regard to dimensional changes are applicable for the test rods in the SCOUT and PROTOTYPE bundles (see Sections 3 3 and 3.4, respectively).

The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the Batch H fuel are identical to those of the Batch D, F, and G fuels to be included in Cycle 6. Thus, the chemical or metallurgical performance of the Batch H fuel will remain unchanged from the performance of the Cycle 5 fuel.

4.2 Hardware Modifications to Mitigate Guide Tube Wear All standard fuel assemblies which will be placed in CEA locations in Cycle 6 will have s$ainless 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

11. A total of four Batch F demonstration fuel assemblies, containing modified guide tubes rather than stainless steel sleeves to mitigate guide tube wear, will be placed in CEA locations in Cycle 6. These four assemblies contain the same modifications as those described in Appendix E of Reference 1.

4 3 Thermal resign The thermal performance of composite fuel pins that envelope the various l

fuel assemblies present in Cycle 6 (fuel batches F, G and H; Batch D assemblies from Unit 2 Cycle 3; assembly DC42 from Unit 1 Cycle 4) have been evaluated using the FATES 3 version of the fuel evaluation model (References 12 and 13). 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

6. In addition, the SCOUT and PROTOTYPE test pins were analyzed, and found not limiting with respect to thermal performance over their respective burnup ranges.

14 5.0 NUCLEAR DESIGN , ,

Cyclo 6 of Unit 1 will ba tha first fual cycle for either Calvert Cliffs Unit for which the average discharge exposure will be as high as 33,700

%D/T. All analyses address fuel exposure explicitly and the small increase in average discharge exposure over that of the reference cycle has not yielded, of itself, significant ' changes in core parameters. 1 Furthermore, those assemblies with burnups in excess of 24,000 LT/T at '

ECC contain maximum 1-pin peaks which are substantially below the maximum 1-pin peak in the core (See Section 6.2).

5.1 Physics Characteristics

5.1.1 Fuel Management The Cycle 6 fuel management employs a mixed central region as described in Section 3, Figure 3-1. The fresh Batch H fuel is comprised of two sets of assemblies, each having a unique enrichment in order to minimize radial power peaking. There are 40 assemblies with an enrichment of 4.00 wt% U-235 and 32 assemblies with an enrichment of 3 55 wt% U-235 and 8 poison -

shims per assembly. With this loading, the Cycle 6 burnup capacity for full power operation is expected to be between 13,200 %T/T and 13,800

' ET/T, depending on the final Cycle 5 termination point. The Cycle 6 core characteristics have been examined for Cycle 5 terminations between 12,000 and 13,000 LT/T and limiting values established for the safety analyses.

The loading pattern (see Section 3) is applicable to any Cycle 5 termination point between the stated extremes.

Physics characteristics including reactivity coefficients for Cycle 6 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 safety analyses are -different from those displayed in Table 5-1 and are

, typically chosen to conservatively bound predicted values with accommodation for appropriate uncertaint'es and allcwances.

Table 5-2 presents a summary of CEA shutdown worths and reactivity

, allowances for the end of Cycle 6 zero power steam line break accident and a ecmparison to reference cycle data. The ECC zero power steam line break 4

15 w:s selected sinca it is tha most limiting ztro pow r stram lina break accid;nt, and thus provid:s tha basis for establishing tha Tcchnical Specification required shutdown margin.

The power dependent insertion limit (PDIL) for bank 2 insertion has changed relative to the reference cycle. The new PDIL curve is shown in Section 9 Table 5-3 shows the reactivity wcrths of various CEA groups calculated at full power conditions for Cycle 6 and the reference cycle.

5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC6, M006 and ECC6, respectively, that are characteristic of the high burnup end of the Cycle 5 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 higher burnup end of the Cycle 5 shutdown window tends to increase the power peaking in this axial central region of the core for Cycle 6.Theplanar@adialpowerdistributionsfortheaboveregionwith '

CEA Group 5 fully inserted at beginning and end of Cycle 6 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 5 shutdown window.

The radial power distributions described in this section are calculated data without uncertainties or other allowances. However, the single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel asse=bly 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 6. These conservative values, which are used in Sea. tion 7 of this document, establish the allowable limits for power peaking to be observed during operation.

The range of allowable axial peaking is defined by the limiting conditions for operation covering axial shape index (ASI). '41 thin these ASI limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum three-dimensional er total peaking

frcter anticip5ted in Cyclo 6 during nor=al bass lord, all rods out ep ration at full pcwcr is 1.86, not including unctrtcinty allowances and augmentation facters.

5 1 3 Safety Related Data 5131 Ejected CIA Data ihe maximum reactivity worths and planar radial pcwer peaks associated with an Ejected CEA event are shewn in Table 5-4 for Cycle 6 and the reference cycle (Reference 14). These values enecmpass the worst conditions anticipated during Cycle 6 for any expected Cycle 5 terminatien point. The values shown for Cycle 6 are the safety analyses values which are conservative with respect to the actual calculated values.

5 1 3 2/5 1 3 3 Drepped CEA/ Augmentation Facters The Cycle 6 safety related data for these sections are identical to the safety related data'used in the Reference Cycle, as presented in Sections 5.1 3 2 and 5 1 3 3 of Reference 1.

5.2 Analytical Input to In Ccre Measurements In-ccre detector measurement constants to be used in evaluating the reload cycle pcwer distributions will be calculated in the canner described in Reference 15, which is the same method used for the reference cycle.

, 53 Nuclear Design Methodology Most analyses have been performed in the same manner and with the same mechedologies used for the reference cycle analyses. The cnly analytical tools which have been changed are:

a) The DIT assembly spectrum code (References 16 and 17) which is based on integral transpcrt theory was introduced into the Cycle 6 design.

This change has led to a6reement with =easurements en reactivity and pcwer distributiens that is substantially improved frem previcusly used cress-section generation methods (Reference 18).

17 DIT-btsed cross sections wero g;ncreted for both ROCS rnd PDQ nuclcar design calculations for all hot zero power and hot full powei conditions. Tne use of DIT based cross sections was previously presented in the ANO-2 Cycle 2 reload analysis and subsequently approved in Reference 19 by the NRC for Cycle 2.

b) The calculation of space-time scram insertions (FIESTA, Reference 20).

5.4 Uncertainties in Measured Power Distributions The power distribution measurement uncertainties which are applied to Cycle 6 are the same ar crese applied to the Reference Cycle, as presented in Section 5.4 of Reference 1.

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Ttble 5-1 Calvert Cliffs Unit 1 Cycle 6 Ncminal Physics Characteristics Units Reference Cycle Cycle 6_

Dissolved Eeron Dissolved Bcron centent for Criticality, CEAs Uithdrawn Hot Full Power, Equilibrium Xenen, BOC PPM 1010 1025 Baron Worth Hot Full Pcwer BOC PPM /".Ar 101 106 Hot Full Pcwer EOC PPM /%af 83 85 Reactivity Ccefficients (CEAs Withdrawn)

Moderater Te=perature Coefficients, Het Full Pcker, Equilibrium Xenon Beginning of Cycle 4 C

-0.1 10 Af/ F -0.2 End of Cycle 10-4Af/ F -1 9 -2.1 Doppler Coefficient Hot Zero Pcwer BOC 10-54f/F -1.55 -1.48 Hot Full Pcwer EOC 10-5ar/ F -1.21 -1.24 0

Hot Full Pcwer EOC 10-3A (/ F -1.40 -1.47 Total Celayed Neutron Fraction,# eff BOC .00628 .00631 EOC .00521 .00541 Neutron Ceneration Time, /

• EOC 10-Osec 24.4 23 8 EOC 10-6 sec 29 7 30.2

19 Tcblo 5-2 Calvert Cliffs Unit 1 Cycle 6 Limiting Values of Reactivity Worths and Allowances for Hot Zero Power Steam Line Break, % Af End-of-Cycle (ECC)

Reference Cycle Cycle 6

1. Worth of all CEA's Inserted 94 10.2

2. Stuck CEA Allowance 2.2 2.6 3 Worth of all CEA's Less Highest Worth CEA Stuck Out 72 7.6

4. Zero Power Dependent

Insertion Limit CEA Bite 2.0 17 5 Calculated Scram Wo rth 5.2 5.9

6. Physics Uncertainty (10% of Item 5) 0.5 0.6 7 Net Available Scram Worth (Item 5 mJnus Item 6) 4.7 53

8. Technical Specification Shutdown Margin 43 53 9 Margin in Excess of Technical Specification Shutdewn Margin 0.4 0.0

20 Tcble 5-3 - -

Calvert Cliffs Unit 1 Cycle 6 Reactivity 'dorth of CEA Regulating .

GrcupsatHotFullPcwer,%af Beginning of Cycle End of Cycle Regulating Reference Reference CEA's Cycle Cvele 6 Cycle Cycle 6 Group 5 0.49 0.48 0.57 0.65 Grcup 4 0 32 0.31 0 39 0.33 Group 3 0 97 0.84 0 93 1.04 Notes Values shown assu:ne sequential group insertien

21 Table 5-4 Calvert Cliffs Unit 1 Cycle 6 C7A Ejection Data Limiting Values steference Cycle Cycle 6 Safety Analysis Value Safety Analysis Value Maximum Radial Power Peak Full Power with Bank 5 in::erted; worst CEA ejected 3.15 2.75 Zero power with Banks 5+4+3 inserted; worst CEA ejected 9,4 9.4 Maximum Ejected CEA Worth ( g )

Full power with Bank 5 inserted; worst CEA ejected 0.22 0.22 Zero power with Banks 5+4+3 inserted; worst CEA ejected 0.63 043 Notes

1) Uncertainties and allcwances are included in the above data.

2) The Cycle 6 safety analysis values are conservative with respect to the actual Cycle 6 calculated values.

0.77 1.04 I

0.82 1.09 1.10 0.85 1.14 0.87 1.10 0.94 1.21 1.04 1.03 1

0.87 1.13 0.82 1.21 0.86 1.23 0,83 4

0.82 1.11 0.83 0.95 0.87 1.18 0.87 2.07 1.09 0.94 1.21 0.88 1.16 1.04 0.96 0.93 X

1.10 1.21 0.86 1.18 1.03 1.16 0.97 0.92 0.77 0.85 1.05 1.23 0.87 0.96 0.96 1.18 0.91 1.04 1.14 1.03 0.83 1.07 0.93 0.92 0.91 0.64 NOTE: X MAXIMUM 1-PIN PEAK - 1.56 l

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CALVERT CLIFFS UNIT I CYCLE 6 l ASSEMBLY RELATIVE POWER DENSITY AT BOC, c=ivert cliff, 5-1 Nuclear P:wer Pice.t EQUILIBRIUM XENON

4 0.75 0.96 0.78 1.03 1.09 0.83 1.08 0.86 1.06 0.92 1.17 1.04 1.04 s

0.86 1.16 0.85 1.25 0,88 1.28 0.87 0.78 1.06 0.85 0.99 0.88 1.16 0.89 1.08 1

1.03 0.92 1.25 0.89 1.12 1.03 0.99 0.97 1.09 1.17 0.88 1.15 1.02 1.14 1.01 0.99 0.75 0.83 1.05 1.28 0.89 0.98 1.01 1.30 1.01 0.96 X 1.08 1.04 0.87 1.08 0.97 0.99 1.01 0.74 NOTE: X-MAXIMUM 1-PIN PEAK-1.47 gas $#E$TSfcco- CALVERT CLIFFS UNIT I CYCLE 6 Figure cerve-r cntr, ASSEMBLY RELATIVE POWER DENSITY AT 7 GWDlT Nuclect P:wer Plan, EQUILIBRIUM XENON 5-2

O.78 0.96 0.81 1.04 1.15 0.85 1.05 0.91 1.07 0.94 1.16 1.04 1.03 0.91 1.25 0.90 1.31 0.90 1.31 0.89 X

0.80 ,,

1.07 0.90 1.01 0.89 1.11 0.88 1.05 1.04 0.94 1.31 0.89 1.05 0.96 0.93 0.92 1.15 1.16 0.90 1.11 0.95 1.05 0.96 0.95

,. 0.78 0.85 1.04 1.31 0.88 0.93 0.95 1.28 0.98 0.96 1.05 1.03 0.89 1.05 0.92 0.95 0.98 0.76 NOTE: X-MAXIMUM 1-PIN PEAK-1.49 l

t BAtmcRE CALVERT CLIFFS UNITI CYCLE 6 s *u r, l gas & Etsciatc co. ASSEMBLY RELATIVE POWER DENSITY AT EOC, Nuc r  ! nt EQUILIBRIUM XENON 5-3

0.77 1.04 l

0.84 1.12 1.11 0.82 1.04

.- X 0.75 1.05 0.96 1.21 0.98 l//b 0.75 0.82/ 0.76 1.23 0.89 1.23 0.81 I/b

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0.84 .l.05 0.77 0.94 0.91 1.24 0.92 1.11 1.12 0.96 1.24 0.93 1.23 1.14 1.04 1.01 1.12 1.22 0.90 1.24 1.13 1.26 1.05 0.99

~

, '0.77 0.82 0.98 1.23 0.92 1.04 1.04 1.21 0.87 1.04 1.04

/

' ///

7 0 0.81 1.11 1.01 0.99 0.87

'//

/

/0.

77/

l/// \ I//

l NOTE: X-MAXIMUM 1-PIN PEAK-1.59

/

'// CEA BANK 5

'/// LOCATIONS

^

GAS E E T IC Co' CALVERT CLIFFS UNITI CYCLE 6 Fisure coivert ci;rr, ASSEMBLY RELATIVE POVER DENSITY WITH BANK 5 Nuclear Power P! ant INSERTED, HFP, BOC 5-4

a 0.78 0.96 0.81 1.08 1.19 0.83 0.96 0.74 1.01 0.9E 1.20 0.99 I////

0.74 0.82 1.35 0.95 1.33 0.86

'//d x 0.81 ,,

1.01 0.82 0.99 0.94 1.19 0.94 1.11 1.07 0.97 1.35 0.95 1.14 1.06 1.02 1.00 s

1.18 1.19 0.95 1.19 1.05 1.15 1.02 0.99 0.78 0.83 0.99 1.33 0.94 1.01 1.02 1.26 0.89 0.96

/'///

/ 0.74/

/ //

0.96 0.86 1.11 1.00 0.99 0.89 0.44/

I/ /b {////

NOTE: X MAXIMUM 1-PIN PEAK 1.53

/

'// CEA BAN K 5 W/ LOCATIONS 8^ '* 0 R' CALVERT CLIFFS UNIT I CYCLE 6 Ficure gasc 'Ne t b;$ -

ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 INSERTED, HFP, EOC 5-5 Nuclect Power ?!cnt

6.0 THERMAL-HYDRAULIC DESIGN 6.1 DNER Analysis Steady state DNBR analyses of Cycle 6 at the rated power level of 2700 Kit have been performed using the TORC computer code described in Reference 1, the CE-1 critical heat flux correlation described in Reference 2, and the simplified modeling methods described in Reference 3 A variant of TCRC called CETOP, optimized for simplified modeling applications, was used in this cycle to develop the " design thermal margin codel" described generically in Reference 3 Details of CETOP are discussed in Reference 4; a similar discussion of CETOP methodology was submitted on the Arkansas Nuclear One Unit 2 (ANO-2) docket in Reference 5 and on the Calvert Cliffs docket in Reference 12. CETOP was approved for

~

use on ANO-2 in Reference 6. In general, this code differs from earlier versions of TORC only in that enthalpy transport coefficients are used to improve modeling of coolant conditions in the vicinity of the hot subchannel and in that more rapid equation-solving routines are used.

Direct comparisons show that CETOP models tend to be slightly more conservative than TORC design models in computing minimum CNBR for limiting

. cases. (Note that application of the methods of Reference 3 assures that design models set up with either TC'RC or CETOP are always conservative relative to detailed TORC analyses.) CETOP is used only because it reduces computer costs significantly; no margin gain is expected or taken credit for.

. Table 6-1 contains a list of pertinent thermal-hydraulic design parameters used for both safety analyses and for generating reacter protective system setpoint information. Also note that the calculational factors (engineering heat flux facter, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in Table 6-1 have been ccabined statistically with other uncertainty factors at the 95/95 confidence / probability level (Reference 7) to define a new design limit on CE-1 minimum CNBR when iterating on power as discussed in Reference 7 l

i Investigations have been made to ascertain the effect of the CEA guide tube l

l

28 wear problem and tha sleeving repair en ENER margins as cstablished by this '

typa of anclysis. 2a findings were report:d to tha NRC in Reference 8 which concluded that tha wear proble= and the sleeving repair do net l adversely affect DNBR =argin.

