ML20086K763

Cycle 11 Reload Analysis Rept

ML20086K763
Person / Time
Site:	Calvert Cliffs
Issue date:	12/10/1991
From:	BALTIMORE GAS & ELECTRIC CO.
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ML20086K756	List: ML20086K752 ML20086K760 ML20086K763
References
NUDOCS 9112130294
Download: ML20086K763 (93)

Text

.

NITACllMENT.]

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CAINEltT CI,lFFS UNIT 1 CYC1.E 11 REl.OAD ANAINSIS ItEl'OltT The marked up Technical Spectrication pages have been removed from Section 9 and are included as Attachment 1.

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CALVERT CLIFFS UNIT 1 CYCLE 11 LICENSE SUBNITTAL Table of Contents Section 1.

Introduction and Summary 2.

Ope-

ng History of the Previous Cycle 3.

General Description 4.

Fuel System Design

5. -

Nuclear. Design 6.

Thermal Hydraulic Design 7.

Transient Analysis 8.

ECCS Performance Analysis 9.

Technical. Specifications 10.

Startup Testing 11.

References

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1.0 INTRODUCTION

AND

SUMMARY

This report provides an evaluation of design and performance for the operation of Calvert Cliffs Unit 1.at full rated power of 2700 MWt during its eleventh fuel cycle.

All planned operating conditions remain the same as those for Cycle 10. The core will consist of 132 presently operating Batch L and M assemblies, I reinserted Batch K*

assembly which had been discharged at the end of Unit 1 Cycle 9, and 84 fresh Batch H assemblies.

The fresh Batch N fuel will employ the GUARDIAN" Cosign.

This design features an enhanced debris tra) ping capability av improved fuel pin resistance to debris.

It is seing implemented to ntnimize the incidence of debris induced failure and, thereby, to luer the activity level of the reactor coolant.

A quite dissimilar leature, i.e., small flow holes, was introduced in Unit 2 Cyr'e 9 for the samepurpose(seeReference1).

Four of the irradiated Batch H assemblies are the Advanced Nuclear fuel (ANF) demonstration assemblies carried over from Cycle 10 (see Reference 2 for discussion).

These four demonstration assemblies contain Gadolirium bearing fuel pins and were included in the Cycle 10 core as part of the effort to qualify ANF fuel for 24 month cycle operation.

The analyses presented in the main body of this report support the neutronic modelling of the core, the safety evaluation of the core, and the acceptable performance of the C E fuel to be contained in Cycle 11. The discussion in the Appendix of Reference 2 supp%s the performance of the ANF fuel.

The fuel management for Cycle 11 will be similar to that employed for the reference cycle, Unit 2 Cycle 9 (see below), and for Cycle 10, which was the first Unit I cycle to use 1,w-leakage fuel management.

However, to further reduce the fluence to the critical vessel wold, beyond what was initially achieved in the change to low-leakage fuel management in Cycle 10, two modifications are being introduced in Cycle 11.

First, fresh fuel on the periphery near the critical weld is being replaced with twice burned fuel. This change will be referred to as low-fluence fuel management.

Second, Guide Tube Flux Suppressors (GTFSs), which are essentially CEA fingers, are being placed in selected guide tubes near the periphery.

Plant operating requirements have created a need for flexibility in the Cycle 10 termination point.

This need has been met by using an End-of-Cycle 10 window ranging from 17,500 MWD /T to 20,500 MWD /T in the Cycle 11 analyses.

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 11 conditions would be enveloped, provided the Cycle 10 termination point falls within this end of-cycle burnup range. The analysis presented herein supports a Cycle 11 length which varies from 18,700 to 20,600 MWD /T, depending upon the Cycle 10 shutdown burnup, including a coastdown in inlet temperature to 537'F and a coastdown in power to approximately 77%.

1-1

c>

The evaluations of the reload core characteristics have been conducted with respect to the Calvert Cliffs Unit 2 Cycle 9 safety analysis described in Reference 1.

Unit 2 Cycle 9 will hereafter be referred to as the " reference cycle" in this report, unless otherwise noted.

This is the appropriate reference cycle because its design / safety basis is the one most recently reported to the NRC and because the basic system characteristics of the two reload cores are very similar, Unit 2 Cycle 9 being a 24-month cycle using low-leakage fuel management.

Specific core differences have been accounted for in the Unit 1 Cycle 11 analyses.

In all cases, it has been concluded that either the reference cycle analyses envelope the new conditions or the revised analyses discussed herein continue to show acceptable results. Where dictated by variations from the previous cycle (Unit 1 Cycle 10, Reference 2), proposed modifications to the existing plant Technical Specifications are provided and are justified by the analyses discussed herein.

A large proportion of these proposed modifications are identical or nearly identical to those approved (Reference 3) for the reference cycle.

The 'Jnit 1 Cycle 11 analyses used the same methodology as the reference cycle with two significant exceptions.

First, improved physics methods have been implemented, including the accounting for anisotropic scattering effects and the application of the Nodal Expansion Method (NEM) for solving the coarse mesh flux.

Second, the improved fuel performance methodology that justifies internal pin pressures above system pressure (No Clad-Lift-Off) has been applied.

Similar to the modifications made for Unit 2 Cycle 9, changes in the LCO monitoring function are being proposed to acconcodate the usc of the CECOR 3.3 computer code.

Specifically, the CECOR 3.3 code would replace the INCA code and would be combined in an on-line network, as described in Appendix B of Reference 1, to monitor cot;1iance with the LHR and DNB LCOs.

The Unit 2 Cycle 9 Technical Specifications provided, as a contingency, for the continued use of the INCA incore monitoring system.

The requested Technical Specification changes for Unit 1 Cycle 11 have not been drafted to permit the use of INCA because the per formance of the CECOR 3.3 monitoring system for Unit 2 Cycle 9 has been satisfactory.

The performance of Combustion Engineering (C-E) 14x14 fuel at extended burnup was discussed in Reference 4 and approved in Reference 5 for 1-pin burnups up to 52,000 MWD /T.

The further performance of C-E 14x14 fuel at Calvert Cliffs for burnups beyond those considered in Reference 4 was discussed in the Supplement (Reference 6) to the Unit 2 Cycle 9 License Submittal (Reference 1).

That Supplement supports operation to a 1-pin burnup of 60,000 MWD /T, a value considerably in excess of the maximum 1-pin burnup projected for Cycle 11 (approximately 56,900 MWD /T).

1-2

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-2,0

. 0PERATING HISTORY Of THE PREVIOUS CYCLE Calvert Cliffs Unit 1 is presently operating in its. tenth fuel cycle -

utilizing. Batch K, L, and M fuel assemblies.

Calvert Cliffs Unit 1 Cycle.10 began operation on-July 1,- 1988 and reached approximate full power conditSons _ on July 9,1988.' Cycle-10 startup testing was reported to the NRC in Reference 1.

The reactor has operated up to

- the present - time with ~ core. reactivity,- power distributions and peaking factors closely Criloving the calculated predictions.

It is presently estimated that Cyde *0 will terminate-on or about

~

March : 6, 1992.

The Cycle 10 termination point can vary between 17,500 MWD /1 and 20,500 MWD /T to accommodate the plant schedule-and still be within the assumptions of the Cycle 11 analyses.

As of October 1, 1991, the Cyc1s 10 burnup had reached 15,557 MWD /T.

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a 3.0 ENERAL DESCRIPTION The Cycle 11 core will consist of the numbers and types of assemblies and fuel batches as described in Table 3-1.

The fuel management will entail the removal of 85 irradiated assemblies:

4 Batch L.12 Batch L*,

48 Batch K, and 21 Batch K* assemblies. These assemblies will be replaced by 84 fresh aasemblies:

12 unshimmed Batch N assemblies, 20 4-shimmed (B C) Batch NX assemblies, and 52 4

8-shimmed (B C) Batch N/ assemblies at 4.20 wt% U-235 enrichment; and by one h assembly which was discharged at the end of Unit 1 Cycic 9.

Figure 3-1 shows the fuel management pattern to be employed in Cycle 11.

This fuel management will be similar to those employed for the reference cycle (Unit 2 Cycle 9) and Unit 1 Cycle 10.

The only

ignificant change is the use of twice burned fuel in Quarter Core (QC) Locations 2 and 54 (see Figure 3-1) and the placement of Guide Tube Flux Suppressors (GTFSs) in QC Locations 1, 2, 45, and 54. The purpose of these changes is to further reduce the fluence to the critical vesse' weld beyond what was initially achieved in the change to low-leakage fuel management in Unit 1 Cycle 10.

The critical weld is a vertical weld located along the center line of the cere, i.e., adjacent to the upper right corner of QC Location 2.

The use of all twice burned fuel in QC Locations 1, 2, 45, and 54 i

results in what is being termed low-fluence fuel management.

l Figure 3-2 shows the locations of the fuel and B C burnable absorber 4

rods within the NX, N/, and M* shimmed assemblies, and the placement of the fuel only and Gadolinium burnable absorber fuel bearing rods in the ANF demonstration assemblies (MX).