The thermal =argin model developed for Cycle 6 is not affected by the slightly higher fuel burnups experienced in Cycle 6. Be pcwer distributions used to develop the thermal margin model acccunt fcr fuel burnup explicitly, and have been found to be no mere limiting than power distributions used to develop thermal medels in previous cycles.

6.2 Effects of Fuel Rod Bowing on DNBR Margin The fuel red bewing effects on DNB margin for Calvert Cliffs -1 Cycle 6 have been evaluated according to the guidelines set fcrth in Reference 9 A total of 137 fuel assemblies will exceed the NRC specified DNB penalty threshold burnup of 24 GWD/T (Reference 9) during Cycle 6, the maxi =um assembly burnup rea$hing 42.8 GWD/T by the end of cycle. Fcr these assemblies which will experience a burnup of between 24 and 28 3 GWD/T at any time during Cycle 6, the mini =um best esti= ate margin available relative to more lirdting peaking values present in other asse=blies is greater than 10%. He DNB rod bow penalty for this burnup range, as determined via an inte polation of the data centained in Reference 9, varies from 0 to 1.4%. For assemblies which experience burnups in excess of 28 3 GWD/T, up to a maximum of 42.S GWD at ECC6 fer one_ assembly, the minimum best estimate margin available is considerably greater than 20%.

The ENB rod bow penalty for this latter burnup range, as determined via interpolation /extrpolatien of the data contained in Reference 9, varies from 1.4 to 6 3%. In su= mary, for both burnup ranges, the magnitude of the margin available is considerably in excess cf the correspcnding ENB red bcw penalty and, consequently, no pcwer penalty for fuel rod bewing is required in Cycle 6.

29

. .. TABLE 6-1 Calvert Cliffs Unit 1 Then11al-Hydraulic Parameters at Full Pcwer Reference General Characteristics Unit Cycle 5*_ Cycle 6**

Total Heat Output (core only) MWT 2700 2700 106 Btu /hr 9215 9215 Fraction of Heat Generated in .975 .975 Fuel Rod Primary System Pressure Nominal psia 2250 2250 Minimum in steady state psia 2200 Maximum in steady state psia 2300 Inlet Temperature 'F 550 548 Total Reactor Coolant Flow gpm 370,000 381,600 (steadystate) 106 lb/hr 139.0 143.8 Coolant Flow Through Core 106 lb/hr 133.9 138.5 Hydraulic Diameter ft 0.044 0.044 (nominal channel)

Average Mass Velocity 106 lb/hr-ft 2 2.51 2. 61 Pressure Drop Across Core psi 10.4 11.1 (minimum steady state flow irreversible op over entire fuel assembly)

Total Pressure Drop Across Vessel psi 32.4 34.4 (based on nominal dimensions and minimum steady state flow)

Core Average Heat Flux (accounts Stu/hr-ft 2 186,435*** 184,266*****

for above fraction of heat generated in fuel rod and axial densification factor)

Total Heat Transfer Area (Accounts ft 48,192*** 48,748*****

for axial densification factor)

Film Coefficient at Average Conditions Stu/hr-ft2_.F 5765 5930 Average Film Temperature Difference *F 32 31 Average Linear Heat Rate of kw/ft 6.23*** 6.16*****

Undensified Fuel Rod (accounts for above fraction of heat generated in fuel' rod)

Average Core Enthalpy Rise Btu /lb 68.8 66.5 Maximum C' lad Surface Temperature 'F 657 657

30 Table 6-1 (cont.)

1 Calculational Factors Reference Cycle 5 Cycle .6 Engineering Heat Flux Factor 1.03 1. 03 ****

Engineering Factor on Hot Channel Heat 1.02 1. 02 ****

Input Rod Pitch and Clad Diameter Factor 1.065 1. 06 5 ****

Fuel Dens 1fication Factor (axial) 1.'01 1.01 NOTES

$ Design inlet temperature and nominal primary system pressure were used to calc" late these parameters

• • Due to the statistical combination of uncertainties describad in References 7, 10 and 11, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.

• • • Based on a generic value of 1100 shims.

• • • • These factors have been combined statistically with other uncertainty factors at 95/95 confidence / probability level (Reference 7) to define a design limit on CE-1 minimum D?lBR when iterating on power as discussed in Reference 7.

• • • • • Based on Cycle 6 specific value of 672 shims.

e A

e 4

31' 7.0 TRANSIENT ANALYSIS This section presents the results of the Baltimora Gas & Electric Calvert Cliffs Unit 1. Cycle 6 Non-LOCA safety cnalysis at 2700 liWt.

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 Act.idents All DBEs listed in Table 7-1 were evaluated to determine the impact of the extended burnup of Cycle 6. DBEs for which a reanalysis was performed due to changes in methodology (given in References la, lb, lc and 2), changes

- in' key input parameters, and/or the effects of the extended burnup of Cycle 6 8

were explicitly included. For DBEs for which no key transient input parameter

, changed from the reference cycle value, justification is provided to show that the extended burnup of Cycle 6 has no adverse impact and that reference cycle analyses are valid for Unit 1, Cycle 6.

- 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.

t

exs I

TABLE 7-1 CALVERT CLI'FFS UNIT 1, CYCLE 6 DESIGN BASIS EVENTS CONSIDERED IN THE fl0N-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 Re-evaluated 7.1.2 Startup of an Inactive Reactor Coolant Not Reanalyzed 4 ,

Pump 7.1.3 Loss of Load Reanalyzed 7.1.4 Excess Load Reanalyzed 7.1.5 Loss of Feedwater Flow Re-evaluated 7.1.6 Excess Heat Removal due to Feedwater Re-evaluated fialfunction 7.1.7 Reactor Coolant System Depressurization Reanalyzed 7.2 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial

• steady state themal margin, maintained by the LCOs,are necessary to prevent exceeding the acceptable limits:

7.2.1 Sequential CEA Group Withdrawal l Reanalyzed 2

7.2.2 Loss of Coolant Flow Reanalyzed 7.2.3 Full Length CEA Drop Reanalyzed 7.2.4 Transients Resulting from the Reanalyzed Malfunction of One Steam Generator 3 7.2.5 Loss of AC Power 2 Re-evaluated 7.3 Postulated Accidents 7.3.1 CEA Ejection Reanalyzed 7.3.2 Steam Line Rupture Reanalyzed 7.3.3 Steam Generator Tube Rupture Re-evaluated 7.3.4 Seized Rotor 2 Reanalyzed 7.3.5 Fuel Misloading Reanalyzed 7.3.6 Fuel Handling Reanalyzed I

Requires High Power and Variable High Power Trip 2

Requires Low Flow Trip 3

Requires trip on high differential steam generator pressure Tech. Specs. preclude this event

33 TABLE 7-2 CALVERT CLIFFS UtlIT 1, CYCLE 6 CORE PARAt1ETERS It1PUT TO SAFETY AtlALYSES FOR DilB At40 CTM (CEllTERLItiE-T0 11ELT) DESIGri LIMITS Reference Cycle Cycle 6 Physics Paraneters Units Values (Cycle 5) Values Radial Peaking Factors ForDriBMarginAnalyses(F[)

~

Unrodded Region ~ 1.62 1.70+'*

Bank 5 Inserted 1.78 1.87+'*

For Planar Radial Component (Fly)'

of3-DPeak(CTriLimitAnalyses)

Unrodded Region 1.62 Bank 5 Inserted 1. 70++ '*

1.78 1.87 '

Maximum Augmentation Factor 1.055 1.055 Moderator Temperature Coefficient 10-4ao/*F -2.5**d.5 -2.5**++.5 Shutdown Margin (Value Assumed %ao -4.3 -5.3 -

in Limiting E0C Zero Power SLB)

Tilt Allowance  % 3.0 3.0 Safety Parameters Power Level MWt 2754 2700*

Maximum Steady State Core Inlet "F 550 548*

Temperature Minimum Steady State RCS Pressure psia 2200 2225*

6 l Reactor Coolant Flow 101bm/hr 133.9 138.5*

fiegative Axial Shape Index LCO Extreme ,Ip .16 .15*

Assumed at Full. Power (Ex-Cores) t on Maximum CEA Insertion at Full Power hI 25 25 Maximun Initial Linear Heat Rate 16.0 16.0

for Transient Other than LOCA KW/ft l Steady State Linear Heat Rate for KW/ft 21.0 21.3 l Fuel CRi Assumed in the Safety l Analyses CEA Drop Time from Removal of Power sec 3.1 3.1 to Holding Coils to 90% Insertion

< Minimun DriBR (CE-1) 1.195 1.23*

l

• For DilBR and CTM calculations, effects of uncertainties on these parameters were accounted for statistically. tiumerical values of these uncertainties and the procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to Drib and CTil l

limits are detailed in References la, lb, lc.

l **The effective initial MTC assumed for the SLB is -2.2x10-4ao/ F.

+The values assumed are conservative with respect to the-Technical Specification limits.

7.1 AllTICIPATED GPERATIOilAL CCCURREllCES FOR WHICH THE RPS ASSURES NO VIOLATI0ft OF LItilTS The events in this category were analyzed for Calvert Cliffs Unit 1 Cycle 6 to determine that acceptable limits on Df1BR, fuel centerline temperature to melt. (CTM), reactor coolant system (RCS) upset precure, and 10CFR100 site Laundary dose rate guidelines will not be exceeded. Protection against exceeding those limits is assured by the reactor protective system (RPS) limiting safety system settings (LSSS) setpoints. The setpoints incorporate the results cf the analyses of all the DBEs. The methodology used to generate the LSSS for the TM/LP and ASI RPS trips is contained in Reference la.

For those events in this section where DNBR and CTil values were calculated and quoted, the calculations were performed using the nominal values of key ilSSS parameters listed in Table 7-2. Uncertainties were accounted for in determining the values of DNBR and CT!i by applying appropriate values of aggregate uncertainties identified in Reference la to the limiting rod power. For those events analyzed to determine that the RCS upset pressure limit or 10CFR100 dose limits are not exceeded, the methods used are the same as previously reported in the FSAR and subsequent reload licensing submittals. Effects of NSSS parameter uncertainties on these linits are not assessed statistically. Instead, applicable uncertainties are assumed to occur simultaneously in the most adverse direction.

9

35 7.1.1 BORON DILUTION EVENT The Baron Dilution event was re-evaluated for Cycle 6 operation at extended burnup to verify that the conclusions reached for the reference cycle' (References 3 and 11) are applicable for Cycle 6.

Inadvertent baron dilution adds positive reactivity and produces temperature '

and power increases. The worst time in life for this event is at beginning of cycle (BOC) when boron concentration is highest and moderator temperature coefficient (MTC) is least negative. As noted in Table 5.1, the MTC at BOC is slightly more negative in Cycle 6 than for the reference cycle, and well within the more limiting MTC assumption employed in evaluating the baron dilution event. Therefore, increased burnup has no adverse effect on this transient. Since key transient parameters have not changed and since extended burnup has no impact on this_ event, the conclusions reached in the reference cycle analysis are valid for Cycle 6 operation.

m.

e*

i l

l a

n --

7.1.2 STARTUP OF All I?lACTIVE REACTOR C00LAfiT PU.MP EVEtiT The startup of an inactive reactor coolant pump event was not analyzed in Cycle 6 because the Technical Specifications do not pemit operation at power (modes 1 and 2) with less than 4 reactor coolant pumps operating.

I

7.1.3 LOSS OF LOAD EVENT -

37 Th3 Loss of Load event was reanalyzed for Cycle 6 to det:rmine that the transient DNBR does not exceed the design limit and that the RCS pressure upset limit of 2750 psia is not exceeded.

The methods used to analyze this event are consistent with those reported in the reference cycle (Reference 3), except that CETOP/CE-1 wa; used instead of TORC /CE-1 to calculate the DNBR.  !

The assumptions used to maximize RCS pressure during the transient are:

a) The event is assumed to result from the sudden closure of the. turbine stop valves without a simultaneous reactor trip. This assumption causes the greatest reduction in the rate of heat removal from the reactor coolant system aod thus results in the most rapid increase in primary pressure and the closest approach to the RCS pressure upset limit.

b) The steam cump and bypass system, the pressurizer spray system, and the power operated pressurizer relief valves are assumed not to be operable. This too maximizes the primary system pressure reached during the transient.

~

The Loss of Load event was initiated at the conditions shown in Table 7.1.3-1. The combination of parameters shown in Table 7.1.3-1 maximizes the calculated peak RCS pressure. As can be inferred from the table, the key parameters for this event are the initial primary and secondary pressures and the moderator and fuel temperature coefficients of reactivity.

The initial core average axial power distribution for this analysis was assumed to be a bottom peaked shape. This distribution is assumed because it minimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure and heat flux increases. The Moderator Temperature Coefficient (MTC) of +.5 x 10-4ap/*F was assumed in this analysis. This MTC in conjunction with the increasing coolant temperatures, maximizes the rate of change of heat flux and the pressure at the time of reactor tri p. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions was used in the analysis. This FTC causes the least amount of negative reactivity feedback to mitigste the transient increases l

in both the core heat flux and the pressure. The uncertainty on the l FTC used in the analyses is shcwn in Table 7.1.3-1. The lower limit on initial RCS pressure is used to maximize the rate of change of pressure, and thus peak pressure, following trip.

As seen from the above discussion, this event is limiting at BOC and hence, extended burnup has no impact. .

The Loss of Load event, initiated from the conditions given in Table 7.1.3-1, results in a high pressurizer pressure trip signal at 8.3 seconds.

At 11.5 seconds, the primary pressure reaches its maximum value of 2550.0 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 3.7 seconds. The secondary pressure reaches its maximum value of 1050.0 psia at 11.4 seconds after initiation of the event.

Table 7.1.3-2 presents the sequence of events fer this event. Figures 7.1.3-1 to 7.1.3-4 shcw the transient behavior of power, neat flux, RCS coolant temperatures, and RCS pressure.

l

l

. The event was also reanalyzed with the initial conditions listed in Table 7.1.3-3 co deter =ine'that the ac: ptable CNBR limit is not exceeded. T'ie minimum transient DNBR calculated for the event is 1.38 as compared to the design limit of 1.23.

The results of this analysis demonstrates tnat during a Loss of Load event the peak RCS pressure and the minimum CNBR do not exceed their respective design limits.

Since the Loss of Load event is limiting at BOC and the DNBR design limits are not exceeded, no fuel pins are predicted to fail and extended burnup has no adverse impact during Cycle 6 operation.

O 4

I

/

e

39 TABLE 7.1.3-1

, KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD ANALYSIS '

TO MAXIli!ZE CALCULATED RCS PEAK PRESSURE s- -

(.. .

Reference

• Units Cycle Cycle 6 Parameter ,

2754 2754 Initial Core Power Level MWt ,

Initial Core , Inlet Coolant *F , 550 '. . 550 Temperature - .

0 '

133.9 Core Coolant Flow ,

X10 lbm/hr 133.9 .

psia 2200 Initial Reactor Ccolant 2200 System Pressure ,

Initial Steam Generator psia 864.0 854.0 Pressure .

Moderator Temperature X10-4ap/*F ,

+.5 +.5 Coefficient ,

Doppler Ccefficient -

.85 .85

~

Hultiplier ,

%Ap -4.7 * -4,7

, CEA L' orth at Trip ,

Time to 90% Insertion of sec ., 3.1 3.1 Scram Rods . ,

Reactor Regulating System Opera' ting Mode Manual Manual Steam Dump and Bypass System Operating tiede Inoperative ,

Inoperative

~ ~

Cycle 5 (Reference 3) ,

~

3 SEQUENCE OF EVENTS FOR - -

-l THE LOSS OF LOAD EVE?h -

l

. TO MAXIMIZE CALCULATED RCS PEAK PRESSURE .

Time (sec) Event -

Setcoint or Value -

I

~ ~

0.0 Loss of Secondary Load , --

3.2 Steam Generator Safety Valves Open 1000 psia 8.3 High Pressurizer Pressure Trip '2422 psia Signal Generated --

e.

9.7 CEAs Begin to Drop Into Core --

9.8 Pressurizer Safety Valves Open 2500 psia 11.4 Maximum Steam Generator Pressure 1050 psia r .. .

11.5 Maxim m RCS Pressure . 2550 psia

k. . , -

13.4 Pressurizer Safety Valves are Fully -

2500 psia

. Closed . .

D O 9

. g e

e e

e a

en 9

e

TABLE 7.1.3-2 40 SEQUENCE OF EVEllTS FOR '

THE LOSS OF LOAD EVElli -

TO MAXIMIZE CALCULATED RCS PEAK PRESSURE C-Time (sec) Event ,

Setooint or Value 0.'O Loss of Secondary Load .

3 . 71 Steam Generator Safety Valves Open 1.000 psia 8.3 High Pressurizer Pressure Trip ~2422 psia Signal Generated .

9.7

~

CEAs Begin to Drop Into Core --

9.8 Pressurizer Safety Valves.Open 2500 psia

~

• 11.4 Maximum Steam Generator Pressure 1050 psia 11.5 Maximum RCS Pressure 2550 psia 13.4 Pressurizer Safety Valves are Fully -

2500 psia Closed .

e f g *

• a e e

e 6

g.

O e

C.

--> ,, nn _,

e

~

,- . KEY PAR &2TERS ASSUMED IN THE LOSS OF LOAD ANALYSIS TO . CALCULATE TRANSIENT MINIMUM DNBR .

.- ~ Reference

• Partmeter -

Units Cycle -

Cycle 6 Initial Core Pcwer Level >

MHt 2754 ,

2700

'F Initial Core Inlet Coolant ..

Temperature 550 ,

',.- 548 0 **

Core Coolant Flow ,

X10 lbm/hr 133.9 .

.138.5 Initial Reactor Ccolant psia -

~ **

System Pressur ..

, 2200

'~2225 Initial Steam Generator psia . . .

Pressure

.- 864.0 8.64.0 .

~.

Integrated Radial Peaking -

• *' +

Factors, FJ (Bank 5 inserted 25%)

1.71 1.75 Moderator Temperature 'X10 '4Ap/*F +.5 ,

+.5 Coefficient . .

Doppler Ccefficient ,. f. . .85 * .85 ,

Multiplier i .

CEA L'arth at Trip

%AP - ,

-4.7

~

. 44*7

~

Time to 907 Insertion of sec

~

. 3.1 . 3.1 Scram Rods ', ,- ,

Reactor Regulating System OperatingNode , Manual' Manual Steam Dump and Bypass System Operating Mode Inoperative ,

Inoperative

'~.

0 Cycle 5 (Reference 3)

• • Effects of uncertainties on these parameters were accounted for statistically.

(See Reference 1)

• The values assumed are conservative 'with respect to the Technical' Specifica:icn limits.

l l .

s o . . .

41 g

9 . **

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t. .. .

120 i i ,

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(- w e 40 i

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20 .

4 l

0 0 20 40 60 80 100 TIME, SECONDS ,

.e BALIN.CRE LOSS OF LOAD EVENT Ficure

~

GAS & ELECTRIC CO.

c.,i ,.,i c n ii. COR;t POW,cR vs u.,,, tnc 7.1.3-1 noa..n, %., ernni

42 120 ,' , , ,

J 100 E

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(. E 40 -

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tu w

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0 20 40 60 80 100 TIME, SECONDS

(' '

gas k^EUb$tc co. LOSS OF L0AD EVEllT Figure c,1-~i cit a. CORE AVERAGE HEAT FLUX vs TIME 7.1.3-2 i Flus le'n P ,w er P!, int j

43 _

(-

2700 , , ., i m .

m -

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a. 2500 -

w .

e . ... .

2 . ..

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t. o .. ,

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cc .

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f2 1900

, ,o ,

.w x -

1700 0 20 40 60 80 100 TIME, SECONDS

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i l ,.

l(

BA LilM OR': R" S

, GAS & ELECTRIC CO. LOSS OF LOAD EVENT CnNere cu'is REACTOR COOLANT SYSTEM PRESSURE vs TTME 7.1.5-3 Nuclear Pcwer Plant

. 44

~

630 , i i

• C -

vi -

T -

E o 610 OUT1.ET .

e w -

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TAVERAGE -

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a s./

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a T

INLET w ~ % ' ~ ~~ " - -

-c -

w< -

x .

530 '

y0 4O 0 0 100 TIME, SECONDS sALTucRE -

"F

^ LOSS OF LOAD EVENT l cI,t! FEE;).

REACTOR COOLANT SYSTEM TEMPERATURES vs TIME 7 1 3-4 c:o<.t...,, e.. c, n!,ini  ;

45 7.1.4 EXCESS LOAD EVE?lT The Excess Load Event was reanalyzed to detennine that the DNBR and CTM design limits are not exceeded during Cycle 6. The methods used to analyze this event are consistent with those report in the referenc cycle (Reference 3) except that CETOP/CE-1 was used instead of TORC /CE-1 to calculate the DNBR.

The analyses included the effects of manually tripping the RCP's on SIAS due to low pressurizer pressure and the automatic initiation of auxiliary feedwater flow on low steam generator level trip signal.

The High Power level and Thermal Margin / Low Pressure (TM/LP) trips provMe primary protection to . prevent exceeding the DNBR limit during this event.

Additional ' protection is provided by other trip signals including high rate of change of power, low steam generator water level, and low steam generator pressure. In this analysis, credit is taken only for the action of the High Power trip in the determination of the minimum transient DNBR.

~

The approach to the CTM limit is terminated by either the Axial Flux Offset trip, Variable High Power Level trip or the DNB related trip discussed above.

The most limiting load increase events at full power and at hot standby conditions, for approach to the DNBR limit of 1.23 (CE-1), are due to the completc opening of the steam dump and bypass valves.-

For conservatism in the analyses, auxiliary feedwater flow rate corresponding to 21% of full power main feedwate'r flow was assumed (i.e.,10.5% of full -

power main feedwater flow per generator). Also, the addition of the auxiliary feedwater to each steam generator was conservatively assumed to occur 180 seconds after reactor trip. The addition of the auxiliary feedwater flow to both steam generators results in anadditional cooldown of the RCS and a potential for a return-to-power (R-T-P) or criticality arising from reactivity feedback mechanisms.

The Excess Load event at full power was initiated at the conditions given in Table 7.1.4-1. The Moderator Temperature Coefficient (MTC) is the only key parameter which is adversely impacted by extended burnup.

4ap/ F TheanalysisconservativelyassumedanEOCMTCvalueof-2.5xlgp/F.

in comparison to the Technical Specification limit of -2.2x10 a Hence, the effects of extended burnup have been explicitly and conser-vatively included in the analysis.

A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions with an uncertainty of 15% was used in the analysis since this FTC causes the least amount of negative reactivity change for mitigating the transient increase in core heat flux. The minimum CEA worth assumed to be available for shutdown at the time of reactor trip for full power operation is 4.3%ao. The analysis conservatively assumed that the worth of boron injected from the safety injection tank is -1.00%ao per 105 PPM.

The pressurizer pressure control system was assumed to be inoperable because this minimizes the RCS pressure during the event and, therefore, reduces the calculated DNBR. All other control systems were assumed to be in manual mode of operation and have no impact on the results of this event.

The Full Power Excess Load event results in a High Power trip at 7.2 seconds.

l The minimum DNBR calculated for the event at the conditions specified in l Table 7.1.4-1 is 1.48 compared to the design limit of 1.23. The maximum '

local linear heat generation rate for the event is 18.1 KW/ft.

i

~

For the Excess Load event initiated frem HFP conditions, SIAS is. generated. .

at 34.3 seconds at which time the RCP's are manually trippec by the operator,. .The coastdown of the pumps decr:ases the rate of decay heat removal and, therefore, keeps the RCS coolant temperatures and pressure at higher values.

Auxiliary feedwater flow is delivered to both steam generators at 187.2 seconds. The feedwater ficw causes additional cooldown of the RCS. The decreas'ing temperatures in combination with a negative MTC inserts positive reactivity which enables the core to approach criticality.' The negativt, reactivity inserted due to the CEAs and Baron injected via the ligh Pressure Safety Injection (HPSI) pumps, however, is sufficient to maintain the core subcritical at all times.

Table 7.1.4-2 presents the sequence of events for an Excess Load event initiated at HFP conditior.s. Figures 7.1.4-1 to 7.1.4-5 show the NSSS response for power, heat flux, RCS temperatures, RCS pressure, and steam generator pressure during this event.

The Zero Power Excess Load event was initiated at the conditions given in Table 7.1.4-3. The MTC and FTC values assumed in the analysis are the same as for the full power case for the reasons previously given.

The minimum CEA shutdown worth available is conservatively assumed to be -4.0"ac. .

The results of the analysis show that a variable high power trip occurs at 35.9 seconds. The r.inimum DNBR calculated during the event is 2.92 and the peak linear heat generation rate is 14.4 K!!/ft.

As with the HFP Excess Load event, an SIAS signal on low pressurizer pressure is generated at 76.6 seconds for the 2ero Power Excess Load event. At 215.9 seconds auxiliary fee'dwater flow is delivered to both steam generators. The additional positive reactivity due to the cooldewn of the RCS is mitigated by the negative reactivity inserted due to CEA's and the baron injected via the HPSI pucps. The core remains suberitical at all times during an Excess Load event initiated from HIP conditions.

The sequence of events for the zerc power case is presented in Table 7.1.4-4 Figures 7.1.4-6 to 7.1.4-10 show the NSSS response for core power, core heat flux, RCS temperature, RCS pressure and steam generator pressure.

For the full and zero power Excess Load events initiated by a full opening l of the steam dur.p and bypass valves, the core remains subcritical even after astcmatic initiation of the auxiliary feedwater flow and following manual trip of the RCP's on SIAS due to low pressurizer pressure. The reactivity transient during a HFP and HZP Excess Load event is less limiting than tra corresponding Steam Line Rupture events (see Section 7.3.2).

In addition, the Excess Load event initiated from HFP ano HZP does not exceed the DNBR and CTit design limits. Since the analysis assumptions conservatively bound the Cycle 6 core parameter values and design limits are not exceeded, no fuel pins are predicted to fail and extended burnup has no adverse impact during this event.

TABLE 7.1.4-1 KEY PARAMETERS ASSUMED FOR FULL POWER EXCESS LOAD EVENT ANALYSIS Reference

• Parameter Units Cycle Cycle 6 Initial Core Power Level MWt 2754 2700 +

~

Core Inlet Temperature 'F 550 548 +

ReactorCoolankSystemPressure psia 2200 2225 +

0 Core Mass Flow Rate X10 lbm/hr 133.9 138.5+

Moderator Temperature Coefficient X10-4ao/*F -2.5 -2.5 CEA Worth Available at Trip %Ao -4.3 -4.3 Doppler Multiplier .85 .85 Inverse Baron Worth PPM /%Ao 105 105 Auxiliary Feedwater Flow Rate lbm/sec 175.0/S.G. 175.0/S. G.

~

High Power Level Trip Set' point  % of Full Power 112 110 Low S. G. Water Level Trip Setpoint ft. 30.9 30.9 RTD Response Time sec 8.0 12.0

• Reference cycle is Cycle 5, Reference 3.

+For DNBR calculations, effects of uncertainties on these parameters were combined statistically (see Reference 1) -

• e

+

l l

I l

. . w TABLE 7.1.4-2 . ,

. . . SEQUENCE OF EVENTS FOR THE E'(CESS LOAD EVENT AT FULL PCWER TO CALCULATE MINIMUM CNER Time (sec) Event .

Set:cint er Value 0.0 Complete Opening of Steam Dump and --

Bypass Valves at Full Pcwer ,

7.2 High Pcwer Trip Signal Ger.erated 110". of full pcwe9 7.6 Trip Breakers Open .

8.1 CEA's Begin to Drop Into Core .

8.6 Maximum Power; 113.2", of full pct Maximum Lncal Linear Heat Rate Occurs, KW/ft 18.1

-9.0 Minimum DNBR Cccurs 1.48 10.6 Low Steam Generator Level Trip Setpoint Reached 30.9 ft 34.1 Pressurizer Empties --

34.3 Safety Injection Actuation Signal Initiated; 1556 psia Manual Trip of RCP's 52.5 Main Steam Isolation Signal 548 psia 63.1 Rampdcwn of Main Feedwater Flow Ccmpleted 5". cf full power

' main feedwater fle 96.5 Pressurizer Begins to Refill --

132.5 Isolation of Main Feedwater Flew to Bnth --

S' team Generators 187.2 ' Auxiliary Feedwater Flow Delivered to Both 175.0 lbm/sec to Steam Generators each steam generator 600.0 Ocerator Terminates Auxiliary Feedwater ..

Flow to Both Steam Generators

49 TABLE 7.1.4-3 J

XEY PARAMETERS ASSUMED FOR HOT STAND 3Y EXCESS LOAD EVENT ANALYSIS Reference

• Parameter Units Cycle Cycle 6 Initial Core Power Level MWt 1 1+

Core Inlet Tcmperature 'F 532 - 532+

Reactor Coolant System Pressure psia 2200 2225+

0 Core Mass Flow Rate X10 lbm/hr. 137.1 141.35 +

Moderator Temperature X10'4ap/*F -2'. 5 -2.5 Ccefficient ,.

CEA Worth Available at Trip %ap -4.0 -4.0 Doppler Multiplier .85 .85 Inverse Boron Worth PPM /%Ao 100 100 Variable High Power Trip  % of full 40 40 Setpoint power Low S. G. Water Level Trip ft. 30.9 30.9 Setpoint

• Auxiliary Feedwater Flow .. lbm/sec 175.0/S.G. 175.0/S. G.

Rate RTD Response Time sec 8.0 12.0

• Reference Cycle is Cycle 5 in Reference 3.

l

+ For DNBR calculations, effects of uncertainties on these parameters were combined statistically (see Reference 1).

l l

l

w TABLE 7.1.4-.i

. . . SEQUENCE OF EVENTS FOR EXCESS LOAD EVENT AT FOT STANDSY CONDITIONS TO CALCULATE MINIMUM CNBR Time (sect Event Setooint or Value 0.0 Steam Dump and Bypass Valves Open to --

Maximum Flow Capacity

~

Variable High Power Trip Signal Generatad 40% of full powG 35.9 36.3 Trip Breakers Open --

36.0 CEAs Begin to Drop Into Core --

36.9 Core Power Reaches !!aximum; 40.4% of full power Maximum Local Linear Heat Rate 14.4 KW/ft 37.6 Minimum DNBR (CE-1)- 2.92 72.3 Pressurizer Empties --

76.6 Safety Injection Actuation Signal Generated; 1556 psia Manual Trip of RCS Coolant Pumps --

82.6 Main Steam Isolation Signal Generated 548 psia 88.7 Low Steam Generator Water Level Trip Setpoint 30.9 ft Reached -

~

106.8 Pressurizer Begins to Refill --

162.6 Isolation of Main Feedwatar Flew to Both --

Steam Generators 215.9 Auxiliary Feedwater Flow Delivered to 175.0 lbm/sec Both Steam Generators to each steam generator 600.0 Ocerator Terminates Auxiliary Feedwater --

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61 7.1.5 LOSS OF FEEDWATER FLOW The reference cycle analysis (Cycle 2, Reference 8) for the Loss of Feedwater Flow event was re-evaluated for Cycle 6 operation at extended burnup to verifiy that the RCS upset pressure limit of 2750 psia is not exceeded.

The Loss of Feedwater Flow event, like the Loss of Load event, is a heatup transient and is more limiting at BOC. Therefore, extended burnup has no adverse impact en this event.

The Loss of Feedwater Flow event results in a less severe pressure transient than the Loss of Load event. Hence, the conclusions reached for the loss of Load event (see Section 7.1.3) are applicable for the Loss of Feedwater Flow event.

4 e8 i

, . - , e - - . - , ,- ,-

9 7.1.6 EXCESS HEAT REMOVAL CUE TO FEECWATER MALFUtlCTICtl The reference cycle analysis (Cycle 2, Reference 8) for the Excess Heat Removal due to Feedwater Malfunction event was re-evaluated for Cycle 6 operation at extended burnup to verify that the DtlBR limit is not exceeded.

The Loss of High Pressure Feedwater Heaters is the most adverse feedwater malfunction event in terms of the cooling action on the RCS. This event like the Excess Load event is more limiting at EOC. This event has the same effect on the primary system as a small increase in turbine demand (approximately 9%) which is not matched by an increase in core power. As a result, the DriBR degradation associated with this event is less severe than for the Excess Load event where a larger effective increase in turbine demand (45") is analyzed.