The Unit 1 Cycle 11 fuel management pattern will accommodate Cycle 10 tarminetton burnups l

from 17,500 MWD /T tc 20,500 MWD /T.

l Ti.e Cycle 11 core loading pattern is 90* rotationally symmetric.

That is,- if one quadrant of the core were rotated 90' into its neighboring quadrant, each assembly would be aligned with a similar assembly. This similarity includes batch type, number of fud rods, initial enrichment, burnup, and GTFS placement.

Figure 3-3 shows the beginning of Cycle 11 assembly burnup distribution for a Cycle 10 termination burnup of 17,500 MWD /T. The initial enrichment of the fuel assemblies is also shown in Figure l

3-3.

Figure 3-4 shows the end of Cycle 11 assembly burnup distribution.

The end of Cycle 11 core average exposure is approximately 34,500 MWD /T and the average discharge exposure is approximately 43,050 MWD /T. The end of cycle burnups are based on a Cycle 10 length of 20,500 MWD /T and a Cycle 11 length of 18,700 MWD /T.

L i

i 3-1

1 v

3.1 ANF DEMONSTRATION ASSEMBLIES Four demonstration assemblies, manufactured by Advanced Nuclear Fuel (ANF) and designated as Batch MX, were loaded into Unit 1 Cycle 10.

These assembD.es contained Gadolinium as the burnable absorber material;- they will be returned to the core ir, Cycle 11 for their i

second cycle of irradiation.

The fuel and burnable absorber (Gd 0 23 in 00 ) pin arrangement for these demonstration assemblies is shown 3

in Figure 3-2.

A further discussion concerning these assemblies is contained in the Appendix to Reference 1.

3.2 GUIDE TUBE FLUX SUPPRESSORS Due to Pressurized Thermal Shock considerations, Guide Tube Flux Suppressors (GTFSs) are being added to the core for Cycle 11 to minimize the flux to the critical vessel weld.

These GTFSs are discrete abscrbers in the form of standard B C CEA fingers (see 4

Section 4.1.3).

They will be placed in selected outer guide tubes of those peripheral assemblies nearest the weld and in all quarter core symmetric locations to maintain quarter core symmetry (see j

Figure 3-5). There will be a total of 24 GTFSs in the 6 outer guide tubes closest to the core axes on each " flat" (4 offset outer assemblies).

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TABLE 3-1 l

CALVERT CLIFFS UNIT 1 CYCLE 11 CORE LOADING Initial Assembly Non-Fuel Initial Total Total' Average Batch Average Bearing BC Non-Fuel No. of 4

E,C Rods Loaajpg Bearing Fuel Assembly No. of Enrichment Burnup (MWD /T)

B er Assem.

f ams B /in) D C Rods Rods Desiar,ation Assem.

(wt% U-235)

E0Ellill E0011(2) 4 N

12 4.20 0

17,500 0

0 0

2112 NX 20 4.20 0

20,650 4

.036 80 3440 N/

52 4.20 0

25,000 8

.036 416 8736 M(3) 16 4.06 14,100 34,400 0

0 0

2816 us da AX(3,5) 4 3.89 20,250 42,300 0

0 0

704 M*(3) 76 4.07 21,250 44,500 12

.036 912 12464 L(3) 36 4.04 28,850 40,200 0

0 0

6336 K*(4) 1 3.40 27,650 42,750 0

0 0

176 Total 217 13,750 34,500 1408 36,784 36,763(6)

(1) Cycle 10 burnup of 17,500 MWD /T.

(2) Cycle 10 burnup of 20,500 MWD /T and Cycle 11 burnup of 18,700 MWD /T.

(3) Carried over from Cycle 10 to Cycle 11 of Unit 1.

(4) Reinserted; discharged at the end of Cycle 9 of Unit 1.

(5) ANF Gadolinia bearing demonstration assemblies.

(6) Provision has been made for the displacement of up to 21 fuel rods by stainless steel replacement rods during reconstitution efforts.

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l X

- BOX NUMBER Y

- BATCH I

2 L

L 3

4 5

6 7

L M

N NX NX 8

9 10 11 12 13 L

NX N

N/

M*

N*

14 15 16 17 18 13 20 L

M N*

N/

M*

N/

M*

21 22 23-24 25 26 27 28 L

NX M*

N/

M*

N*

N*

N/

?9 30 31 32 33-34 35 36 N-N N/

N*

N/

N*

-N/

L i

i 37 38 39 40 41 42 43 44 Nl l

45 46 47 48 49 50 51 52 53 NX N*

N/

M*

N/

M*

N/

NJ, 54-L 55 56 57~

58 59 60 61 62 NX N*

M*

N/

L N/

Nj, K*

++ Gadolinia ANF Assemblies BALTIMORE GAS & ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE CALVERT CLIFFS CORE MAP 3-1 NUCLEAR POWER PLANT 3-4

r s

i FIGURE 3-2 l

CALVERT CLIFFS UNIT 1 CYCLE 11 FUEL AND SHIM LOCATIONS 4-SHIM NX ASSEMBLY BP BP kUb!

hub!"

ful!'

fulf' fulf' BP BP 8-SHIM N/ ASSEMBD' BP BP hub hub BP BP hub BP BP kOb hub BP BP

.0$6 gm B-10/in/o U-235 #uel rodchburnab'ieabsorberrod NX N/:

4.20 w Blank BP 3-5

,w.

FIGURE 3-2 (Continued)

CALVERT CLIFFS UNIT 1 CYCLE 11 FUEL AND SHIM LOCATIONS 12-SHIM M* ASSEMBLY or 12-Sti!M MX ASSEMBLY BP BP kub!'

kub!'

BP BP EbY' BP BP Guide Guide Tube -

BP BP Iube BP BP 1

Blank; M*: 4.07 w/o U-235 fuel rod; MX: 4.08 w/o U 235 fuel rod BP M*: _ 0.036 gm B-10/ inch burnable absorber rod

.MX: ANF 10 w/o Gd 0 mixed with natural uranium burnable absorber rod 23 3-6

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- BATCH 1

L 2

L YYY

- INITIAL ENRICHMENT 4.04 4.04 ZZZZZ

-_BURNUP(MWD /T) 31,900 29,800 3

L 4

N 5

M 6 NX 7 NX 4.04 4.20 4.06 4.20 4.20 24,600 0

14,700 0

0 8

L 9 NX 10 M 11 N/ 12 M* 13-M*

4.04 4.20 4.06 4.20 4.07 4.07 29,000 0

13,400 0

22,500 17,300 14 L 15 N 16 M* 17 N/ 18 M* 19 N/ 20 M*

4.04 4.20 4.07 4.20 4.07 4.20 4.07 28,900 0

22,600 0

22,600 0

23,300 21 L 22 NX 23 M* 24 N/ 25 M* 26 N* 27 M* 28. N/

4.04 4.20 4.07 4.20 4.07 4.07 4.07 4.20 24,600 0

22,700 0

22,800 19,000 22,000 0

29 N 30 M 31 N/ 32 M* 33 N/ 34 M* 35 N/ 36 L

4.20' 4.06 4.20 4.07

-4.20-4.07 4.20 4.04 0

13,400 0

22,800

- 0 17,100 0

29,500 37 M 38 N/ 39 M* 40 M*

41 M* 42 M* 43 M* 44 N/

4.06 4.20 4.07 4.07 4.07 4.07 4.07 4.20 45 L 14,700

- 0 22,500 19,000 17,100 21,500 22,300 0

4.04 31,700 46 NX 47 M* 48 N/ 49 M* 50 N/ 51 M* 52 N/ 53 MX 4.20 4.07 4.20 4.07 4.20 4.07 4.20 3.89 54 L

0 22,300 0

21,900

.0 22,200 0

20,200 4.04 29,700 55 NX 56 M* 57 M* 58 N/ 59 L 60 N/ 61 MX 62 K*

4.20 4.07 4.07

'4.20 4.04 4.20 3.89 3.40 0

17,300 23,300 0

29,500 0

20,200 27,600 EOC 10 - 17,500 MWD /T BALTIM0RE GAS & ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE CALVERT CLIFFS ASSEMBLY AVERAGE BURNUP AT BOC 3-3 NUCLEAR POWER PLANT AND INITIAL ENRICHMENT DISTRIBUTION 3-7

oy

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- BATCH 1

L 2

L ZZZZZ

- BURNUP (MWD /T) 40,400 39,900 3

L 4

N 5

M 6 NX 7 NX 35,100 16,400 32,800 20,900 21,800 8

L 9 NX 10 M 11 N/

12 M*

13 M*

40,200 19,800 36,000 24,800 45,600 40,700 14 L 15 N

16 M* 17 N/ 18 M* 19 N/ 20 M*

40,200 19,800 45,100 25,100 46,100 25,000 46,500 21 L 22 NX 23 M* 24-N/ 25 M* 26 M* 27 M* 28 N/

35,100 19,800 45,200 25,100 46,100 41,900 45,500 25,500 29-N 30- M 31 N/ 32 M* 33 N/ 34 M* 35 N/ 36 L

16,400 36,000 25,100 46,100 25,000 40,800 24,900 50,600 37 M 38 N/ 39 ' M* 40 M* 41 M* 42: M* 43 M* 44 N/

45 - L 32,800 24',800 46,100 41,900 40,800 44,200 45,600 25,100 40,300 46 NX 47 M* 48.N/ 49-M* 50 N/ 51 M* 52 N/ 53 MX 54 L 20,900 45,300 25,000 45,500 25,000 45,500 24,300 42,300 39,900 55.NX 56 M*

57 M* 58 N/ 59 L 60 N/ 61 MX 62 K*

21,800 40,700 46,500 25,500 50,600 25,100 42,300 42,700 EOC 10 - 20,500 MWD /T E0C 11 = 18,700 MWD /T

-BALTIM0RE-GAS & ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE CALVERT CLIFFS ASSEMBLY AVERAGE BURNUP AT EOC 3-4 NUCLEAR-POWER PLANT-A.