The Excess Load event reanalysis (see Section 7.1.4), including exter.ded burnup effects, showed that the D!iBR limit is not exceeded. As stated previously, this event is less limiting than the Excess Load event. Hence, the conclusions reached for the Excess Load event are also applicable for this event.

e*

7.1.7 RCS DEPRESSURIZATION EVENT

-The RCS Depressurization event was reanalyzed for Cycle 6 to detemine the-pressure bias term input to the TM/LP trip. As stated in CENPD-199-P (Reference 4) the bias factor accounts for measurement system processing delays duri.ng RCS depressurization. The trip setpoints incorporating a bias factor at least this large will provide adequate protection to prevent the DNBR SAFDL from being exceeded during the transient.

The analytical method used in the reanalysis of this event is consistent with the reference cycle analysis (Rtfarence 12) except that CETOP/CE-1 was used instead of TORC /CE-1 to calculate DNBR.

The assumptions used to maximize the rate of pressure decrease and, conse-quently, the fastest approach to DNBR SAFDLs are:

1. The event is assumed to occur due to an inadvertent opening of both pressurizer relief valves while operating at rated themal power.

This results in a rapid drop in the RCS pressure and, consequently, a rapid decrease in DNBR.

2. The initial axial power shape and the corresponding scram worth versus insertion used in the analysis is a bottom peaked shape. This power distribution maximizes the time required to teminate the decrease in DNBR following a trip.

3. The charging pumps, the pressurizer heaters and the pressurizer backup heaters were assumed inoperable. This maximizes the rate of pressure decrease and, therefore, the rate of approach to DNBR SAFDL.

As seen from the hbove discussion, the key transient parameters for this event are independent of burnup and hence extended burnup has no impact on this event.

The analysis of this event shows that the pressure bias factor is 35 psia.

Hence, the use of a pressure bias tem which is as large as that detemined for the RCS Depressurization event will prevent exceeding the DNBR SAFDL.

4

7.2 ANTICIPATED OPERATIONAL OCCURRENCES WHICH ARE DEDEN0ENT ON INITIAL OVERPOWER MARGIN AND/OR RPS TRIPS FOR PROTECTION AGAINST VIOLATION OF LIMITS ,

The events in this category were analyzed for Calvert Cliffs Unit 1 Cycle 6 to determine the initial margin that must be maintained by the Technical Specification LCO limits such that acceptable DNBR, CTit, RCS upset pressure, and 10CFR100 site boundary dose -limits will not be exceeded during any of these events. The initial margin required to prevent the appropriate limits -

from being exceeded for any of these events was determined using the initial conditions specified in Table 7-2. For each event, conditions were chosen to assure that sufficient initial overpower margin is available at the initiation of the most limiting A00 in this category. The method of generating LCO is discussed in Reference lc.

As noted in Section.7.1, the initial conditions used in the evaluation of upset pressure limit and site boundary doses will differ from those given in Table 7-2 because the effects of NSSS parameter uncertainties on these limits are not combined statistically.

s

65 7.2.; CEA WITHDRAWAL EVENT The CEA Withdrawal event was reanalyzed for Cycle 6 to detennine the initial margins that must be maintained by the LCOs such that the DNBR and fuel centerline to melt (CTil) design limits will not be exceeded in conjunction with the RPS (Variable High Power Trip or Axial Flux Offset Trip).

3 As stated in CEN-121(B)-P, (Reference 2), the CEA Withdrawal event is e now classified as one for which the acceptable DNBR and centerline to melt limits are not violated by virtue of sufficient initial steady state thermal margin provided by the DNBR and Linear Heat Rate (LHR) related Limiting Conditions for Operations (LCOs). Depending on the initial conditions and the reactivity insertion rate associated with the CEA Withdrawal, the Variable High Power Level trip in conjunction with the initial steady. state LCOs, prevents DNBR limits from being exceeded. An approach to the CTM limit is terminated by either the Variable High Power Level Trip or the Axial Flux Offset Trip. The analysis only took credit for the Variable High Power Trip to determine the required initial overpower margin for DNBR.,using CETOP/CE-1.

The key parameters for the CEAW event initiated from both HFP and HZP are the reactivity insertion rate due to rod motion, the MTC and the FTC. The analysis conservatively assumed a positive MTC value since this in Lombina-tion with increasing doolant temperatures inserts a positive reactivity and thus maximizes the power and heat flux transients. The analysis also assumed a BOC FTC with a 15% uncertainty. This in combination with increasing fuel temperatures inserts the least amount of negative reactivity due to Doppler feedback. This minimizes the transient minimum DNBR. The range of reactivity insertion rates due to rod motion, along with the values of other parameters assuned in the analjsis is given in Table 7.2.1-1.

The only key parameters which are impacted by extended burnup are itTC and FTC. Since it is conservative to assume BOC values for these parameters, extended burnup has no adverse impact on this event.

The zero power case was analyzed to demonstrate that acceptable DNBR and centerline melt limits are not exceeded. For the zero power case, a reactor trip, initiated by the Variable High Power Trip at 40% (30% plus 10% uncertainty) of rated thennal power, was assumed in the analysis.

The zero power case initiated at the limiting conditions of operation results in a minimum CE-1 DNBR of 1.57. Also, the analysis shows that the fuel-centerline temperatures are well below those corresponding to the acceptable fuel centerline melt limit. The sequence of events for the zero power case is presented in Table 7.2.1-2. Figures 7.2.1-1 to 7.2.1-4 present the transient behavior of core power, core average heat flux, RCS coolant temperatures, and the RCS pressure for the zero power Case.

uu Protection against exceeding the DNBR limit for a CEA Withdrawal at full power is provided by the initial steady state thermal margin which is maintained by adhering to the Technical Specifications' LCOs on DNBR margin and by the response of the RPS which provides an automatic reactor trip on high power level. The minimum DNBR for this event, when initiated from the extremes of the LCOs, is 1.47. The analysis shows that the fuel centerline temperatures are well below those corresponding to the acceptable CTM limit. The sequence of events for the full power case is presented in Table 7.2.1-3. Figures 7.2.1-5' to 7.2.1-8 present the transient behavior of core power, core average heat flux, RCS coolant temperatures, and the RCS pressure for the full power case.

TheeventinitiatedfromtheTech. Spec.LCOs(inconjunctionwiththe Variable High Power Trip if required) will not lead to a DL3R or fuel temperature which exceed the DNBR and centerline to melt design limits.

Since the DNBR and CTM design limits are not exceeded for this event and no fuel pins are predicted to fail, it is concluded that extended burnup has no adverse impact during this event.

67 TABLE 7.2.1-1 KEY PARAMETERS ASSUMED IN THE CEA WITHDRAWAL ANALYSIS Reference Parameter Units Cycle

• Cycle 6 Range of Initial Core MWt 1 to 2754 1 - 2700**

Power Level Core Inlet Coolant *F 532 - 550 532 - 548**

Temperature Reactor Coolant System psia 2200 '

2225**

Pressure Moderator Temperature x10~4ap/ F +.5 +0.5 Coefficient Doppler Coefficient .85 .85 Multiplier CEA Worth at Trip - FP x10-2 ap -4.3 -4.7 CEA Worth at Trip - ZP x10-2 ap -4.0 -4.3 Range of Differential x10-4ap/in 0 to 3.2 0 to 3.2 Rod Worth CEA Group Withdawal in/ min 30 30 Rate Holding coil Delay sec 0.5 0.5 Time RTD Response Time sec 8.0 12.0

)

• Cycle 5 (Reference 3) o*For DNBR calculations effects of uncertainties on these parameters were combined statistically.

68 TABLE 7.2.1-2 SEQUEllCE OF EVEIITS FOR CEA WITHDRAWAL FROM ZERO POWER Time (sec) Event Setcoint or Vilue 0.0 CEA Withdrawal Causes Uncontrolled ---

Reactivity Insertion 26.4 Variable High Power Trip Signal Generated 40% of 2700 MWt 26.8 Reactor Trip Breakers Open ---

27.3 CEAs Begin to Drop Into Core ---

27.33 Maximum Core Power 124% of 2700 MWt 28.7 Miximum Heat Flux 54.9". of 2700 MWt 28.7 Minimum CE-1 DilBR l.57 32.1 Maximum' Pressurizer Pressure, 2385 psia

69 TABLE 7.2.1-3 SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM FULL POWER Time (sec) Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled ---

Reactivity Insertion 2.35 High Power Trip Signal Generated 110% of 2700 MWt 2.75 Reactor Trip Breakers Open ---

3.25 CEAs Begin to Drop Into Core ---

3.80 Maximum Core Power 115.3% of 2700 MWt 4.35 Maximum Heat Flux 106.5% of 2700 MWt 4.35 Minimum CE-1 DNBR 1.47

. 5. 6 Maximum Pressurizer Pressure, 2258 psia l

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re 7.2.2 LOSS dF COOLANT FLOW EVENT The Loss of Coolant Flow event was reanalyzed for Cycle 6 to determine the minimum initial margin that must be maintained by the Limiting Conditions for Operation (LCOs) such that in conjunction with the RPS (Low Flow Trip), the DNBR limit will not be exceeded.

The methods used to analyze this event are consistent with those discussed ir. Reference lc except CETOP/CE-1 was used instead of TORC /CE-1 to calculate DNBR.

The 4-pump Loss of Coolant Flow produces a rapid approach to the DNBR limit due to the rapid decrease in the core coolant flow. Protection against exceeding the DNBR limit for this transient is provided by the initial steady state thermal margin which is maintained by adhering to the Technical Specifications LCOs on DNB and by the response of the RPS which provides an automatic reactor trip on low reactor coolant ficw as measured by the steam generator differential pressure transmitters. -

The transient is characterized by the flow coastdown curve given in Figure 7.2.2-1. Table 7.2.2-1 lists the key transient parameters used in the present analysis. As seen from this table, the MTC and FTC are the only key parameters which are impacted by extended burnup. Since this transient is more limiting at BOC, corresponding MTC and FTC values were assumed in the analysis. Hence... extended burnup has no adverse impact on this cycnt.

Table 7.2.2-2 presents the NSSS and RPS responses during a four pump loss of flow initiated at a 0.0 shape index. The Low Flow trip setpoint is reached at 1.00 seconds and the scram rods start dropping into the core at 2.0 seconds. A minimum CE-1 DNBR of 1.23 is reached at 3.00 seconds.

Figures 7.2.2-2 to 7.2.2-5 present the core power, heat flux, RCS pressure, and core coolant temperatures as a function of time.

The event initiated from the Tech Spec LCOs in conjunction with the Low Flow Trip, will ensure that the DNBR design limits will not be exceeded.

Since DNBR design limits are not exceeded and no fuel pins are predicted to fail, extended burnup has no adverse impact during this event.

. - 79 TABLE 7.2.2-1 KEY PARAMETERS ASSUttED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter Units Reference Cycle

• Cycle 6, s

Initial Core Power Level MWt 2754 2700+

Initial Core Inlet Coolant *F 550 548+

Temperature 6

Initial Core Mass Flow Rate x101bm/hr 133.9 138.5+

Reactor Coolant System Pressure psia 2200 2225+

Modr-ator Temperature Coefficient x10-4t.p/*F +.5 +.5 Doppler Coefficient Multiplier -- 1.00 1.00**-

LFT Response Time sec .5 .5 <

CEA Floiding Coil Delay sec 0.5 0.5 CEA Time to 90% Insertion sec 3.1 3.1 (Including Holding coil Delay)

CEA horth at Trip (all rods out) x10-2Ap -5.60 -5.60 Unrodded Radial Peaking Factor 1.62 1.70+'++

(FT) 4-Pump RCS Flow Coastdown Figure 7.2.1-1 Figure of Reference 3 7.2.2 -1

• Cycle 5 (Reference 3) oo Since this is a second order effect and the most limiting Donnier nultiplier varies durir.g the transient, a nominal value is used.

+

For DNBR calculations, effects of uncertainties en these parameters were combined statistically.

The values assumed are conservative with respect to the Technical Specification linits.

80 TABLE 7.2.2-2 -

SEQUENCE OF EVENTS FOR LOSS OF FL0tf Time (sec) Event Setcoint or Value 0.0 Loss of Power to all Four Reactor ----

Coolant Pumps 1.00 Low Flcw Trip Signal Generated 93% of 4-Pump Flow 1.50 Trip Breakers Open ----

2.00 Shutdown CEAs Begin to Drop Into Core ----

3.00 Minimum CE-1 DNBR 1.23 5.70 Maximum RCS Pressure, psia 2308.0

1

) f0 8 '

1 4-PUMP C0ASTDOWN

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do 7.2.3 FULL LEdGTN CEA OROP EVElh

- The Full Length CEA Dr p cvent was reanaly:ed for Cycle 6 to determinc the initial thermal margins that must be maintained by the Limiting '

Conditions for Operatien (LCOs) such that tr.e DtGR and fuel centerline to '

melt design limit will not be exceeded.

The metheds used to analyze this event are censistent with those discussed in Reference 1 except CETOP/CE-1 'was used instead of TORC /CE-1 tn calculate DNSR. l Table 7.2.3-1 lists the key input parameters used for Cycle 6 and ccmpares l them to the reference cycle values. Conservative assumptions usad in the analysis include:

1. The most negative moderator and fuel temperature coeffi~cients of reactivity (including uncertainties), because these coefficients produce the minimum RCS ccolant temperature decrease upon return to -

100% power level and lead to the minimua DNBR.

2. Charging pumps and propertional heater systems are assumed to be inoperable during the transient. This maximizes the pressure drop during the event.

3. All other systems are assumed to be in manual mode of operation and .

have no impact en this event.

As'seen from the above d,iscussion, the MTC and FTC are the only key parameters.

Table 7.2.3-1 shows that the analysis which are impacted by extended burnup.

conservatively assumed an EOC MTC value of -2.5x10-4e/*F in comparison the to the Technical Specification limit of -2.2x10-4 e /*F. In addition, Hence, the effects analysis assumed an EOC FTC value with a 15% uncertainty.

of extended burnup have been explicitly and conservatively included in the analysis.

The event is initiated by dropping a full length CEA cver a pericd of .

1.0 second. The maximum increases in (integrated and planar) radial peaking factors in either redded or unredded planes were used in all axial regions of the core ence the power returns to the initial level. Values of 16t were assumed for these peak increases at full pcwer. The axial power shape in the hot channel is assumed to remain unenanged and hence the increase in the 3-D peak is propertional to the maximum increase in radial peaking factor of 16%. Since there is no trip assumed, and the secondary side conti-nues to demand 100% power, the peaks will stabilize at these asymptotic values l

~

after a few minutes.

I Table 7.2.2-2 presents the sequence of events fer the Full Length CIA Dr p The event initiated at the conditions described in Table 7.2.3-1.

transient behavicr of key NSSS parameters are presented in Figures 7.2.3-1

[

to 7.2.3-4. -

1 i The transient initiated at the most negative shape index LCD ( .19) and at the maximum power level alicwed by the LCO, results in a minimum l CE-1 DfGR of 1.23. A maximum allcwaole initial linear heat generation rate of 18.36 KW/ft could exist ss an initial condition without exceedina ~

21.3131/ft during this transient. This amoun of margin is asrured by j

setting the Linear Heat Ra a related LCOs based on the more limiting allowable linear heat rate for LOCA.

l

87 Consequently, it is concluded that the Full Length CEA Drop event initiated from the Technical Specification LCOs will not exceed the DNBR and centerline to melt design limits.

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- TABLE 7.2.3-1 XEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DRCP ANALYSIS Units Reference Cvele* Cvele 6 Parameter 2754 2700 +

Initial Core Power . Level MWt 550 548 +

Core Inlet Temperature ' F-

. Reactor Coolant System Pressure psia 2200 2225

• 6 133.9 138.'5 +

Core Mass Flow Rate X10 1bm/hr ,

Moderator Temperature Coefficient X10-4@/*F -2.5 -2.5 1.15 1.15 Doppler Coefficient Multiplier --

% Insertion of 25 25 Mikimum CEA Insertion at Allowed Power Bank 5

~

%@ unrodded .04 .04 Dropped CEA Worth .04 -

PDIL .04 ,

.Most Negative Axial Shape Index ,' -

Allowed at Full Power (LCO): ,

Ex-ccre .16 .15++

---

.19 Incore '

Unrodded Region 1.16 . 1.16 Integrated and Planar Radial 1.16 Peaking Distortion Factor Bank Inserted 1.16 . .

Region (Full Power)

~

• Cycle 5 (Reference 3) .

+ For DNBR calculations, effects of uncertainties on the3e parameters were ccmbined statistically. (See Reference 1)

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~ - TABLE 7.2.3-2 SEQUENCE OF EVENTS FOR CEA DROP Time (sec) Event Setpoint or Value 0.0 CEA Begins to Drop ----

CEA Fully Dropped -0.04%Ap 1.0 ,

1.1 Core Power Reaches Minimum 92.2% of 2700 MWt 4.2 Core Heat Flux Reaches Minimum 98.1% of 2700 ffWt 300. ' ~ ' Heat" Flux Reaches Final Value 100% of 2700 MWt 300. Core Inlet Temperature Reaches Minimum 546.5'F 300. RCS Pressure Reaches Minimum 2204.3 psia 300. Minimum DNBR Reached 1.23 e

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94 7.2. 4 A00'S RESULTING FRCM THE !!ALFUNCTICN OF ONE STEAM GEilERATOR The transients resulting frcm the malfunction of one steam generator were analyzed for Cycle 6 to determine the initial margins that must be maintained by the LCO's such that in conjunction with the RPS (Asymmetric Steam Ganerator Protective trip), the DNBR and fuel centerline melt design -

limits are not exceeded. ,

The methods used to analyze these events are consistent with those reported in the reference cycle ~ (Reference 3), except that CETOP/CE-1 was used instead of TORC /CE-1 to calculate the DNBR. - -

The four events which affect a single generator are identified below:

1. Loss of Load to One Steam Generator .- .

2. Excess Load to One Steam Generator

3. Loss of Feedwater to One Steam Generator

4. Excess Feedwater to One Steam Generator .

Of the four events described above, it has been determined that.the Loss of Load to One Steam Generator (LL/lSG) transient is the limiting .

asymmetric event. Hence, only the results of this transient are reported.

1 The event is initiated by the inadvertent closure of a single main' steam isolation valve. Upon the loss of load to the single steam generator, its pressure and temperature increase to the opening pressure of the secondary safety valves. The intact steam generator " picks up" the lost load, which causes its temperature and pressure to decrease. The cold leg asymmetry causes an inlet temperature tilt which results in an azimuthal power tilt, increased PLHGR and a degraded CNSR.

The LL/lSG was initiated at the conditions given ir$ Table 7.2.4-1. The MTC is the only key parameter which is adversely inpa:ted by extended burnup. The analysis conservatively assumed an EOC HTC value of -2.5x10-4ao/*F in comparison to the Technical Specification limit of -2.2x10-4ao/*F. Hence, the effects of extended burnup have been explicitly and conservatively included in the analysis.

l Table 7.2.4-2 presents the sequence of events for the Loss of Load to One Steam Generator. A reactor trip is generated by the Asymmetric Steam Generator Protection Trip at 2.6 seconds based on high differential pressure between the steam generators. The transient behavior of key NSSS parameters are presented in Figures 7.2.4-1 to 7.2.4-5.

A maximum allowable initial linear heat generation rate of 19.3 K./ft could exist as an initial condition without exceeding 21.3 KN/ft during this transient. This amount of margin is assured by setting the Linear Heat Rate LCO based on the mere limiting allowable linear heat rate for LOCA.

The minimum transient DNBR calculated for the LL/lSG event is 1.43 as compared to the minimum acceptable DNBR of 1.23.

~ _ _ . _ - . __ _ ~ _ _ _

95 The event initiated from the extremes of the LC0 in conjunction with the ASGP tri.p will not exceed the DitBR and centerline to melt design limits.

-Since the DilBR and CTM design limits are not exceeded and no fuel pins are predicted to fail, extended burnup has no adverse impact during this event.

g .-

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4

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.* m TABLE 7.2 A -1 KEY PARAMETERS ASSUMED IN THE ANALYSIS CF LOSS OF LOAD'TO ONE STEAli GENERATCR' Refer'ence

~

Units Cycle

• Cycle 6 Parameter 2754 2700 +

Initial Core Power MWt .

550 548 +

Initial Core Inlet *F -

Temperature psia 2200 2225+

Initial Reactor Coolant System Pressure ,

-2.5 -2.5 Moderator Temperature x10-4e/*F Coefficient

-- 0.85 0.85 Doppler Coefficient -

Hultiplier

}

Cycle 5 (Peference 3)

. + For DMBR calculations, effects of uncertainties on these parameters were combined statistically. (See Reference 1) e e

e O

97 s .

- TABLE 7.2.4-2 .

SEQUENCE OF EVENTS FOR ~

LOSS OF LOAD TO ONE STEAM GENERATOR Event Setpoint or Value Time (sec) 0.0 Spurious closure of a single main steam ----

isolation valve 0.0 . Steam flow from unaffected steam generator "

increases to maintain turbine power 2.6 ASGPT* setpoint reached (differential pressure) .

175 psid ,

3.2 Dump and Bypass valves are open . ----

3.5 ,

Trip breakers open .

4.0 CEAs begin to insert ----

4.0 . Safety valves open on isolated steam generator 1000 psia

~

Minimum tNSR occurs 1,43 5.5 I.'10.1 Maximum steam generator pressure 1050 psia-

.* ASGPT - Asymm.etric Steam Generator Protection Trip i

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103 7.2.5 LOSS OF ALL N0ti-EMERGEtiCY AC POWER EVEllT The Loss of All Non-Emergency AC Power event was re-evaluated for Cycle 6 operaticn at extended burnup to detemine that Dt1BR design limit is not exceeded and to verify that the site boundary doses are within those reported in the reference cycle (Cycle 2, Reference 8) analysis.

For the first few seconds of the transient, the Loss of All ? ton-Emergency AC Power behaves like a Loss of Flow event. Therefore, the transient minimum DilBR calculated for the Loss of Flow event (see Section 7.2.2) is applicable for this. event also.

The key parameters which detemine the site boundary doses are:

1. Primary and secondary coolant activities

2. Primary to secondary leak rate

3. Secondary steam flow

4. Steam generator partition factors Of these key parameters, only primary and secondary coolant activity, in principle, is burnup dependent. However, the reference cycle analysis assumed the maximum Technical Specification limits for the primary and secondary coolant activities (i.e.,1.0 uCi/gm primary coolant activity and 0.1 uCi/gm secondary coolant activity) and these Technical Specification limits are not going to be changed for Cycle 6 operation.

Since the Technical Specification limits on primary and secondary activities will remain at the reference cycle values and since none of the other key transient parameters have changed, it is concluded that the results and conclusions reported for the reference cycle are valid for Cycle 6 operation at extended burnup.

104 7.3 POSTULATED ACCIDENTS The events in this category were analyzed for Calvert Cliffs Unit 1, Cycle 6 to ensure acceptable consequences. For these transients, some amount of fuel failure is acceptable providEd the predicted site boundary dose does not exceed the 10CFR100 guidelines.

l I

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--- - - --r-.- - - - - - ,

7.3.l_ CEA EJECTION EVENT The CEA Ejection event was reanalyzed for Cycle 6 to determine the fraction. '

of fuel pins that experience.DNB during the event and show that, under the

' assumption that fuel which experiences DNB fails, the radiological

' consequences are within the guidelines of 10CFR100.

The analytical. method used in the reanalysis of this event is consistent l with the reference cycle analysis (Reference 5) except that CETOP/CE-1 with a DNB41imit of 1.23 was used instead of TORC /CE-1 to calculate DNER.

An ejected CEA (control element assembly) is assumed to occur as a result of either a complete circumferential break of the control element drive mechanism (CEDM) housing or the CEDM nozzle of the reactor vessel. The ejection of the CEA inserts positive reactivity into the core. The addition of positive reactivity causes a rapid rise in power and heat flux.

When the power excursion reaches the Variable High Power Trip (VHPT) setpoint, a reactor trip is initiated. The Doppler fuel temperature .

coefficient mitigates the power rise while the heat flux continues to increase. Insertion of negative reactivity due to scram rod motion

! causes the heat flux ta decrease. The decreasing heat flux terminates the Departure from Nucleate Boiling (DNB).

The key parameters used in'this event are listed in Table 7.3.1-1. With these key parameters selected to add conservatism, the procedure outlined in Reference 5 is then used to detemine the number of fuel pins which exceed the DNBR design limit. A conservative pin census distribution (a histogram of the number of pins in radial peak intervals of .05 nomalized' to the maximum pre- and post-ejected radiahpeaks)::isiused .

I to detemine the number of pins which experience DNB. Although C-E does not equate the onset of DNB with pin failure, the analysis conservatively assumed that all pins which experience DNB fail.

To bound the most adverse conditions during the cycle, the most limiting

! of either the Beginning of Cycle (BOC) or End of Cycle (E0C) parameter values were used in the analysis. A BOC Doppler defect was u ud since it produces the least amount of negative reactivity feedback to mitigate the

> transient. A t10C moderator temperature coefficient (MTC) of +.5x10-4ao/*F was used since a positive MTC, in conjunction with increasing ecolant temperatures, results in positive reactivity addition which enhances the power and heat flux excursion. An EOC delayed neutron fraction was used in the analysis to produce the highest power and heat flux rise during the event.

i To account for extended burnup considerations, a core average gap thermal conductivity value ccrresponding to a burnup of 50,000 MWD /MT was used.

This gap conductivity is conservative since it lowers fuel temperatures and thus results in less Doppler reactivity feedback to mitigate the rise in core power and heat flux during the event. This minimizes the transient minimum DNBR and thus maximizes the predicted number of fuel pin failures.

In addition, the site boundary dose calculations were performed assuming a radioisotope (i.e., I, Kr, and Xe) gap concentration corresponding to 50,000 MWD /MT. Hence, the analysis of this event conservatively included extended burnup effects.

i' The full power and zero power cases were analyzed assuming the value of

.05 seconds for the total ejection time which is consistent with the FSAR and previous reload license submittals.

_ _ _ _ _ _u _____

106 The full power case assumes the core is initially operating at 2700 MWt while the zero power ejection assumes an initial power level of 1 MWt.

A Variable High Power Trip is conservatively assumed to initiate at 110%

of 2700 MWt for the full power case and 40% (30% + 10% uncertainty) of 2700 MWt'for the zero power case to terminate the event.

The results of the full and zero power CEA Ejection analyses are presented in Table 7.3.1-2. The power trarsient produced by a CEA Ejection initiated at .the maximum allowed ~ power. level:is shown in Figure 7.3~.1-1.

Similar results for the zero power case are shown in Figure 7.3.1-2.

The number of fuel pins experiencing DNB is 11.0% for the ejection from full power and 6.3% for ejection from zero power. A maximum site boundary thyroid dose of 79.0 Rem and a Whole Body Dose (WBD)~ of .62 Rem occurs for the full power ejection.

The analysis of this event explicitly included the effects of extended burnup and since the resultant site boundary doses are within the 10CFR100 limits, it is concluded that the consequences of the CEA Ejection event are acceptable for Cycle 6 operation at extended burnup.

107 TABLE 7.3.1-1 KEY PARAMETERS ASSullED IN THE CEA EJECTION ANALYSIS Unit 1 Parameter Units Reference Cycle + Cycle 6 Full Power Core Power Level MWt 2754 2700*

Moderator Temperature x10-4ap/*F +.5 +.5 Coefficient Ejected CEA Worth %ap 0.22 0.22 Delayed Neutron Fraction .0044 .0044 Maximun Post Ejected 3.15 2.75 Radial Peak CEA Bank Worth at Trip %ao -3.0 -3.0 Doppler Multiplier .85 .85 High Power Trip Setpoint  % of rated power 112 110 Core Inlet Temperature 'F 550 548*

RCS Pressure psia 2200 2225*

Zero Power Core Power Level MWt 1 1*

Ejected CEA Worth %ao 0.63 0.63 Maximun Post Ejected 9.40 9.40 Radial Peak CEA Bank llorth at Trip %ao -1.50 -1.50 Doppler Multiplier .85 .85 Variable High Power  % of rated power 40 40 Trip Setpoint Core Inlet Temperature 'F 532 532*

RCS Pressure psia 2200 2225*

• For DNBR calculations, effects of uncertainties on these parameters were combined statistically.

,See Reference 5

1R@

TABLE 7.3.1-2 CEA EJECTI0tt RESULTS Unit

• Parameter Units Reference -Cycl e+ Cycle 6 Full Power Total flumber of Fuel Pins Which  % 10.7 11.0 Experience D?tB 0 - 2 Hr Site Boundary Thyroid Rem 76.8 79.0 Dose 0 - 2 Hr Site Boundary Whole Rem 0.25 0.62 Body Dose Zero Power Total flumber of Fuel Pins Which  % 6.6 6.3 Experience DilB 0 - 2 Hr Site Boundary Thyroid Rem 47.4 45.0 Dose 0 - 2 Hr Site Boundary Whole Rem 0.16 0.36 Body Dose

+See Reference 5

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111 i

7.3.2 Stean Line Rupture Event ,

i The Steam Line Rupture (SLB) event was analyzed for Cycle 6 to determine that the critical heat flux is not exceeded during this event.

i

[

The analysis included the effect of automatic initiation of auxiliary feedwater flow in three (3) minutes from the initiation of a Low S.G.

Pressure trip, a manual trip of the Reactor Coolant Pumps on Safety Injec- i tion Actuation Signal due to low pressurizer pressure, and an MSIV closure time of 12 seconds.

The analysis assumed that the event is initiated by a circumferential rupture of a 34-inch (inside diameter) steam line at the steam generator main steam line nozzle. This break size is the most limiting, since it causes the greatest rate of temperature reduction in the reactor core region.

! The SLB event.was analyzed with the' assumption of a three minute delay between the time of reactor trip on Low S.G. Pressure and the time when Auxiliary Feedwater (AFW) flow is delivered to the affected steam generator. This is conservative with respect to the expected time of AFW initiation since the _

generation of the AFW signal actually occurs at the time of the Low Steam Generator Water Level trip signal which occurs later than the Low S.G. Pressure trip. The analysis assumes, therefore, that AFW flow is delivered to tne steam generator sooner than the flow is actually available resulting in a conservative prediction of the resulting cooldown.

A conservatively high value of the AFW flow was calculated assuming that '

4 all auxiliary feedwater pumps are operable.. An AFW flow value of 21", of full power feedwater flow was used in the analysis. This value accounts l .for pump run-out due to reduced back pressure. In addition, the-analysis conservatively assumed that all the AFW flow is fed only to the damaged steam generator.

The analyses assumed that the main feedwater flow is ramped down to 5".

of full power feedwater flow in 20 seconds and that the main feedwater isolation valves are completely closed in 80 seconds after a low steam generator pressure or a main steam isolation signal. These assumptions are consistent with Technical Specification limits (see Table 3.3-5).

The manual trip of the RCPs is assumed to result in no flow mixing at the core inlet plenum. Thus, cold edge temperatures were used to calculate the moderator reactivity insertion during the cooldown of the RCS following an SLB.

The two SLB cases considered in cFjunction with automatic initiation of auxiliary feedwater flow and manual trip of RCPs are:

1. 2 Loop - Full Load (2754 MWt)

2. 2 Loop - No Load (1 MWt)

The 1 Loop - Full Load and 1 Loop - No Load cases were not analyzed since Technical Specifications prohibit operation in these modes.

l

The Two Loop-2754 MWt case was initiated at the conditions listed in Table 7.3.2-1. The Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis corresponds to end of life, since this MTC results in the greatest positive reactivity change during the RCS cooldown caused by the Steam Line Rupture. Since the reactivity change associated with moderator feedback varies significantly over the moderator density covered in the analysis, a curve of reactivity insertion versus density rather than a single value of MTC, is assumed in the analysis. The moderator cooldown curve assumed is given in Figure 7.3.2-1. The moderator cooldown curve given in Figure 7.3.2-1 was conservatively calculated assuming that on reactor scram, the Control Element Assembly is stuck in the fully withdrawn position which yields the most severe combination of scram worth and reactivity insertion.

The reactivity defect associated with the fuel temperature decrease is also based on an end of life Doppler defect. The Doppler. defect based on an end of life Fuel Temperature Coefficient (FTC), in conjunction with the decreasing fuel temperatures, casues the greatest positive reactivity insertion during the Steam Line Rupture event. The uncertainty on the FTC assumed in the analysis is given in Table 7.3.2-1. The a fraction assumed is the maximum absolute value including uncertainties for end of life conditions. This too is conservative since it maximizes subcritical multiplication and thus, enhances the potential for Return-To-Power (R-T-P).