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FIGURE 3 5 CALVERT CLIFF 5 UNIT 1 CYCLE 11 Gul0E TUBE FLUX SUPPRESSOR LOCATION I

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1 2

3 4

i __ _

5 6

7 8

9 to 11 12 13 s

14 15 16 17 18 19 20 21, 72 23' 24 25 26 27 28 29 30 31 32 33 34 35 36

[7 38 39 40 41 42 43 44 45 46 47 48 69 1

$9 ti 52 53 54 55 56 57 58 59 (C

41 62 63 64 65 66 67 68 69 70 71 72 73 74 75l 76 77 78 79 C0

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83 e

85 26 87 88 89 90 91 92 93 94 95 96 97 V8 99 e

e 100 101 e e

102 103 104 105 106 107 108 109 110 111 112 113 114 115 116 e

e 117 118 o e

119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 e

135 e e 134 136 137 138 139 140 141 142 143 144 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 181 182 123 184 185 186 187 128 189 190 191 192 193 194 190 196 197 198 199 200 201 202 203 204 205 2C L 207 208 209 210 211 212 213 l

214 215 216 217 e

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o e GUIDE TUBE FLUX SUPPRESSOR 3-9

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4.0 FUEL SYSTEM DESIGN 4.1 MECHANICAL DESIGN 4.1.I f 4.1JgJiign The mechanical design of the Batch N reload fuel is similar to that of the Batch L fuel described in the reference cycle submittal (Calvert Cliffs Unit 2 Cycle 9. Reference 1) with the exception of the two items discussed below. The mechanical designs of the Unit 1 Batch M, L, and K fuel assemblies were described in Reference 2.

The main difference between the Unit 1 Cycle 11 Batch N and a.

Unit 2 Cycla 9 Batch L designs is the employment of different debris-resisgantfeatures.

Batch H has been provided with the new GUARDIAN " Design, which entails new grid and fuel pin designs, in place of the small flow hole debris-resistant feature in the Unit 2 Cycle 9 Batch L fuel.

The GUARDIAN" Design employs a redesigned Inconel spacer grid assembly that improves the grid assembly's capability to entrap debris, and redesigned rods that have long, solid Zircaloy 4 end caps to absorb any wear induced by the debris trapped within the Inconel spacer grid assembly.

The rods are secured by a detent spring feature that holds the rods in-place but permits reconstitution, if necessary.

The specific changes in the fuel design are the following:

(1) The overall length of both the fuel and burnable absorber rods is being increased by 0.466 inches, from 146.763 inches to 147.229 inches.

(2) The guide tube length is also being incressed by 0.466 inches, from 148.788 inches to 149.254 inches, to maintain the same shoulder gap.

(3) The length of the lower end cap of both the fuel and burnable absorber rods is being increased by 1.562 inches, from 0.688 inches to 2.250 inches, to provide a solid zircaloy region in the area where debris is to be trapped.

(4) The plenum regions are being reduced by 1.096 inches due to the net effect of increasing the length of the lower end cap (1.562 inches) and the length of the rod (0.466 inches).

(a)

Fuel rod - from 8.375 inches to 7.279 inches.

(b) Bur'.able absorber rod - from 8.B75 inches to 7.779

• oches.

4-1

(5) The height of the lower end fitting is being reduced by 0.191 inches, from 2.937 inches to 2.746 inches.

"T" stanchions are being added to aid in fuel handling.

(6) The position of the active fuel region along with that of the B C burnable absorber region is being raised by 1.371 4

inches due to the net effect of increasing the length of the lower end cap (1.562 inches) and decreasing the height of the lower end fitting (0.191 inches).

(7) The height of the upper end fitting is being decreased by 0.275 inches, from 5.766 inches to 5.491 inches, to keep the overall length of the bundle unchanged. However, this j

change in height is accomplished by decreasing the 1

compression region for the Holddown Spring without a change in the dimensirn of the spring.

Consequently, there is an initial increase in the holddown load of 82 lbs. and some decreased capacity for assembly growth (See Section4.1.2),

b.

The Batch N fuel has larger Zircaloy spacer grid assemblies l

than used in prior batches.

The Zircaloy spacer grid l

assemblies were redesigned so that the fuel rods located along the periphery of the fuel bundle would be able to receive more coolant flow when in contact with adjacent bundles.

This change was accomplished by increasing the size of the outer pin cell through enlargement of the outside envelope of the spacer grid assembly, i

4.1.2 Dimensiop_al Chanaes All C-E fuel-assemblies in Cycle 11 were reviewed for shoulder gap clearance and fuel assembly clearance using the SIGREEP models described in Reference 3.

This review included consideration of the decreased capacity for assembly' growth of the new Batch N reload l

fuel which employs the GUARDIAN Design (see discussion above in Section 4.1.1.a.7).

All clearances were found to be adequate for Cycle J.

In addition, the increased holddown load does not produce any unacceptable effects on the fuel or associated reactor internal components.

l 4.1.3

[EA Desics During the Unit 2 Cycle 8 to Cycle 9 refueling oub.ge, significant swelling was observed near the tip of one of the " weak" A1 0 23 fingers of the low strength center CEA.

An assessment, which included measurements of the observed swelling of this CEA finger-l l

and of other CEA fingers of the same design, determined that an adverse reaction was occurring at the int'erface between the A10 l

pellets that comprise the bulk of the pellet stack and the zircaf,oy _

3 l

slug that resides at the bottom of the pellet stack.

The affected CEA was replaced during the Unit 2 cutage with a new CEA that corrected the swelling problem, f

4-2 i

F

s a

Unit I was in the midst of Cycle 10 when this problem was identified.

Since the Unit 1 center CEA's design was identical to the Unit 2 CEA which experienced the above described problem, a concern was raised regarding continued operation of Cycle 10.

Evaluations determined that Cycle 10 operation could continue safely and that, indeed, the center CEA, in view of its low reactivity worth, could be exempted from conformance to alignment requirements.

Emergency Tech Specs supporting this position were approved as a contingency against potential future problems (which have not occurred to-date).

During the Unit 1 Cycle 10 to Cycle 11 outage the center CFA will be replaced with a new CEA identical in design to the replacement CEA loaded into Unit 2 during its Cycle 8 to Cycle 9 Outage (see above).

This CEA corrects the swelling problem by replacing the zircaloy slug at the bottom of the stack with a stainless steel slug.

The replacement of the zircaloy slug eliminates the potential for hydriding of the zirconium which took place due to the loss of the zircalloy oxide coating caused by the abrasion between the Al 023 and zircaloy materials.

In all other aspects the design of the new CEA is identical to that of the CEA being replaced.

4.1.4 Guide Tube Flux Suppressors At selected core locations (see Figure 3-5) neutron flux suppressors, called Guide Tube Flux Suppressors (GTFSs), will be installed in the fuel bundle assemblies.

The basic design of the GTFSs is identical to that of control rod fingers with regard to the spacer pellets, and Inconel 625 cladding.

The B,,C pellets, AL 023 pellet column in the active core region has the following core centerline with 10.25 inches,C located symmetrica h around the configuration: 116.2 inches of B, of AL 0 spacers at each end.

23 The GTFSs are seated on the bottom end fitting of the CEA guide

-tubes and are held in place by a spring loaded mechanism similar to that used previou for the neutron source assemblies.

The upper end of the GTFSs are axially loaded and aligned at the top by the fuel alignment plate.

4.2 THERMAL DESIGN The thermal performance of a composite standard uranium dioxide fuel pin that envelopes the various assemblies which will be present in Cycle 11 (Batches K, L, M, and N) has been evaluated using the FATES 3B version of the fuel evaluation model (References 4, 5 and 6) in conjunction with the maximum pressure methodology (No-Clad-Lif t-Off) described in Reference 7.