Thekeyinputparametersimpactedbyextendedb$rnupareMTCandFTC. Even though no singular value of MTC is assumed in the analysis, the moderator cooldown as a function of moderator density is calculated assuming an initial MTC equal to the Technical Specification MTC limit of -2.2x10-ho/*F. In addition, an EOC FTC value in combination with a 15% uncertainty was assumed in the analysis. Hence, the analysis conservatively included the effects of extended burnup.

The minimum CEA worth assumed to be available for shutdown at the time of reactor trip at the maximum allowed power level is 7.02 ho. This available scram worth was calculated for the stuck rod which produced the moderator cooldown curve in Figure 7.3.2-1.

The analysis conservatively assumed that on Safety Injection Actuation Signal one High Pressure Safety Injection pump and one' Low Pressure Safety Injection pump fail to start. The analysis also assumed a conser-vatively low value of baron reactivity worth of -1.O ne per 95 PPM.

The conservative assumptions on feedwater flow were discussed previously.

The feedwater flow and enthalpy as a function of time are presented in Figures 7.3.2-2 and 7.3.2-3, respectively.

Table 7.3.2-2 presents the sequence of events for the full power case initiated at the conditions given in Table 7.3.2-1. The reactivity inser-tion as a function of time is presented in Figure 7.3.2-4. The resconse of the flSSS during this event is given in Figures 7.3.2-5 through 7.3.2-9.

The results of the analyses show that STAS is actuated at 15.3 seconds, at which time the Reactor Coolant Pumps n nanually tripped by the operator.

The nanual trip of RCPs slows down the rate of primary heat removal and thus delays the time when the affected steam generator blows dry. The affected steam generator blows dry at 97.8 seconds and teminates the s lewn @f th@ RCS. Th@ geak reactivity attained prior to delivery of

113 l

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of 8.8%, consisting of 3.'6T decay heat and 5.2% fission power is produced at 110.8 seconds. The continued production of decay heat from the fuel I after temination of blowdown, causes the reactor coolant temperatures to increase. This in turn reduces the magnitude of the positive moderator l reactivity inserted and thus the total reactivity becomes more negative.

The delivery of auxiliary feedwater flow to the affected steam generator 4

tat 183.9 seconds initiates a further cooldown of the RCS which results i in more positive reactivity insertion. The positive reactivity insertion j causes the. core to return to criticality. The peak criticality attained

! is +0.056%ap at 449.2 seconds. The reactivity transient is terminated by

{ the boron injected via the High Pressure Safety Injection Pumps. A peak R-T-P of 6.0%, consisting of 2.8% decay heat and 3.2% fission power is produced at 581.2 seconds.

The MacBeth correlation (Reference 6) with the Lee non-uniform mixing correlation factor (Reference 7) results in a post-trip minimum DNBR of 1.31' compared i to the DNBR limit of 1.30 during a SLB event initiated from hot full power conditions. Thus, critical heat fluxes are not exceeded during this event. _

t I Two Loop-No Load case was initiated at the conditions given in Table 7.3.2-3.

The moderator cooldown curve is given in Figure 7.3.2-10. The cooldown curve corresponds to an end of life MT3. An and of life FTC was also used ,

for the reasons previously discussed in connection with the two loop-2754 MWt case.

I The minimum CEA worth assumed to be available for shutdown at the time of rentor trip at the zero power level is 5.3%ao. This available scram worth  !

was calculated for the stuck rod which produced the moderator cooldown curve in Figure 7.3.2-10. A maximum inverse baron worth of 90 PPM /%ao was conser-vatively assumed for the safety injection during the no load case. The feed-water flow and m enthalpy used in the analysis are presented in Figures 7.3.2-11 and 7.2.J-12, respectively.

. Table 7.3.2-4 presents the sequence of events for the Two Loop-No Load case initiated from the conditions given in Table 7.3.2-3. The reactivity inser-tion as a function of time is presented in Figure 7.3.2-13. The NSSS response during this event are given in Figures 7.3.2-14 to 7.3.2-18.

The results of the analysis show that SIAS is actuated at 12.9 seconds, at which time the RCPs are manually tripped by the operator. Auxiliary feed-  ;

water flow is . initiated at 184.2 seconds which continues the cooldown of the RCS. Thus, the total cote reactivity approaches criticality. The peak reactivity attained is .243%ap at 291.7 seconds and a peak power of 2.8% occurs at 319.6 seconds. The addition of baron via High Pressure ,

Safety Injection mitigates the reactivity transient.

The MacBeth correlation with the Lee non-uniform mixing correlation factor results in a post-trip minimum DNBR of 1.39 compared to the DNBR limit of 1.30 during a SLB event initiated from hot zero power conditions. Thus, critical heat fluxes are not exceeded during this event.

The Steam Line Rupture event initiated from HFP and HZP conditions with automatic initiation of auxiliary feedwater flow and manual trip of RCPs on SIAS due to low pressurizer pressure shows that the DNBR limits are not exceeded. Since the DNBR limits are not exceeded and no fuel pins are predicted to fail, it is concluded that the consequences of the SLB event rmfgggstablg for Cycle 6 operation at extended burnup.--

LO90 TABLE 7.3.2-1 KEY PARAMETERS ASSUMED IN THE STEAM LINE RUPTURE ANALYSIS 2-LOOP-2754 MWT Reference

_fParameter Units Cycle

• Cycle 6 Initial Core Power MWt 2754 2754 Initial Core Inlet *F 550 550 Temperature Initial RCS Pressure psia 2300 2300 Initial Steam Generator psia 853 871 Pressure Low Steam Generator psia 548 548 Pressure Analysis Trip Sstpoint Safety Injection Actuation psia 1556 1556 Signal

• Minimun CEA Worth Available b -7.15 -7.02 at Trip Doppler Multiplier 1.15 1.15 3

Moderator Cooldown Curve b vs density See Figure 7.3.2-1 ofR re c 11 Inverse Boron Worth PPM / % 105 95 Effective MTC x10-4e/*F -2.2 -2.2 s Fraction (including .0060 .0060 uncertainty)

• Cycle 5 (Reference 11) i

115 TABLE 7.3.2-2 SEQUEllCE OF EVEllT3 FOR STEAM LINE RUPTURE EVEllT WITH AUTOMATIC IflITIATIO!! 0F AUXILIARY FEEDWATER AND MANUAL TRIP OF REACTOR COOLANT PUMP 2-LOOP-2754 MWT Time (sec) Event Setpoint or Value 0.0 Steam Line Rupture Occurs ---

2.5 Low Steam Generator Pressure Trip 548.0 psia Signal Occurs 3.4 Main Steam Isolation Valves Begin to 548.0 psia Close 3.4 Trip Breakers Open ---

3.9 CEAs Begin to Drop Into Core ---

14.2 Pressurizer Empties ---

15.3 Safety [njection Actuation Signal 1556.0 psia 15.3 Reactor Coolant Pumps Manually 1556.0 psia Tripped 15.4 Main Steam Isolation Valves ---

Completely Closed 23.4 Main Feedwater Rampdown 5% of Full Power Feedwater Flow 45.3 High Pressure Safety Injection Pumps ---

Start 83.4 Main Feedwater Isolation ---

97.8 Affected Steam Generator Blows Dry ---

109.3 Peak Reactivity, Prior to Auxiliary -0.13 %

Feedwater Flow 110.8 Peak Return to Power 8.8% of 2700 MWt 183.9 Auxiliary Feedwater Flow Initiated 350 lbm/sec to Ruptured Steam Generator 449.2 Peak Reactivity Post Auxiliary +0.056 h Feedwater Flow 581.2 Peak Return to Power Post Auxiliary 6.0% of 2700 MWt Feedwater Flow 600.0 Operator Isolates Ruptured Steam ---

Steam Generator and Terminates Auxiliary

' TABLE 7.3.2-3 XEi PARNIETERS ASSU!!ED IN THE STEAM LIllE RUPTURE AftALYSIS 2-LOOP fl0 LOAD Reference Parameter Units Cycle

• Cycle 6 Initial Core Power Mut 1.0 1.0 Initial Core Inlet *F 532 532 Temperature Initial RCS Pressure psia 2300 2300 Initial Steam Generator psia 899 900 Pressure Low Steam Generator Pressure psia 548 548 Analysis Trip Setpoint Safety Injection Actuation psia 1556 1555 Signal Minimum CEA Worth Available h -4.3 -5.3 at Trip Doppler Multiplier 1.15 1.15 Moderator Cooldown Curve  % vs density See Figure 7.3.2-10 See Figure of Reference 11' 7.3.2,10 Inverse Boron Worth PPM / b 100 90 Effective MTC x10-4e/*F -2.2 -2.2 8 Fraction (. including .0060 .0060 uncertainty)

• Cycle 5 (Reference 11)

• 117 TABLE 7.3.2-4 SEQUENCE OF EVEliTS FOR STEAM LIf4E RUPTURE EVE!iT WITH AUTOMATIC INITIATI0ft OF AUXILIARY FEEDWATER AND MANUAL TRIP OF REACTOR COOLANT PUMP 2-LOOP-N0 LOAD Time (sec) Event Setpoint or Value .

0.0 Steam Line Rupture Occurs ---

2.8 Low Steam Generator Pressure Trip 548.0 psia Signal Occurs 1

3.7 Main Steam Isolation Valves 8egin 548.0 psia to Close 3.7 Trip Breakers Open ---

4.2 CEAs Begin to Drop Into Core --- -

~~ ~ ~~~

10.6 Pressurizer Empties ---

12.9 SafethInjec.ionActuationSignal 1556.0 psia 12.9 Reactor Coolant Pumps Manually 1556.0 psia Tripped 15.7 Main Steam Isolation Valves ---

Completely Closed

~

42.9 High Pressure Safety Injection ---

Pumps Start 84.2 Main Feedwater Isolation ---

184.2 Auxiliary Feedwater Flow Initiated 350 lba/sec to Ruptured Stean Generator 291.7 Peak Reactivity .243 %

319.6 Peak Power 2.8% of 2700 Mwt 600.0 Operator Isolates Ruptured Steam ---

Generator and Terminates Auxiliary Feedwater Flow

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136 )

l 7.3.3 STEAM GENERATOR TUBE RUPTURE EVENT The Steam Generator Tube Rupture (SGTR) event is re-evaluated for Cycle 6 operation at extended burnup to verify that the site boundary doses are within those reported in the reference cycle (Reference 11) analysis.

The key parameters which detennine the site boundary doses are:

1. Primary and secondary coolant activities

2. Primary to secondary leak rate

3. Secondary steam flow

4. Steam generator and condenser air ejection partition factors Of these key parameters, only primary and secondary ccolant activity is, in principle, burnup dependent. The reference cycle analysis assumed the maximum Technical Specification limits for the primary and secondary coolant activities (i.e.,1.0 pCi/gm primary coolant activity and 0.1 uCi/gm secondary coolant activity) and these Technical Specification limits are not going to be changed for Cycle 6 operation.

Since the Technical Specification limits on primary and secondary activities will remain at the reference cycle values and since none of other key transient parar.eters have changed, it is concluded that the results and conclusions reported for the reference cycle are valid for Cycle 6 operation -

at extended burnup.

137 7.3.4 SEIZED ROTOR EVEtiT The Seized Rotor event was reanalyzed for Cycle 6 to demonstrate that the RCS upset pressure limit of 2750 psia will not be exceeded and that only a small fraction of fuel pins are predicted to fail during this event. In addition, this event was' analyzed to show that the site boundary doses are'within.the limits of 10CFR100.

The methods used to analyze this event are consistent with the reference cycle (Reference 3) analysis except that. CETOP/CE-1 with a D?iBR limit of 1.23 was used instead of TORC /CE-1 to calculate DtlBR.

The single reactor coolant pump shaft seizure is postulated to occur as a consequence of a mechanical failure. The single reactor coolant pump shaft seizure results in a. rapid reduction in the reactor coolant flow to the three-pump value. A reactor trip for the Seized Rotor event is initiated by a low coolant flow rate as detennined by a reduction in the sum of the steam generator hot to cold leg pressure drops. This signal is compcred to a . setpoint which is a function of the initial number of operating reactor coolant pumps. For this event, a trip will be initiated when the flow rate drops to 93". of the initial flow. .

The initial conditions and key parameters assumed for the Seized Rotor event are listed in Table 7.3.4-1. As seen from this table, the analysis con-

^ servatively assumed a BOC MTC value of +0.5x10-4.o/*F.t This MTC, in conj,unc-tion with increasing coolant temperatures, inserts positive reactivity and thus enhances the power and heat flux excursion. The analysis also assumed a BOC FTC with a'15% uncertainty. The BOC FTC,in conjunction with increasing fuel temperatures, inserts the least amount of negative reactivity due to Doppler feedback. These assumptions minimize the transient minimum DtlBR and maximize the predicted number of fuel pin failures.

A conservatively " flat" pin census distribution (a histogram of the number of pins versus radial peak in intervals of 0.01 normalized l

to the maximum peak) is used to determine the number of pins that experience l DtlB. The site boundary dose calculations were perfonned assuming a radia-l isotope gap concentration corresponding to 50,000 MWD /MT.

i The key transient parameters which are impacted by extended burnup are MTC, FTC and radioisotope gap concentrations. As explained previously, this transient assumes BOC values for MTC and FTC and, consequently, extended burnup has no adverse impact on these parameters. In addition, the radioisotope gap concen-i trations assumed in calculating site boundary doses correspond to a burnup

! of 50,000 MWD /MT. Hence, extended burnup effects have been conservatively included in the analysis of this event.

i Table 7.3.4-2 presents the flSSS and RPS response for a Seized Rotor

event initiated from an axial shape index value of .24. The pressurizer pressure reached a maximum value of 2313 psia at 3.50 seconds. Figures 7.3.4-1 through 7.3.4,4 give the response of the core power, core average heat flux, RCS pressure, and coolant temperatures during the transient.

138 The results of the analysis show that no more than 3.0% of fuel pins are predicted to experience DNS. The resultant site boundary doses, calculated using the assumptions given in Table 7.3.4-3, are; 0-2 Hr Thyroid Dose = 3.60 REli (DEQ I-131) 0-2 Hr Whole Body Dose = 0.4 REM (DEQ Xe-133)

For the case of the loss of coolant flow resulting frem a seizure of a reactor coolant pump shaft, the Low Flow Trip in conjunction with the DNB LCOs,' limits the predicted number of oin failures to only a small fraction of

.the total pins in the core. The resultant site boundary doses are well within

he 10CFR100 limits. In addition, the maximum RCS pressure experienced during the event is well below the upset pressure limit of 2750 psia.

The analysis of this event explicitly included the effects of extended burnup. Since the site boundary doses are within 10CFR100 limits, it is concluded that the consequences of the Seized Rotor event are acceptable for Cycle 6 operation at extended burnup.

sr

• f39 TABLE 7.3.4-1 KEY PARAMETERS ASSUMED IN SEIZED ROTOR ANALYSIS Refdrence Unit 1 Units Cycle

• Cycle 6 Parameter Initial Core Power Level MWt 2754 2700**

Core Inlet Coolant Temperature ' F- 550 548**

6 138.5**

4-Pump Core Mass Flow Rate 10 1bm/hr 133.9 6 106.8**

3-Pump Core Mass Flow Rate 101bm/hr 103.4 Reactor Coolant System Pressure psia 2200 2225**

Moderator Temperature Coefficient X10~4ap/*F +.5 +.5 Doppler Coefficient Multiplier -- .85 .85 CEA Worth at Trip ,

%Ao -5.6 .

-5.6 1.70**,+

Integrated Radial Peaking 1.62 Factor with Tilt; F4

~*24****

Axial Shape Index .16

• Cycle 5 (Reference 3)

• • Uncertainties on these parameters were combined ' ca ut"cally.

+ The values assumed are conservative with resp n t to t.e Technical Specification limits.

k .

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TABLE 7.3.4-2 SEQUENCE OF EVENTS FOR SEIZED ROTOR Time (Sec) Event Setroint or Value 0.0 Seizure o# One Reactor Coolant Pump 93% of Initial 4-Pump 0.0 Low Flow Trip Setpoint Reached Flow -

0.50 Trip Breakers Open 1.00 CEAs Begin Dropping into Core --

3.50 Maximum RCS Pressure 2313 psia .

O e

I

. e 141 TABLE 7.3.4-3 ASSUMPTI0tlS FOR SEIZED ROTOR DOSE CALCULATI0ft

1. During first 1800 see there is no operator action.

Tius, the steam releases are to the condenser via the bypass valves and the atmosphere via the atmospheric dump valves.

2. Beyond 1800 seconds steam dump is to condenser via operator action.

3. Primary activity released to secondary assumed to be released without mixing within the S.G. secondary liquid.

4. Primary to secondary leak rate

• 1 GPM

5. Initial Deactor Coolant System Maximum 1

• 0 uCi Allowable Tech Spec Concentration (DEQ I-131)* gm

6. Initial Steam Generator Maximum Allowable 0.1 uti Concentration (DEQ I-131)* ' gm

7. Reactor Coolant System Maximum Allowable 100/E Concentration of tioble Gases (DEQ Xe-133)*

8. Partition Factor for Primary Release Through 1.0 Atmospheric Dumps

9. Partition Factor for Secondary Steam Release 0.1 Through Atmospheric Dump

10. Feed Pump Turbine Air Ejector Partition .0005 Factor 3

11. Atmospheric Dispersion Coefficient 1.8x10-4sec/m

12. Breathing Rate 3.45x10-4 3m /sec 6

13. Dose Conversion Factor (I-131) 1.486x10 REti/Ci

14. All Activity in the Gap is Assumed to be Released to the Coolant

15. Calculated Activity in the Gap Corresponds to a Burnup of 50,000 MWD /MT

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^ SEIZED ROTOR EVENT Figure GAS E i C O.

cai.,, r e a n h REACTOR COOLANT SYSTEM TEMPERATURES vs TIME 7.3.4-4 Nuclear Power P!act - -

146 735 Analysis of Fusi Misicading Evant Method of Analysis The analysis of the fuel misloading event was performed in two steps. He first step was to determine which fuel loading errors are detectable. This was done by calculating symetric dual CEA red worths for a variety of misloaded cores at hot zero power, beginning of cycle conditions and by applying a criterien for detectability to these results. The criterion fer detectability was chosen as a 3 5c difference between the worth of any dual CEA and the average worth of all dual CEAs within a given CEA bank. This criterion is sufficiently low to allow for detection of major fuel loading errors and sufficiently high to preclude false indication of fuel loading errors due to measurement uncertainties and ner=al power tilts.

The second step of the analysis was the evaluation of the consequences of normal operation with an undetectable fbel loading error. This was done by calculating the incTeases in maximum 1-pin peaks at hot full power throughout the ' operating cycle for the misloaded cores and comparing these increases to the initial steady state thermal margin maintained by the limiting conditions for cperation.

Results Table 7 3 5-1 sumarizes the cypes of misloading events fcund to be detectable and undetectable for Calvert Cliffs 1 Cycle 6. Table 7 3 5-1 lists the detectability of similar misloading events analyzed for the reference cycle (Reference 9). Leading errors listed as detectable would be identified during the hot zero pcwer testing and 'culd be ccrrected prior to power generation. Mcwever, interchange or rotation of assemblies shewn as undetectable wculd not produce CEA worth asymetries in excess of the 3 5 c detectability criterien and, therefore, would be undetectable during hot zero pcwer testing. The most adverse loading errer(s) within each type of undetectable error was analy:ed to deter =ine the consequences of ner=al operation with such an unidentified error.

i Table 7 3 5-2 displays the calculated pin peaking facter increases at hot full power conditiens for the undetectable misicadings shewn in Table

full power conditions for ths undstect:ble mislordings showi in Tablo 7 3 5-1 for c:lvtrt Cliffs Unit 1 cycle 6. Included in Tablo 7 3 5-2 is corresponding informatun from the reference cycle. Information showT1 in Table 7 3 5-2 represents the most adverse misleading effects throughout Cycle 6.

Table 7 3 5-2 indicates that the most adverse undetectable loading error is the interchange of a fresh shimmed assembly with a once-burned assembly.

' Pin peaking effects for the fresh shimed assembly include the result of

^ the burnout of the poison shims throughout the cycle. The maximum radial pin peaking factor associated with this misloading results in a calculated radial peaking factor which is 10% above the Technical Specification limit including appropriate uncertainties. This increase in radial peaking above the Technical Specification limit does not cause the fuel safety limits to be exceeded because of the initial steady state thermal margin maintained by the limiting conditions for operation. These limiting conditions for operation provide 17% margin on DNB and 35% margin I

on peak linear heat generation rate.

e

148 TABLE 7.3.5-1 Detectability of Fuel Misicading Events Calvert Cliffs I Fuel Type Misloaded With Reference Cycle (I) Cycle 6 Fresh Any other assembly Detectable Detectable Unshimed Fresh Once-burned fuel Undetectable Undetectable Shimed Twice-burned fuel Detectable Detectable Thrice-burned fuel -

Detectable Once-Burned ( Once-burned fuel Undetectable Undetectable Twice-burned fuel Detectable Undetectable

.. Thrice-burned fuel Undetectable All Rotation Undetectable Undetectable U) Reference 9 .

(21 Results with regard to detectability for the intercha.nge of two twice-burned fuel assemblies are the same as indicated for once-burned fuel; results for the interchange of twice- and thrice-burned fuel are the same as for the interchange of once- and twice-burned fuel.

l l

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_ . . . . . . . _-. _=

.. . 149 TABLE 7.3.5-2 Maximum Pin Peaking Factor Increases Calvert Cliffs' Unit I Cycle 6 Calculated Pin Peaking Factor Increase at HFP, %(I)

Within Misloaded (4)

(21 (3) or Potentially Type of Misloading For Entire Core Relative to Tech Spec Limiting Assemblies Interchange of a 4% +10% 14"

_ fresh shimmed (Ref. cycle: 3%) (Ref. cycle:l% above) (Ref. Cycle: 13%)

ex a eb Interchange of Less than 1% -24% 7%

two assemblies within an exposure (Ref. cycle: (Ref. cycle: more (Ref. cycle,. 3". )

i batch less than 2%) than 5% below)

Interchange of a once-burned assem- 3j g +4% 13%

bly with a twice-burnedassembly(5)

Interchange of a once-burned assem- Less than 1% -24'." 8%

bly with a thrice-t burned assembly 4% -12% 5%

Hisrotation of an Assembly (Ref. cycle: 3%) (Ref. cycle:4".below) (Ref. cycle: 3%)

t (1) Applicable information from the reference cycle (Reference 9) is included in parenthesis, i (2) Percent increase in the maximum pin power in the core during the cycle for the misloaded core as compared to the maximum pin power in the core during the cycle for the normally loaded core.

I. (3) Percent difference between the maximum pin peaking factor in the core during the cycle for the misloaded core and the Technical Specification limit.

(4) Maximum increase in the peak pin within an assembly as compared to the peak pin within the same assembly in the normally loaded core at the same time in the cycle.

I (5) Results for the interchange of twice- and thrice-burned fuel are less severe.

l

. 150 736 Fuel Handling Event The consequences of a fuel handling incident due to increasing the burnup to 45,0C0 M'nT/T has been investigated for Calvert Cliffs fuel. The results of this investigation deconstrate that the present results shewn in Reference 10 will not be changed by increasing the burnup.

The dose rate at the site boundary will not increase because the gas gap inventory will be less than the gas gap inventory shown in Reference

10. The Reference 10 gas gap activity is based en the hottest fuel assembly in the core, independent of time (burnup) during the fuel cycle.

Since the radioactive fission products which centribute significantly to the dose rate at the site boundary reach maximum concentrat:.sas at relatively low burnt.ps, the only significant influence of burnup is the increased release from the fuel pellet for a given fuel temperature, comonly called " enhance =ent". The fuel temperature, in turn, is prinicpally dependent on linear heat rate.

The predicted linear heat rate for Calvert Cliffs nit 1, Cycle 6 fuel rods has been calculated to determine the radioactive fission product release to the gas gap. The maximum linear heat rate fer rods with burnups between 33,000 M'nT/T and 45,000 MnT/T is abut 7.5 kw/ft; this maximum occurs at 33,000 Kat/T and the linear heat rate decreases to less than 6.0 kw/ft for the maximum fuel red exposure at ECC6. The maxi =um fuel temperature is not high enough to have significant diffusion-type release from the fuel; the method of release will be primarily from knock-out or recoil. Ccnsequently, the radioactive fission product release to the gas gap will be less than 1% of the inventory; this releae is based en the ANS 5.4 Standard, "Methed fer Calculating the Fractional Release of Velatile Fissicn Products from Oxide Fuel." The release of less than 1% is over a facter of ten lower than that assu=ed in the fuel handling accident in Reference 10.

'

• 151 8.0 ECCS PERFORMANCE ANALYSIS 8.1 Introduction and Summary An ECCS performance analysis was performed for Calvert -Cliffs Unit I Cycle 6 to demonstrate compliance with 10CFR50.46 which presents the NRC Acceptance Critga for Emergency Core Cooling Systems for Light-Water-Cooled Reactors . The analysis justifies an allowable peak linear heat generation rate (PLHGR) of 15.5 kw/ft which is the same as the Cycle 5 limit. The method of analysis and detailed results which support this value are presented herein.

8.2 Method of Analysis The NRC approved C-E large break evaluation model (2) was used for this analysis. The model was used to re-evaluate ECCS performance for the

~

limiting large break LOCA. reflood hydraulic calculations employed in the The blowdown Cycle 5 evaluation and refg apply to Cycle 6 since there are no changes to RCS hardware characteristics. Therefore, only the hot rod clad temperature and oxidation calculations were performed to evaluatT3}he fuel rod conditions specific to Cycle 6. .The NRC approved STRIKIN-II code was used for this. purpose.

• Burnup dependent calculations were performed to determine the limiting conditionfortheECg)performanceanalysis. The nuclear and fuel thermal performance (FATES 3 ) data used as input to the ECCS analysis considered high burnup effects specific to the Cycle 6 reload. Two STRIKIN-II temperature calculations were performed: one at.the burnup with the maximum initial fuel stored energy and one at high burnup in order to verify that the low burnup maximum fuel stored energy case is limiting.

The temperature calculations for both cases were performed for the 1.0 DgPD* break. The break spectrum analysis performed for Unit I Cycle

. 2 determined that the 1.0 DES /PD is the limiting break for high density fuel since this break yields the highest residual fuel stored energy at the end of blowdown as well as the highest peak clad temperature.

The PARCH (5) code was not utilized in the Cycle 6 evaluation. Use of the steam cooling heat transfer coefficients calculated by PARCH would improve late reflood heat transfer and reduce the peak clad temperature reported herein.

, *0ES/PD - Double-Ended Slot at Pump Discharge l

~

152 8.3 Results Table 8-1 presents the results calculated for the most limiting burnup for Cycle 6 and the results for a case at high burnup which resulted in blowdown rupture. A list of the significant parameters displayed graphically for the limiting case is presented in Table 8-2. A summary of the fuel parameters is shown in Table 8-3.

The results of the evaluation confirm that 15.5 kw/ft is an acceptable value for the PLHGR in Cycle 6. As shown in Table 8-1, the peak clad temperature and maximum local and core wide clad oxidation values for the limiting case are below the 10CFR50.46 acceptance criteria limits.

As shown in Table 8-1, the case analyzed at the burnup with the maximum initial stored energy in the fuel (3000 MWD /MTU) resulted in the highest peak clad temperature. An initial fuel average temperature of 2213*F and-fuel pin internal pressure of 1251 psia were calculated for this burnup assuming operation at 15.5 kw/ft. The transient results presented in Figures 8-1 to 8-6 are very similar to those predicted for Cycle 5. The fuel cladding is predicted to rupture as well as achieve its peak temperature during the reflood period (at 32 seconds and 249 seconds, respecti vely).

The case analyzed at high burnup (34,000 MWD /MTU) resulted in a peak clad temperature of 2024*F,14*F lower than the maximum initial fuel stored energy case. An initial fuel average temperature of 2127'F and fuel pin internal pressure of 2191 psia were calculated for this case assuming operation at 15.5 kw/ft. (This is very conservative since a rod at such a 4

high burnup would actually be at a power below 15.5 kw/ft.) At 15.5 kw/ft the pin pressure was sufficient to cause rupture during the blowdown period. However, the initial stored energy at that burnup is not sufficient to cause a higher peak clad temperature than the case at 3,000 MWD /MTU, as shown in Table 8-1.

8.4 Conclusion As discussed above, conformance to the ECCS criteria is sumarized by the analysis results presented in Table 8-1. The most limiting case results in a peak clad temperature of 2038'F, which is well below the acceptance limit of 2200*F. The maximum local and core wide values for zirconium oxidation percentages, as shown in Table 8-1, remain well below the acceptance limit values of 17% and 1%, respectively. Therefore, operation of Unit I Cycle 6 at a plHGR of 15.5 kw/ft and a power level of 2754 MWT (102% of 2700 MWT) results in compliance with the 10CFR50.46 acceptance criteria.

TABLE 8-1

SUMMARY

OF ECCS PERFORMANCE RESULTS FOR CALVERT CLIFFS I CYCLE 6 FOR Tile LIMITING BREAK SIZE (1.0 DES /PD)

PEAK CLAD TIME OF PEAK TIME OF CLAD OXIDATION CASE BURNUP TEMPERATURE TEMPERATURE CLAD RUPTURE LOCAL CORE-WI DE Limiting Case (Maxinom Initial- MWD Fuel Stored Energy) 3,000 MTU 2038 F 249. Sec 31.9 Sec 8.5% 4.51%

liigh Burnup Case MWD 2024*F 248. Sec 10.6 Sec 8.4% 4.51%

(Blowdown Rupture) 34,000IITif G

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TABLE 8-2 CALVERT CLIFFS I CYCLE 6 ANALYSIS PLOTS FOR LIMITING CASE Variable Figure Number Peak Clad Temperature 8-1 Hot Spot Gap Conductance 8-2 Peak Local Clad Oxidation 8-3 Temperature of Fuel Centerline, Fuel Average, Clad and Coolant at Hottest Node 84 Hot Spot Heat Transfer Coefficient 8-5 Hot Rod Internal Gas Pressure 8-6 1

_. .. . _ = .

- - 155 TABLE 8-3 CALVERT CLIFFS I CYCLE 6 FUEL PARAMETERS Quantity Value Reactor Power Level (102% of Nominal) 2754 MWT Average Linear Heat Rate (102% of Nominal) 6.44 KW/FT Peak Linear Heat Generation Rate (PLHGR) ~

Hot Assembly, Hot Channel 15.5 KW/FT Peak Linear Heat Generation Rate (PLHGR)

Hot Assembly, Average Channel 13.14 KW/FT 2

• Gap Conductance at PLHGR 2025 BTU /HR-FT *F

• Fuel Centerline Temperature at PLHGR 3634 'F

• Fuel Average Temperature at PLHGR 2213 'F

• Hot Rod Gas Pressure 1251 psia

• Hot Rod Burnup 3000 MWD /MTU i

i

• Fuel rod parameters given are those which yield the limiting ECCS performance results.

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. . 159 FIGURE 8-4 CALVERT CLIFFS I CYCLE 6

'l.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG TEMPERATURE OF FUEL CENTERLINE, FUEL AVERAGE, CLAD AND COOLANT AT HOTTEST N0DE 35CC i 3000, -

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1/b2 9 0 TECHNICAL SPECIFICATICNS 2e Technical Specification changes which must be made in order to make the Calve'rt Cliffs I Technical Specifications valid fer the operation of Cycle 6 are presented in this section. Table 9-1 presents a stm=ary cf the Technical Specification changes and Table 9-2 presents the explanatiens fer the changes stenarized in Table 9-1. Following Tables 9-1 and 9-2, for each Technical Specification which must be modified, either the existing page with the intended modificaticn or the already modified page with a new figure is provided.

e

163 Tablo 9-1 Calvert Cliffs I Cyclo 6 Technical Specification Changes Change Tech Spec # Action the following Tech Spec changes were originally requested in Reference 1, and are unchanged from those requested in Reference 1.

1 Figure 2.1-1, Replace Figure 2.1-1 with enclosed page 2-2 Figure 2.1-1 2 Figure 2.2-1 Replace Figure 2.2-1 with enclosed page 2-11 Figure 2.2-1 3 B.2.1.1 Change LHGR to centerline melt limit page B2-1 from 21 kw/ft to 213 kw/ft.