FATES 3B received NRC approval for application to the Calvert Cliffs reactors in Reference 8.

The thermal performance analysis was performed with a history that modeled the power and burnup levels representative of the peak pin at each burnup interval, from beginning of cycle to end of cycle i

4-3

a

. burnups. The analysis-included the reduction in internal pin volug in the fresh Batch N fuel due to the introduction of the GUARDIAN Design.

The burnup range ' analyzed was in excess of that expected for Cycle-11.

The maximum fuel pin internal pressure was verified to remain below the No Clad Lift-Off critical pressure.

P

.4 4-4 i.

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I.

~ ' - '

4-5.0 NUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS 5.1.1 Fuel Manaaement The Cycle 11 fuel management employs a low-fluence pattern along with GTFSs, as doscribed in Section 3. Figures 3-1 and 3-5.

This-arrangement of fuel and CEA -fingers, i.e., GTFSs, results in very low-fluence to the critical pressure vessel weld which is located along the center line of the core, i.e., adjacent to the upper right corner-of Quarter Core Box No. 2.

Due to the shift in power and fluence away from that part of the periphery near the critical weld, the Cycle 11 fuel and GTFS arrangement yields slightly high power levels in the interior of the core.

t The fresh Batch N fuel is comprised of three sets of assemblies, all using non-poison fuel pins of just one enrichment (4.20 wt% U-235) with each set containing a unique number of B,C shims per assembly.

The unique number of shims per assembly was chosen to minimize radial-power peaking and to control BOC MTCs. S)ecifically, Batch N consists of 12 unshimmed assemblies,.20 assemblues with 4 B C shims-4 per assembly, and 52 assemblies win 8 B C shims per assembly. With 4

this loading,' the Cycle 11 burnup capacity for full power operation is expected to be between 17,700 MWD /T and 19,600 MWD /T, depending on the final Cycle 10_ termination point.

-The Cycle 11 core characteristics have been examined for Cycle 10 terminations between 17,500 and 20,500 MWD /T and limiting values established for the safety analyses.

The loading pattern (see Section 3) is applicable to any Cycle 10 termination point between the stated extremes.

Physics characteristics for Cycle 11 are listed in Table 5-1 along with the corresponding' values - from the reference cycle (Reference 1).

Please note that the values of parameters actually employed in safety analyses are different from those disolayed in Table 5-1 and are typically chosen to conservatively bound predicted values with accommodation for appropriate uncertainties anc allowances.

Table 5-la presents a summary of CEA shutdown worths and reactivity-allowances for the end of Cycle 11 zero power steam line break accident-and a comparison to reference cycle data.

The E0C zero--

power stean line break accident was selected since it is the most limiting zero power transicnt with respect to reactivity _

requirements and, thus, provides the basis for verifying the Technical' Specification required shutdown margin.

Table 5-2 shows the reactivity worths of the three CEA groups which are allond in the core during critical conditions.

These reactivity worths were calculated at full power conditions for Cycle 11 and the reference cycle. The CEA configurations are identical to 5-1

e.

s those of the reference cycle.

The power dependent insertion limit (PDIL) curve is also the same as that of the reference cycle.

l 5.1.2 P_gwer Distributions figures 51 through 5-3 illustrate the all rods out (AR0) integrated 1

radial power distributions at BOC11, H0Cll and E0011, respectively, that are characteristic of the high burnup end of the Cycle 10 shutdown window.

The high burnup end of the Cycle 10 shutdown window tends to increase the integrated 1-pin radial power peaking.

The integrated radial power distributions with CEA Group 5 fully inserted at beginning and end of Cycle 11 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 10 shutdown window.

The reactivity level and location of the ANF assemblies together yleid maximum 1-pin peaks in the ANF assemblies which are predicted to be at least 30% below the maximum 1-pin peak in the core for standard operating conditions (see Figures 51 through 5-5).

It should be noted that the Gadolinium in these demonstration assemblies has burned out to residual levels by BOC11.

The radial power distributions described in this section are calculated data without uncertainties or other allowances. However, the single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice.

For both DNB and kw/ft safety and setpoint analyses in either rodded or unrodded configurations, the power peaking values actually used are higher than those expected to occur at any time during Cycle 11.

These conservative values, which are used in Section 7 of this document, establish the allowable limits for power peaking to be observed during operation.

The range of a110wable axial peaking is defined by the Limiting Conditions for Operation (LCOs) that are dependent on Axial Shape Index (ASI).

Within these ASI limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes.

The maximum three-dimensional or total peaking factor anticipated in Cycle 11 during normal base load, all rods out operation at full power is 1.91, not including uncertainty allowances.

5.1.3 Safety Related Data (Ejected CEA and Drop CEA Data)

The Cycle 11 safety related data for this section are identical to the safety related data used in the reference cycle.

5.1.4 GUARDIAN" Desion The fresh Batch N fuel in Cycle 11 will use the GUARDIAN

• Design.

This design includes a longer fuel pin end cap which results in the active region of the fresh fuel being raised 1.371 inches relative to the irradiated fuel (see Section 4.1.1).

This offset between the fresh and irradiated fuel has been considered in the Cycle 11 5-2

s neutronic analyses and adjustments to the safety /setpoint/

performance data have been made, where appropriate.

5.2 ANALYTICAL INPUT TO IN CORE MEASUREMENTS In-Core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the same manner as those for the reference cycle.

Moreover, the same change in the LCO monitoring function, i.e.,

the use of the CECOR 3.3 computer code, is being proposed for Unit 1 Cycle 11 as was proposed and approved for Unit 2 Cycle 9 (See Section 5.2 of Reference 1).

The Technical Specification changes associated with the use of CECOR 3.3 are contained in Section 9.

5.3 NUCLEAR DESIGN METHODOLOGY Analyses have been performed in the same manner and with similar methodologies to those used for the reference cycle analyses.

However, several method improvements have been implemented for Unit 1 Cyclu 11.

Those improvements include the use of anisotropic scattering within pin cells and anisotropic neutron currents at pin cell interfaces in the DIT code.

Improvements were also made in the ROCS code via the application of the Nodal Expansion Method (NEM),

which is based upon a

fourth order

solution, and by the implementation of assembly discontinuity factors.

The improved methodology is discussed in approved topical reports (References 2 and 3) with the exception of the application of assembly discontinuity - factors which is, today, a widely accepted and used procedure in the industry.

The new DIT/ ROCS methodology is being implemented to improve core power distribution predictions.

This improvement is obtained via the calculation of more accurate global radial power and local fuel pin power distributions.

The global radial power distribution is particularly enhanced through the better modelling of the power sharing between neighboring assemblies.

The local pin power distribution is improved via the better modelling of both intra and inter pin phenomena.

In addition to improved power distribution predictions, the new methodology improves upon the calculation of control rod worths.

The remaining impact of the improved methodology on the reload analyses performed for Unit 1 Cycle 11 has been essentially limited to the application of a revised set of biases and uncertainties determined specifically for use with the improved methodology.

The effect of the new bias and uncertainty values pertaining to scram worth / shutdown mvgin calculations can be seen on Line 6 of Table 5-la.

The old bias (4%) and old uncertainty (9%) shown for the reference cycle (Unit 2 Cycle 9) resulted in a total allowance of 13%.

The new bias (0%) and new uncertainty (7%) that were applied to the Unit 1 Cycle 11 calculated data yielded a reduced total allowance of 7%.

5-3

s The neutronic modeling of the ANF Gadolinium demonstration assemblies was perfonned in accordance with the methods described in C-E's Gadolinium topical report (Reference 3).

5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties to be applied to Cycle 11 are the same as those applied to the reference cycle which presently uses the CECOR 3.3 code. These uncertainties are the same as those used historically with the INCA code (Reference 4).

Although the basic neutronic modeling used in design / safety analyses and also in the generation of data used by the incore monitoring code, CECOR 3.3, has been changed, as discussed in Section 5.3, the measurement uncertainties presented in Reference 4 have been verified as continuing to be applicable.

5-4

y a

4 TABLE 5-1 CALVERT CLIFFS UNIT 1 CYCLE 11 NOMINAL PHYSICS CHARACTERISTICS Reference Cycle Unit 1 UD111 (Unit 2 Cycle 9)

Cycle 11' Dissolved Boron Hot Full Power, All Rods Out,-Equil. XE Boron Content for Criticality at BOC :

ppm 1456 1482 Inverse Boron Worth Hot Full Power, BOC ppm /%Ap 122 124 Hot Full Power, EDC ppm /%Ap 85 87 Moderator Temperature Coefficient Hot Full. Power, Equil. Xe, CEAs Withdrawn-

. Beginning of Cycle 1(T'Ap/* F 0.04

-0.04 End of Cycle 10ap/* F

-2.3

-2.5 BOCll data were calculated using the early Cycle-10 shutdown burnup of

+

17,500 MWD /T; EOC11 data were calculated using appropriate end of Cycle 11 conditions.

5-5 4

w m

e--

-~w,

-m m

-r s

TABLE 5-la CALVERT CLIFFS UNIT 1 CYCLE 11

-LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR THE END-OF-CYCLE (EOC) HOT ZERO POWER (HZP)

STEAM LINE RUPTURE ACCIDENT, E p Reference Cycle Unit 1

.(linit 2 Cvq d

. Cycle 11 1.

Worth of all CEA's Inserted 9.7 8.1 2.

Stuck CEA Allowance 1.5 1.2 3.

Worth of all CEA's less Worth of Most Reactive Stuck CEA 8.2 6.9 4.

Power Dependent Insertion Limit CEA' Bite at Zero Power 2.0 2.0

5. -Calculated Scram Worth 6.2 4.9 6.

Physics Uncertainty plus Bias 0.8 0.4 7.

Net Available Scram Worth 5.4 4.5 8.,- Technical Specification Shutdown 5.0 4.5 Margin 9.

Margin in Excess of Technical Specification' Shutdown Margin 0.4 0.0 5-6

e o

l TABLE 5-2 CAlifERT CLIFFS UNIT 1 CYCLE 11 REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %Ap Beainnino of Cycle End of Cycle Regulating Reference Unit 1 Reference Unit 1 CEA's Cycle *

[1cle_11 Cycle

• Cycle 11 Group 5 0.37 0.36 0.40 0.45 Group 4 0.72 0.75 0.00 0.89 Group 3 0.90 0.86 1.03 1.05 t

!!91 1 Values shown assume sequential group insertion; values are biased.

• Unit 2 Cycle 9 5-7

X

- BATCH 1

L 2

L ZZZ

- RPD 0.25 0.33 3

L 4

N 5

M 6 NX 7 NX 0.41 0.93 0.87 1.16 1.21 8

L 9 NX 10 M 11 N/

12 M*

13 M*

0.42 1.11 1.17 1.36 1.06 1.10 14 L

15 N

16 M* 17 N/

18 M* 19 N/ 20 M*

0.42 1.12 1.05 1.37 1.07 ! 1.32 1.04 4

21 L 22 NX 23 M* 24 N/ 25 M*

25 M*

27 M* 28 N/

0.41 1.11 1.05 1.35 1 05 1.04 1.05 1.32 29 N 30 M 31 N/ 32 M* 33 N/ 34 M* 35 N/

36 L

+

0.93 1.17 1.37 1.05 1.30 1.06 1.26 0.90 M*i,40 M*l41 M*l42 M*i43 M*l at N/

37 M 38 N/ 39 45 L

0.87 1.36 1.07 1.04,1,06;0.97(0.98,.5.25 0.25 46 NX 47 M* 48 N/{49 M* 50 N/ 51 M*j52 N/

c.

MX 54 L

1.16 1.06 1.32 1.05 1.27-0.99 1.20 0.89 0.33 55 NX 56 M*

57 M* 58 N/

59 L 60 N/ 61 MX' 62 K*

t+

1.21 1.10 1.04 1.32 0.90 1.25 0.89 0.72 NOTE: + = MAXIMUM 1-PIN PEAK - 1.60

++ - MAXIMUM l-PIN PEAK IN ANF ASSEMBLY - 1.02 BALTIM0RE GAS & ELECTRIC CO, CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURF CALVERT CLIFFS ASSEMBLY RELATIVE POWER DENSITY 5'

NUCLEAR POWER PLANT AT BOC, EQUILIBRIUM XENON 5-8

no X

- BATCH 1

L 2

L ZZZ

- RPD 0.28-0.35 3

L 4

N 5

H 6 NX 7 NX 0.41 0.86 0.83 1.11 1.15 8

L 9 NX 10 M 11 N/ 12 M*

13 M*

0.42 1.04 1.08 1.32

-1.03 1.07 14 L 15 N 16 M* 17 N/ 18 M* 19 N/ 20 M*

0.42 -

1.04 1.00 1.34 1.05 1.34 1.04 21 L 22 NX 23 M* 24 N/ 25 M* 26 M* 27 M* 28 N/

+

0.41-1.04-0.99 1.34 1.04 1.06 1.07 1.38 29 N 30

-M 31 N/ 32 M* 33 N/ 34 M* 35 N/ 36 L

0.86-1.08 1.34 1.04 1.35 1.11 1.35 0.96 37 M 38 N/ 39 M* 40 M*

41 M* 42 M* 43 M* 44 N/

45 L

0.83 1.32 1.05 1.06 1.12 1.04 1.06 1.37 0.28 46 NX 47 M* 48 N/ 49 M* 50 N/ 51 M* 52 N/ 53 MX

++

54 L

1.10 1.03 1.34 1.07 1.36 1.06 1.32 0.97 0.'35 55 NX 55 M* 57 M* 58 N/ 59 L 60 N/ 61 MX 62 K*

+-

++

1.15 1.07 1.04-1,38 0.96 1.37 0.97 0.82 NOTE: + = MAXIMUM l-PIN PEAK - 1.54

++ = MAXIMUM 1-PIN PEAK IN ANF ASSEMBLY = 1.09 BALTIMORE

. GAS & ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE.

CALVERT CLIFFS-ASSEMBLY RELATIVE POWER DENSITY 5-2 NUCLEAR POWER PLANT AT 9.5 GWD/T, EQUILIBRIUM XENON 5-9 9

,e

$5 e

TX

~- BATCH 1

L 2

.L ZZZ

- RPD 0.32 0.41 3

L 4

N

-5 M

6 NX 7 'NX 0.45 0.89 0.86 1.15 1.19 8

L 9 NX 10 M 11 N/ 12 M* 13 M*

0.46 1.07 1.07 1.33 1.03 1.06 14 L 15 N 16 M* 17 N/ 18 M* 19 N/ 20 M*

0.46 1.05 0.99 1.34 1.03 1.33 1.02 21 L 22 NX 23 M* 24 N/ 25 M* 26 M* 27 M* 28 N/

0.45

'l.07 0.99 1.34 1.02 1.01 1.03 1.35 29 N 30 M 31 N/ 32 M* 33 N/ 34 M* 35 N/ 36 L

0.89 1.07.

1.34 1.01 1.31 1.06 1.33 0.90 37.

M 38 h/ 39 M* 40 - M* 41 M* 42 M* 43 M* 44 N/

.+.

45 L

0.86 1.33 1.03 1.01 1.06 0.99 11.03 1.35 0.32 46 NX 47 M* 48 N/ 49 M* 50 N/ 51 M* 52 N/ 53 MX

++

54 L

1.14-1.02 1.33 1.03 1.33 1.03 1.32 0.96

'0.41 55 NX 56 M* 57 M* 58 N/ 59 L 60 N/ 61 MX 62 K*

++

1.19 1.06 1.02 1.35 0.96 1.35 0.96 0.83 NOTE: + = MAXIMUM 1-PIN PEAK = 1.53

++ - MAXIMUM l-PIN PEAK IN ANF ASSEMBLY = 1.06 BALTIMORE GAS.& ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE CALVERT CLIFFS ASSEMBLY-RELATIVE POWER DENSITY 5-3 NUCLEAR POWER PLANT AT EOC, EQUILIBRIUM XENON L

5-10 i

i.

.o e

CEA BANK 5 1

L 2

L LOCATION 0.21 0.27 p

3' L

4 H

5 til6 NX 1

NX 0.43 0.94 0.82 0.96 0.90 8

L 9 NX 10 M 11 N/

12 M*

13 M*

0.45 1.18 1.21 1,32 0.88 0.59p 14

-L 15 N 16 M* 17 N/

18 M* 19 N/ 20 M*

0.45 1.21 1.14 1.44 1.08 1.22 0.91 21 L 22 NX 23 M* 24 N/ 25 M* 26 M*

27 M* 28 N/

0.44-1.18 1.13 1.45 1.13 1.10 1.07 1.32 29 N 30 M 31 N/ 32 M* 23 N/ 34 M* 35 N/ 36 L

+

0.94 1,20 1.44 1.13 1.40 1.15 1.34 0.96 37 M 38 N/ 39 M* 40 M* 41 M* 42 M* 43 M* 44 N/

45 L

0.82 1.32 1.07 1.09 1.15 1.07 1.07 1.36 0.21 46 NX 47 M* 48 N/ 49 M* 50 N/ 51 M* 52 N/ 53 MX

++

54 L

0.96 0.88 1.23 1.07 1.36 1.07 1.31 0.97 0.27 55 NX 56 M*

57 M* 58 N/ 59 L 60 N/ 61 MX 62 K*

++

0.90 0.59p 0.91 1.32 0.96 1.36 0.97 0.67p NOTE: + = MAXIMUM 1-PIN PEAK - 1.68

++ = MAXIMUM l-PIN PEAK IN ANF ASSEMBLY - 1.12 BALTIMORE-GAS & ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE CALVERT CLIFFS ASSEMBLY RELATIVE POWER DENSITY 5-4 NUCLEAR POWER PLANT WITH BANK 5 INSERTED, MFP, BOC 5-11