4 B.2.1.1, B.2.2.1 Change minimum DNBR value from 1.195 pages B2-1, B2-3 to-1.23 as indicated on noted pages B2-5, B2-6 5 B.2.1.1, B.2.2.1 Change high power level trip and maximum pages B2-1, B2-4 high power level trip actuation from 112% of rated thermal power to 110%

6 B 3/4.2.5, Change minimum DNBR of 1.195 to minimum B 3/4 2-2 'page DNBR of 1.23 The following Tech Spec changes were originally requested in Reference 1.

However, these changes have been revised from those requested in Reference 1.

7 Figure 3 2-2 Replace Figure 3 2-2 with enclosed page 3/4 2-4 Figure 3 2-2 T

8 3/4.2.2.1 from Pages 3/4 2-6, Changetocalculated 41.620 value 4 1.650 and F o{ >FE620to 3/42-7 31.650,chliingeFigure33-3to l Figure 3 2-3a and change Tech Spec numbers frca 3/4.2.2 to 3/4.2.2.1 I -

9 323 Change calculated value of ' from page 3/4 2-9 61.62g to 41.650, change r > 1.620F {p to Fr71.650, insert action item (b) and change Figure 3 2-3 to 3 2-3a The following Tech Spec changes have been added to the list of changes requested in Reference 1.

10 B.2.2.1 Revise description of TM/LP trip, and page B2-7 change allowance frca 92 psia to 40 psia 1

164 Tabla 9-1 (centinued) .

Change Tech Scec # Action 11 3 1.1.1 20cCF.,

page 3/4 1-1 Change shutdown margin, T[E,>:

from 14 3% 3k/k to 15 3%

12 Figure 3.1.2 Replace Figure 3 1.2 with enclosed page y 4 1-27 Figure 3 1.2 13 4.2.1 3 Change Figure 3 2 3 to Figure 3 2-3b 0 se 3/4 2-2 14 3/4.2.2.2 Insert Pages 3/4 2-7a and 3/4 2-7b Pages 3/4 2-7a, after Page 3/4 2-7 3/4 2-7b 15 Figure 3 2-3 Replace Figure 3 2-3 with page 3/4 2-8 enclosed Figure 3 2-3a 16 Figure 3 2-3b Insert Figure 3 2-3b after Page F 4 2-8 Page 3/4 2-8a 17 Figure 3 2-4 Replace Figure 3 2-4 with page 3/4 2-11 enciesed Figure 3 2-4 18 3 2.6 .. Add " Core Pcwer" to item (d) page F 4 2-14 19 Table 3 2-1 Replace Table 3 2-1 with page y 4 2-15 enclosed Table 3 2-1 20 Table 3 3-2 Replace Table 3 3-2 with page 3/4 3-6 enciesed Table 3 3-2 21 Table 3 3-5 Change MSIV respense time page y 4 3-21 from 6 9 to 12 9 seconds 22 4.7.1 5 Increase the MSIV test closure time page 3/4 7-9 from 3 6 seconds to 4.0 seconds 23 B 3/4 1.1.1 and Change EOC shutdown =argin, T -

B 3/4 1.1.2 frcm >,u.3%.sk/k to >5 3% A k/k, avg >' 200*F page B 3/4 1-1 and change ECC shuidewn require =ent top;4.5% 1 k/k 24 531 Re=cve li=it en enrich =ent in page 5-4 description of reload fuel asse=blies l

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165 Ttblo 9-2 Explanations for Cycle 6 Tech Spea Changes Change Tech Scec # Explanation-1 Figure 2.1-1 Thermal Limit Lines have been changed to reflect higher radial peaking factors and implementation of margin recovery programs.

2 Figure 2.2-1 The G R LSSS has been changed to reflect higher radial peaking factors and the implementation of margin recovery programs 3 B.2.1.1 GGR to centerline melt is being raised based upon Cycle 6 analyses 4 B.2.1.1, The minimum DNBR has been increased to B.2.2.1 1.23 to be censistent with Statistical Combination of Uncertainties 5 B.2.1.1, Statistical Cembination of Uncertainties B.2.2.1 . has removed the 2% power uncertainty from the transient analyses 6 B 3/4.2.5 The minimum DNBR has been increased to be consistent with Tech Spec B.2.1.1 7 Figure 3 2-2 The GR LCO is being changed as a result of higher radial peaks, the implemen-tation of margin recovery programs, and the addition of Figure 3 2-3b to take crepitforthecalculatedvalueof when monitoring the G R LCO F 3 w1Ehtheexceredetectorsystem. This change (# 7) and Change Nos. 8,13,14,15 and 16 are made to avoid unnecessary i power level changes resulting from temp-l orary on-line computer outages 8, 14 3/4.2.2.1, Theplanarradialpeakingfactor, 3/4.2.2.2 Fx , is being raised for Cycle 6, anEtheuseofFigure32-3isbeing l

l expandedtotakeegeditforthecalcu-when monitoring lated the G Rvalue of F*Ehe excore detector LCO with system 9 323 The integrated radial peaking factor, F f, is being raised for Chcle6,andBASSSisbeingimplemented i

Tabla 9-2 (centinued)

Change Tech Scec # Explanation 10 B.2.2.1 The W LP basis'has been adjusted to be consistent with Statistical Cembination of Uncertainties (SCU) and the bias has been changed as a result of the imple-mentation of SCU and the recatager-ization of CEX4 (A' conservatve bias value relative to the Transient Analysis results has been incorpcrated) 11 3 1.1.1 The shutdown margin has been increased to yield acceptable results for the ECC HZP SLB event 12 Figure 3 1.2 The PDIL is being changed to yield accep-table results for the ECC HZP SLB event and the allowable BASSS cperating region is beir.g indicated 13 4.2.1 3 The maxi =um allowable fraction of RATED THERMAL PO'4ER is being changed to take crept for the calculated value of Fx when monitoring the LHR LCO -

w1!htheexcoredetectorsystem 14 (See Change 8) 15 Figure 3 2-3 Radial _ceaking facters, both F xyT and Fr *, are being raised for Cycle 6, and new Fx 1 li=its for excoremonitoringoftheLHRLCOare

, being added 16 Figure 3 2-3b Figure 3 2-3b is being added to take creptferthecalculatedvalueof Fx when monitoring the LHR LCO w1Ehtheexcoredetectorsystem.

17 Figure 3 2-4 The DNB LCO has been changed to reflect higher radial peaking facters and the implementatien of margin recovery programs.

18 3 2.6 The AXIAL SHAPE INDEX, Core Power will be

, maintained within the limits shown on l Table 3 2-1 19 Table 3 2-1 The ASI limits have been =cdified fcr the. implementation of SASSS 20 Table 3 3-2 The RTD delay time used in the Cycle 6 analysis has been increased from 8 to 12 seconds

167 Table 9-2 (continued)

Change Tech Spec # Explanatien 21 Table 3 3-5 The MSIV closure time used in the Cycle 6 analysis has been increased frem 6 to 12 seconds 22 4.7 1.5 The MSIV test closure time is being increased as a result of the increase in MSIV closure time 23 B 3/ 4 1.1.1 and The shutdown margin has been increased B 3/4 1.1.2 to make it consistent with Specification 3 1.1.1 24 531 The specification of the enrichment limit in reload fuel assemblies has been removed to streamline the description and to permit the use of higher enrich-ment fuel S

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. i of RATED TH. ERMAL POWER . .

2-11 Ac.endment No. U , 24 CALVERT CLIFFS - UNIT l

' ' 197

11.0 REFERENCES

References (Chapters 1 through 5)

1. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), "Fifth Cycle License Application," dated September 22, 1980

2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Supplement 1 to Fifth Cycle License Application," dated November 4,1980.

3 Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Report of Startup Testing for Cycle Five," dated April 8,1981.

4. Letter, A. E. Lundvall, Jr. (BG&E) to R. W. Reid (NRC), " Fourth Cycle License Application," dated February 23, 1979

5. BG&E Calvert Cliffs 1 Slides Depicting SCOUT-1 High Burnup Demonstration P ogram, presented $t a BG&E/C-E/NRC Meeting in Bethesda, Md. on December 20, 1978.

6. Letter, A. E. Lundvall, Jr. (BG&E) to B. C. Rusche (NRC), "Second Cycle License Application," dated October 1, 1976. .

7 Letter, A. E. Lundvall, Jr. (BG&E) to R. W.Reid, " Unit 2 Cycle 2 License Application," dated July 26, 1978.

8. CENPD-187, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding,"

dated June 1975 9 CEN-182(B)-P, " Statistical Approach to Analyzing Creep Collapse of Oval Fuel Rod Cladding Using CEPAN," dated September 1981.

10. CEN-183(B)-P, " Application of CENPD-198 to Zircalloy Ccmponent timensional Changes," dated Septemer, 1981.

t l

l

11. CEl-83(B)-P, "Calv;rt Cliffs Unit 1 Rcccter Operatien with Modified CEA Guide Tubes," dated February 8, 1973, and Letter, A. E. Lundvall, Jr.

(BG&E) to V. Stello, Jr. (NRC), "Rsacter Operation with Mcdified CEA Guide Tubes," dated February 17, 1978.

12. CalPD-139, "C-E Fuel Evaluation Model Tcpical Report," dated July,1974.

13 CEN-161(B)-P, " Improvements to Fuel Evaluation Model," dated July, 1981.

14. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Reanalysis of CEA Ejection Event Against New Criteria," dated May 29, 1981.

15 CENPD-153-P, Revision 1, " Evaluation of ' Uncertainty in the Nuclear Pcwer Peaking Measured by the Self-Pcwered Fixed In-Core Detecter System," dated May 1980.

i

! 16. A. Jonssen, et. al.,, " Discrete Integral Transport Theory Extended to the Case with Surface Sources," Atomkermenergie, Bd_. 24_, 1974.

17 A. Jensson, et. al., " Verification of a Fuel Assembly Spectrum Ccde Based on Integral Theory," Trans. g. Nucl. Sec., 288 (778), 1978.

18. TIS-6363, " Core Physics Validation for the Ccebustion Engineering PWR,"

dated November 12, 1979 19 Letter enclosing Amendment 24 frem Robert A. Clark (NRC) to Wi.rliam Cavanaugh, III, (AP&L), " Operation of ANO-2 Curing Cycle 2," June 19, 1981.

20. CDi-133(B), " FIESTA, A Cne-Dimensional, Two Grcup Space-Time Kinetics cede for Calculating PWR SCRAM Reactivities," dated Neve=ter,1979

199 R7ferenccs (Chacter 6)

1. CEPD-161-P, "TCRC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core", July 1975

2. CEWD-162-P-A (Proprietary) and CEWD-162-A (Nonproprietary),

" Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1, Uniform Axial Power Distribution", April 1975

3. CENPD-206-P, " TORC Code, Verification and Simplified Modeling Methods", January 1977

4. Letter,A.E.Lundvall,Jr.(BG&E) tor.A. Clark (NRC),"Responseto Questions on SCU, CEN-124(B)," dated June 2,1981.

5. Letter, D. C. Trimble (AP&L) to Director, NRR, "CETOP-D Code Structure and Modeling Methods, Response to First Round Questions on the Statistical Combination of Uncertainties Program (CEN-139(A)-P)",

July 15, 1981

6. Final Safety Evaluation Report Supporting Facility Operating License Amendment No. 26 on Docket No.30-368 and Operation of ANO-2 During Cycle 2, July 21, 1981

7. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 2",

January 1980

8. CEN-83(8)-P, "Calvert Cliffs Unit 1 Reactor Operation With Macified CEA Guide Tubes", February 8, 1978 and letter, A. E. Lundvall, Jr. to V. Stello, Jr., " Reactor Operation With Modified CEA Guide Tubes",

February 17, 1978

9. Letter, D. F. Ross and D. G. Eisenhut-(NRC) to D. B. Vassallo and K.

R. Goller (NRC), " Revised Interim Safety Evaluation Report on the l Effects of Fuel Rod Bowing in Thermal Margin Calculation for Light l Water Reactors", February 16, 1977

10. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 1",

December, 1979.

l 11. CEN-124(B)-P, " Statistical Contination of Uncertainties, Part 3",

( March 1980 l 12. CEN-124(B)-P Part 2,"" Response to First Round Questions on the Statistical Combination of Uncertainties Program: CETOP-D Code Structure and Modeling Methods," May 1981.

l

200 References for Chanter 7.0 -

la. " Statistical Combination of Uncertainties tiethodology; Part 1; C-E Calculated Local Power Density and Thermal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," CEN-124(B)-P, December 1979.

lb. " Statistical Cembination of Uncertainties !!ethodology; Part 2:

Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units I and II," CEN-124(B)-P, January 1980.

ic. " Statistical Combination of Uncertainties itethodology; Part 3; C-E

. Calculated Local Power Density and Departure from Hucleate Boiling Limiting Conditions for Operation for Calvert Cliffs Units I and II," e CEN-124(B)-P, March 1980.

! 2. " Method of Analyzing Sequential Control Element Assembly Group Withdrawal Event for Analog Protected Systems," CEN-121(B)-P, November 1979.

3. A. E. Lundvall to R. A. Clark, Calvert Cliffs Nuclear Power Plant - I i

Docket No. 50-317, " Amendment to Operating License DPR-53 Supplement 1 5th Cycle License ' Application," November 4,1980.

4. CENPD-199-P, "C-E Setpoint Methodology Topical," April 1976.

5. A. E. Lundvall to R. A. Clark, Calvert Cliffs Nuclear Power Plant Unit No. I and Unit No. 2, Docket Nos. 50-310 and 50-319, " Reanalysis of CEA Ejection Event Against New Criteria," May 29, 1981.

6. R. V. fiacBeth, "An Appraisal of Forced Convection Burn-out Data,"

Proc. Instn. Mech. Engrs. 1965-66, Vol.180, Pt. 3c, pp. 37-50.

7. D. M . Lee, "An Experimental Investigation of Forced Convection Burnout

'in High Pressure Water; Part IV. Large Diameter Tubes at About 1600 psia,"

AEEW-R, November 1966.

8. Letter A. E. Lundvall, Jr. to B. C. Rusche,"Calvert Cliffs Nuclear Power Plant Unit No.1, Docket 50-317, Amendment to Operating License DPR-53 Second Cycle License Application," File-L-037-C, October 1,1976.

9. Letter A. E. Lundvall, Jr. (B3&E) to R. A. Clark (NRC), " Core Misloading Analysis," dated June 11, 1981.

10. Calvert Cliffs Nuclear Power Plant FSAR, Section 14.5, Fuel Handling Incident.

11. Letter A. E. Lundvall, Jr. to R. A. Clark, "Fifth Cycle License Application," September 22, 1980.

12. Letter A. E. Lundvall, Jr. to R. W. Reid, " Fourth Cycle License Application," February 23, 1979.

201 References for Chaoter 8

1. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reacters, Federal Register, Vol. 39, No. 3 -

Friday, January 4,1974.

2. CENPD-132, " Calculative Methods for the CE Large Break LOCA Evaluation Model", August 1974 (Proprietary).

CENPD-132, Supplement 1, " updated Calculative Methods for the CE Large Break LOCA Evaluation Model", December,1974 (Proprietary).

CENPD-132, Supplement 2 " Calculational Methods for the CE Large Break LOCA Evaluation Model", July,1975 (Proprietary).

3. CENPD-135, "STRIKIN-II, a Cylindrical Geometry Fuel Rod Heat Transfer

" Program", April,1974 (Proprietary).

CENPD-135, Supplement 2, "STRIKIN-II, a Cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)", February,1975 (Proprietary).

CENPD-135, Supplement 4, "STRIKIN-II, A cylindrical Geometry Fuel Rod Heat Transfer Program", August,1976 (Proprietary).

. CENPD-135, Supplement 5, "STRIKIN-II, A cylindrical Geometry Fuel Rod Heat Transfer Program", April,1977 (Proprietary).

4. CEN-161(B)-P, " Improvements to Fuel Evaluation Model", July,1981 (Proprietary).

5. CENPD-138, " PARCH - A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", August,1974 (Proprietary).

CENPD-138, Supplement 2-P January,1977 (Proprietary),

a

6. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), "Fifth Cycle License Application", dated September 22, 1980.

7. Letter, A. E. Lundvall, Jr. (BG&E) to B. C. Rusche (NRC), "Second Cycle License Application", dated October 1,1976.

4 Esfcrcnca (Chypter 9)

1. Letter, A. E. Lundvall, Jr. (EG&E) to R. A. Clark (NRC), " Phase I Cycle 6 Reload Application," dated November 19, 1981.

Reference (Chapter 10)

1. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark . (NRC), "Fifth Cycle License Applicatien," dated September 22, 1980.

1 M

i e

s