O CEA BANK 5 1

L 2

L LOCATION 0.27 0.34 p

3 L

4 N

5 M

6 NX 7 NX 0.48 0.91 0.82 0.95 0.89 8

L 9 NX 10 M 11 N/

12 M*

13 M*

0.51 1.15 1.11 1.29 0.85 0.53p 14 L 15 N 16 M* 17 N/ 18 M* 19 N/ 20 M*

0.51 1.15 1.08 1.42 1.04 1.24 0,87 21 L 22 NX 23 M* 24 N/ 25 M* 26 M*

27 M* 28 N/

+

0.48 1.15 1.08 1.45 1.10 1.06-1.05 1.35 29 N 30 M 31 N/ 32 M* 33 N/ 34 M* 35 N/ 36 L

0.91 1.12 1.42 1.09 1,42 1.15 1.42 1.01 37 M 38 N/ 39 M* 40 M*

41 M* 42 M* 43 M* 44 N/

45 L

0.82 1.29 1.04 1.07 1.15 1.03 1.12 1.47 0.27 46 NX 47 M* 48 N/ 49 M* 50 N/

51 M* 52 N/ 53 MX

++

54 L

0.95 0.85 1.24 1.06.

1.42 1.12 1.41 1.02 0.34 55 hX 56 M*

57 M* 58 N/ 59 L 60 N/ 61 MX 62 K*

0.89 0.53 0.87

' 35 1.01 1,47 1.02 0.74p l

p 1

NOTE: + = MAXIMUM 1-PIN PEAK - 1.64

++ = MAXIMUM l-PIN PEAK IN ANF ASSEMBLY - 1.13 B% TIM 0RE GAS & ELECTRIC CO.

CALVERT CLIFFS UNIT 1 CYCLE 11 FIGURE l

CALVERT CLIFFS ASSEMBLY RELATIVE POWER DENSITY 5-5 NUCLEAR POWER PLANT WITH 8ANK 5 INSERTED, HFP, EOC 5-12 l

4 6.0 THERMAL HYDRAULIC DESIGN 6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 11 at the rated power level of 2700 MWt have been performed using the 10RC and CETOP computer codes, the CE-1 critical heat flux correlation and simplified modeling methods, as described in References I through 4 and approved in Reference 5.

The fresh assemblies in Cycle 11 will employ the GUARDIAN" Design along with large envelope Zircaloy grids. This fuel design includes debris-resistant features which result in greater hydraulic resistance than the reference design.

This increase in hydraulic resistance results in a decrease in the inlet flow for the fresh fuel assemblies. The TORR and CETOP models used in the Cycle 11 DNB analyses accounted for this flow reduction.

Table 6-1 contains a list of pertinent thermal-hydraulic design parameters.

The calculational factors (engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in Table 6-1 have been combined statistically 4th other uncertainty faciors using the Extended Statistical Combination of Uncertainties (ESCU) methods, described in Reference 6 and approved in Reference 7, to derive an overall uncertainty allowance which, when used with the CE-1 CHF correlation design limit of 1.15 for 14x14

fuel, provides a

95/95 probability / confidence level-of assurance against DNB occurring during steady state operation or anticipated operational occurrences.

The statistically derived ESCU uncertainty allowance includes a 0.006 DNBR rod bow penalty which accounts for the tdverse effects of rod bowing on CHF for 14x14 fuel with burnup not exceeding 45 GWD/T.

6.2 EFFECTS OF FUEL BOWING ON DNBR MARGIN The effects of fuel rod bowing on DNB margin for Cycle 11 have bcen evaluated using the NRC approved methods described in Reference 8.

Based upon-these methods, a DNBR penalty of 0.006 is required to account for the adverse T-H effects of rod bow at an assembly average burnup of 45 GWD/T.

This penalty is aresently included in the ESCU uncertainty allowance, as discussed a)ove.

For assemblies which will attain burnups greater than 45 GWD/T, the power peaking will be significantly lower than that of other assemblies, providing ample margin to offset increases in the roa bow penalties for these high burnup assemblies.

6-1

TABLE 6-1 CALVERT CLIFFS UNIT 1 CYCLE 11 THERML-HYDRAULIC PARAMETERS AT FULL POWER

• Reference Cycle **

Unit 1 General Characteristics Mail (Unit 2 Cycle 9)

Cycle 11 Total Heat Output (core only)

My 2700 2700 10 BTU /hr 9215 9215 Fraction of Heat Generated

.975

.975 In fuel Rod Primary System Pressure psia 2250 2250 (Nominal)

Inlet Temperature

'F 548 548 Total Reactor Coolant Flow gpg 381,600 381,600 (steady state) 10 lb/hr 143.8 143.8 6

Coolant Flow Through Core 10 lb/hr 138.5 138.5 Hydraulic Diameter ft 0.044 0.044 (nominal channel) 6 2

Average Mass Velocity 10 lb/hr-ft 2.59 2.59 Pressure Drop Across Core psi 11.3 13.1 (steady state flow irreversible ? over entire fuel assembly)

Total Pressure Drop Across psi 34.9 36.7 Vessel (basM on c:eady state flow an6 nominal dimensions) 2 Core Average Heat Flux BTU /hr-ft 186,000***

186,600****

(Accounts for above fraction of heat generated in fuel rod and axial densification factor) 2 Total Heat Transfer Area ft 48,300***

48,150****

(Accounts for axial densification factor) 2 Film Coefficient at Average BTU /hr-ft,.F 5930 5930 Conditions 6-2

\\

TABLE 6 1 (continued)

Reference Cycle **

Unit 1 Gtperal Characteristin Unit (Unit 2 Cvele 91 Cycle 11 Aver 9ge film Temperature

'F 31 31.5 Difference Average Linear Heat Rate of kw/ft 6.27***

6.29****

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

Average Core Enthalpy Rise BTU /lb 66.5 66.5 Maximum Clad Surfcce

'F 657 657 Temperature

[1]rulational Factors Engineering Heat Flux on Hot Channel 1.03+

1.03+

Engineering Factor on Hot Channel 1.02 1.02+

4 Heat Input Rod Pitch and Clad Diameter Fac*.or 1.065+

1.065+

Fuel Densification factor (axial) 1.002 1.002 IID1H

• Due to the Fxtended Statistical Combination of Uvrtainties methodology dcicribed in Referenca 6, the nominal inlet temperature and nominal primary system pressu'J were used to calcuiate some of these parameters.

• • Reference cycle analysis (Unit 2 Cycle 9) is described in Reference 9.

• • • Based on a value of 1296 B C shims plus 21 stainless steel pins.

4

• • • • Based on a value of 1408 B C shtms plus 21 stainless steel pins.

4

• These factors have been combined statistically with other uncertainty factors at the 95/95 confidence / probability level (Reference 6) to derive an overall uncertainty allowance, as discussed in Reference 6 and approved by the NRC in Reference 7.

This allowance was verified to be applicable to Unit 1 Cycle 11.

6-3 l

7.0 IPM151tHT ANALYSIS The Design Basis Events (DBEs) considered for the Unit 1 Cycle 11 non.LOCA safety analyses are listed in Table 7-1.

Core parameters input to the safety analyses for evaluating approaches to DNB and Centerline Temperature to Helt (C1H) fuel design limits are presented in Table 7-2.

For all DBEs the key transient input parameters were equal to or conservative with respect to the referener cycle values (Unit 2 Cycle 9. Reference 1) with one exception. The Shutdown Margin (SDH) at end of cycle decreased from 5.0%ap to 4.5%8p.

This change in SDM, a result of the low fluence fuel management used for Unit 1 Cycle 11, impacted the Steam Line Repture event. The results of tha

.g reanalysis of this event were bounded by those of the referer a g

cycle.

In addiyon to key in >ut parameters, the implementation of the GUARDIAN Design was a'so considered.

This evalaction dettermined that the results for the reference cycle remain bounding.

7-1 f

e TABLE 7-1 CALVERT CLIFFS UNIT 1 CYCLE 11 DESIGN BASIS EVENTS CONSIDERED IN THE NON LOCA SAFETY ANALYSIS Desian Basis FJmtt enalysis Status 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits:

7.1.1 Boron Dilution Not Reanalyzed 7.1.2 Startup of an Inactive Reactor Coolant (1)

Pumn Los' of Load Not Reanalyzed 7.1.3 s

7.1.4 Loss of Feednter Flow Not Reanalyzed 7.1.5 Excess Heat Removal due to Feedwater Not Reanalyzed Malfunction 7.1.6 Reactor Coolant System Depressurization Not Roanalyzed 7.1.7 Excessive Charging Event Not Reanalyzed 7.2 Anticipated Operational Occurrences for which RPS trips and/or stfficient initial steady state thermal margin, maintained by the LCOs, are necessary to prevent exceeding the acceptable limits:

7.2.1 Sequential CEA Group Withdrawal Not Reanalyzed 7.2.2 Loss of Coolant Flow Not Reanalyzed 7.2.3 Full Length CEA Drop Not Reanalyzed 7.2.4 Transients Resulting from the Not Reanalyzed Halfunction of One Steam Generator 7.2.5 Loss of AC Power Not Reanalyzed 7.2.6 Excess Load Not Reanalyzed 7.3 Postulated Accidents 7.3.1 CEA Ejection Not Reanalyzed l

7.3.2 Steam Line Rupture Reanalyzed (2) 7.3.3 Steam Generator Tube Rupture Not Reanalyzed 7.3.4 Seized Rotor Not Reanalyzed 7.3.5 Feed Line Break Not Reanalyzed 7.4 Postulated Occurrence i

7.4.1 Fuel Handling Incident Not Reanalyzed l

Notes (1) Technical Specifications preclude this event during operation.

i (2) This event was reanalyzed to assess the impact of changes in fuel l

management which were principally responsible for the decrease in the end of cycle SDM from 5. Cap to 4.5Y.ap.

The results for Unit 1 Cycle 11 were determined to be less limiting than those previously reported.

Consequently, further discussion is not included herein.

l l

7-2

TABLE 7 2 CALVERT CLIFFS UNIT 1 CYCLE 11 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM DESIGN LIMITS Reference Cycle

• i Physics Parameters Uniti (Unit 2 Cycle 9)

Unit 1 Cvele 11 Radial Peaking Factors F/ (DNB Margin Analyses)

Unrodded Region 1.70**

1.70**

Bank 5 Inserted-1.904**

1.904**

F,/ (CTM Limit Analyses)

Unrodded Region 1.70**

1.70**

Bank 5 Inserted.

1.904**

1.904**

Moderator Temperature 10'4Ap/'F

+.7 - -2.7

+. 7 -. - 2. 7 Coeffkient Shutdown Aargin Mp

-3.5 9 BOC 3.5 9 BOC i

to to

-5.0 9 E0C

-4.5 9 EOC

. Tilt Allowance 3.0 3.0 Power Level MWt 2700**

2700**

Maximum Steady State

'F 548**

548**

Inlet. Temperature Minimum Steady State psia 2200**

2200**

RCS Pressure-Reactor Coolant Flow gpm 381,600**

381,600**

Negative Axial Shape

.15**,+

.15**,+

LC0 Extreme Assumed P

at Full Power (Ex Cores)

Maximum CEA Insertion

% Insertion 35 35 at Full Power of Bank 5 Maximum initial Linear KW/ft 16.0+

16.0+

Heat Rate for Transients' Other than LOCA 73 i

i n

.. - +.,,eE-

~..., -,-,...,,.,

.--.,.<,...,,,-,...,,,,,,n n.,,,,n

..,n-

o o

l TABLE 7-2 (continued)

Reference Cycle

• Safety Parameters Uniti (Unit ? Cycle 9)

Unit 1 Cycle 11 Steady State Linear KW/ft 22.0 22.0 Heat Rate for Fuel CTH Assumed in the Safety Analysis CEA Drop Time from sec 3.1 3.1 Rwval of Power to Hviding Coils to 90%

Insertion MinimumDNBR(CE-1) 1.15**

1.15**

Doppler Coefficient, HFP 10-5 Ap/'F Haximum

-2.4

-2.4 Minimum

-1.0

-1.0 Total Effective Delayed Neutron Fraction, A,77 Maximum

.0070

.0070 Minimum

.0044

.0044 Neutron Generation Time, f*

10-6 3,c Maximum 38.0 38.0 Minimum 16.0 16.0 llDift1

• Reference 1

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

The procedures used in the Statistical Combination of Uncertainties (References 2, 3, and 4) and Extended Statistical Combination of Uncertainties (Reference 5) programs have been applied to DNB and CTH limits.

These procedures have been approved by NRC for the Calvert Cliffs Units in References 6 and 7, respectively.

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

7-4 l

o 8.0 ECCS PERFORMANCE ANALYSIS

8.1 INTRODUCTION

AND SUMKARY An ECCS perfomance analysis was performed for Calvert Cliffs Unit 1 Cycle 11 to demoristrate conformance to the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors (Reference 1).

The limiting Large Break LOCA (LBLOCA) and Small Break LOCA (SBLOCA) were analy7"ed.

Both of those analyses accounted for the use of GUARDIAN grids.

The SBLOCA also justified the plugging of up to 500 tubes per steam generator (the LBLOCA had previously justified the plugging of up to 500 tubes per steam generator).

These two changes required the reanalysis of the hydraulic port lon of both analyses.

The analysis bifies an allowable Peak Linear Heat Generation Rate (PLHGR) of !5.L : w/f t.

This PLHGR is equal to the existing limit for Calvert fitff> ll $.

The method of analysis and results which support thit vclub tre ', resented in the following sections for both LBLOCA and St,'(y'.

8.2 LARGE BPEAK LOSS O' COOLANT ACCIDENT 8.2.1 Method of Analysis The LBLOCA ECCS performance analysis for Calvert Cliffs Unit 1 Cycle 11 was performed using the 1985 Evaluation Model (Reference 2). The 1985 Evaluation Model is approved by the NRC (Reference 3).

The LBLOCA reference cycle analysis for Cycle 11 is the analysis for Unit 1 Cycle 10 (Reference 4).

The method of analysis for Cycle 11 is identical to that used for the reference cycle.

Blowdown hydraulic calculations were performed with the CEFLASH 4A computer code (Reference 5) and refill /reflood hydraulic calculations were performed with the COMPERC Il computer code (Reference 6).

Hot rod cladding temperature and cladding oxidation calculations were performed with the STRIKIN-Il computer code (Reference 7).

Steam cooling heat transfer coefficients were calculated using the PARCH and HCROSS computer codes (Reference 8).

Core wide cladding oxidation was also calculated in this analysis.

The analysis used fuel performance data which bound Cycle 11 and which are expected to bound future cycles.

The fuel performance data were calculated with the FATES 3B computer code (References 9 through II). FATES 3B is approved by the NRC (Reference 12).

Burnup dependent hot rod calculations were performed with STRIKIN !!

to determine the initial fuel conditions which result in the peak cladding temperature (PCT).

The analysis was performed for the 0.0 Double Ended Guillotine at Pump Discharge (DEG/PD) break.

This break size was the limiting break size for the reference cycle.

In addition to the normal differences in fuel related parameters, the differences between Cycle 11 and the reference cycle that were analyzed for Cycle 11 8-1 c

were: (1) the use of GUARDIAN" grids for Batch N fuel assemblies and (2) a reduction of 260 gpm in Low Pressure Safety injection Pump (LPS!P) flow.

The reduction in LPSIP flow was previously evaluated for Unit 2 Cycle 9 (Reference 13).

A comparison of fuel related parameters for Cycle 11 and the reference cycle is shown in Table 8.2-1.

8.2.2 Results The burnup dependent hot rod calculations demonstrated that the burnup with the highest initial fuel stored energy results in the PCT. This occurs at a hot rod burnup of 1000 MWD /HTU.

Table 8.2 2 presents the ECCS performance analysis results for this limiting case.

For comparison purposes, the results for the reference cycle analysis are also presented therein.

As shown in Table 8.2-2, the PCT for the limiting LBLOCA for Cycle 11 is 2014'F.

This is well below the acceptance criterion value of 2200*F.

The maximum local and core wide cladding oxidation percentages of 4.65% and <0.51%

are also well below their respective acceptance criteria values of 17% and 1%.

A list of the significant parameters displayed graphically (Figures 8.21 through B.2-14) for the limiting case is presented in Table 8.2-3.

High burnup fuel conditions result in a maximum cladding temperature of 2002*F, 12*F below the limiting case presented above.

The maximum local cladding oxidation percentage of 7.65% for the high burnup fuel conditions is well below the acceptance criterion value of 17%.

A review of the effects of initial operating conditions on the results for Cycle 11 was performed.

It was determined that over the range of initial operating conditions, as specified in the Technical Specifications, operation of Cycle 11 at a PLHGR of 15.5 kw/ft is acceptable.

8.3 SMALL BREAK LOSS-0F-COOLANT ACCIDENT 8.3.1 Method of Analysis The SBLOCA ECCS performance analysis for Calvert Cliffs Unit 1 Cycle 11 was performed using the C E SBLOCA Evaluation Model (Reference 14).

The SBLOCA Evaluation Model is approved by the NRC (Reference 15).

The SBLOCA reference cycle analysis for Cycle 11 is the analysis for Unit 1 Cycle 8 (Reference 16).

The method of analysis for Cycle 11 is identical to that used for the reference cycle.

Blowdown hydraulic calculations were performed using the CEFLASH 4AS computer code (Reference 17).

Hot rod cladding temperature and cladding oxidation calculations were calculated using the STRIKIN-ll (Reference 7) and PARCH (Reference 18) computer codes.

a The analysis was performed for the 0.1 ft break in the pump discharge leg.

This is the break size that was analyzed in the reference cycle analysis and is the limiting SBLOCA for Calvert Cliffs Units 1 and 2.

The differences between Cycle 11 and the 82

- _ - _ _ =

o reference cycle that were analyzed for Cycle 11 tere:

(1) an 500 tubes per steam generator and (2) the use of GUARDIAN, tubes to increase in the maximum number of plugged steam generator " grids.

Significant core and system parameters used in the Cycle 11 and l

reference cycle analyses are listed in Table 8.3 1.

8.3.2 Results The results of the ECCS performance analysis for the 0.1 ft* break are summarized in Table 8.3 2.

For comparison purposes, the results for-the reference cycle analysis are also presented therein.

The

. PCT is 1991'r and the maximum local cladding oxidation percentage is 7.19% as compared to the acceptance criteria values of 2200'F and 17%, 'oJpectively.

The core wide cladding oxidation percentage (conservatively represented by the rod average oxidation percentage of the hot rod) is less than 0.875% as compared to the criterSn value of 1%.

The important transient parameters which are plotted as a function of time (Figures 8.31 through 8.3-8) are listed in Table 8.3 3.

8.4 CONCLUSION

The PCT for the limiting LBLOCA was calculated to be 2014'F which is well below the acceptance criterion value of 2200'F.

The PCT for the limiting SBLOCA was calculated to be 1991*F.

This is 23'F less than the PCT for the limiting LBLOCA and well below the acceptance criteria value of 2200'F.

The maximum local and core wide cladding oxidation percentages were also determined to be well below the acceptance criteria.

Therefore, operation of Unit 1 Cycle 11 at a PLHGR of 15.5 kw/f t and a power level of 2754 MWT (102% of 700 MWT) is in conformance with the 10CFR50.46 ECCS acceptance criteria.

8-3

.1

TABLE 8.2-1 CALVERT CLIFFS UNIT 1 CYCLE 11 LBLOCA ANALYSIS FUEL RELATED PARAMETERS COMPARED TO THE REFERENCE CYCLE (UNIT 1 CYCLE 10)

Unit 1 Unit 1 fuel Parameter Cycle 10 Cycle 11 Reactor Power Level (102% of Nominal), WT 2754 2754 Average Linear Heat Rate (102% of Nominal), kw/ft 6.45 6.58 Hot Rod Peak Linear Heat Generation Rate, kw/ft 15.5 15.5 Hot Assembly Peak Linear Heat Generation 13.68 13.81 Rate, kw/ft Cap Conductance at PLHGR, BTU /hr-ft' 'F(O 2105 2040 Fuel Centerline Temperature at PLHGR, 'F(4 3633 3621 Fuel Average Temperature at PLHGR, 'F(4 2207 2206 Hot Rod Gas Pressure, psia (U 1209 1227 Hot Rod Average Burnup, MWD /HTU(4 1000 1000 (1) STRIKIN II initial fuel rod parameter at burnup which yields the PCT.

8-4

-n.

~

~ -,

w

TABLE 8.2 2 CALVERT Cliffs UNIT 1 CYCLE 11

SUMMARY

Of ECCS PERf0RMNCE ANALYSIS RESULTS FOR THE LlHITING LBLOCA COMPARED TO THE REFERENCE CYCLE (UNIT 1 CYCLE 10)

Unit 1 Unit 1 Parameter Cycle 10 Cycle 11 Hot Rod Average Burnup, 1000 1000 MWD /MTU Peak Cladding Temperature 1983 2014 (PCT),'f Time of PCT, seconds 247 247 Time of Cladding Rupture, 25.7 26.2 Seconds Maximum Local Cladding 4.14 4.65 0xidation, %

Maximum Core Wide Cladding

<0.51

<0.51 0xidation, %

8-5 l

i

TABLE 8.2-3 CALVERT CLIFFS UNIT 1 CYCLE 11 VARIABLES PLOTTED AS A FUNCTION OF TIME FOR THE LlHITING LBLOCA Variable floure Numbat Core Power 8.2 1 Pressure in Center Hot Assembly Node 8.2-2 Leak Flow Rate 8.2-3 Hot Assembly Flow Rate -(below hot spot) 8.2-4 Hot 4AssemblyFlowRate(abovehotspot) 8.2-5 Hot Assembly Quality 8.2-6 Containment Pressure 8.2-7 Mass. Added to Core During Reflood 8.2-8 Peak Cladding Temperature 8.2 Gap Conductance at Location of PCT 8.2-10 Maximum Local Cladding Oxidation 8.2-11

-Temperature of Fuel Centerline.

8.2-12 4-Fuel Average,-Cladding and Coolant at' Location of PCT Heat Transfer Coefficient at Location-8.2 of PCTL Hot Rod Internal Gas Pressure 8.2-14 8-6

Table 8.3-1 CALVERT CLIFFS UNIT 1 CYCLE 11 SBLOCA ANALYSIS SIGNIFICANT CORE AND SYSTEM PARAMETERS COMPARED TO THE REFERENCE CYCLE (UNIT 1 CYCLE 8)

Unit 1 Unit 1 Parameter Cycle 8 Cycle 11 Reactor Power Level (102% of Nominal), MWT 2754 2754 Average Linear Heat Rate (102% of Nominal), kw/f t 6.37 6.58 Hot Rod Peak Linear Heat Generation Rate, kw/ft 15.5 15.5 Axial Shape Index (Including Uncertainty), ASIU

-0.16

-0.16 Low Pressurizer Pressure Reactor Trip Setpoint, psia 1736"'

1736 0

SIAS Low Pressurizer Pressure Setpoint, psia 1586 )

1586 tr O>

500 Maximum Number of Plugged Steam Generator Tubes per Steam Generator Charging Pump Flow (Minimum) Delivered to Intact 13 13 Pump Discharge Leg, gpm

"'This is the correct value for Unit 1 Cycle 8.

Reference 16 inadvertently listed 1741 psia for reactor trip and 1591 psia for SIAS.

This value was increased to 150 tubes per steam generator for subsequent reloads.

8-7

Table 8.3-2 CALVERI CLIFFS UNIT 1 CYCLE 11

SUMMARY

OF ECCS PERFORMANCE ANALYS,15 RESULTS FOR THE LlHITING SBLOCA (0.1 ft BREAK)

COMPARED TO THE REFERENCE CYCLE (UNIT 1 CYCLE 8)

Unit 1 Unit 1 Parameter Cycle 8 Cycle 11 Peak Cladding Temperature (PCT), 'T 1877 1991 Time of PCT, seconds 1650 1783 Elevation of PCT (from Bottom of Core), ft 10.8 10.8 Maximum Local Cladding Oxidation, %

4.92 7.19 Maximum Core Wide Cladding 0xidation, %

<0.632

<0.875 I

l l

i 8-8 l

1 Table 8.3 3 CALVERT CLIFFS UNIT 1 CYCLE 11 VARIABLES PLOTTED AS A FUNCTION OF TlHE FOR THE LlHITING SBLOCA Variable Fiaure Nuthgr Core Power 8.3-1 Inner Vessel Pressure 8.3-2 Leak Flow Rate 8.3 3 Inner Vassel Inlet flow Rate 8.3 4 Inner Vessel Two-Phase Mixture Height 8.3 5 Heat Transfer Coefficient at Location of PCT 8.3 6 Coolant Temperature at Location of PCT 8.3-7 Peak Cladding Temperature 8.3-8 1

l l

I l

l l

8-9

Figure 8.2-1 CALVERT CLIFFS UNIT 1 CYCLE 11 l

0,6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER i

1.2 i

1.0-O H 0.8 :

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TIME, SECONDS I

8-10

Figure 8.2 2 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NODE 2400 2000 :.

<1600 :

Ui o.

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x 0,,,,,,,,,5 10 15 20 25 0

TIME, SECONDS I

l 8-11 L'

i Figure 8.2 3 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG LEAK FLOW RATE Reactor Coolant Pump Side

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- 100000 I o

ta

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TIME, SECONDS 8-12 n

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_c Figure 8.2-4 CALVERT CLIFFS UNIT 1 CYCLE 11

- 0.6 DOUBLE ENDED GUILLOTINE ilREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY FLOW RATE (BELOW HOT SPOT) c 30 20 o

to m

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Floure 8.2-5 l

CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG i

HOT ASSEMBLY FLOW RATE (ABOVE HOT SPOT) l

{

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Figure 8.2 6 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY OUALITY Node 13, Below Hottest Region Node 13, At Hottest Region Node 15, Above Hottest Region 1.0 p

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Figure 8.2-7

.CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE GO; I

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Figuro 8.2-8 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD Time, see Reflood Rate 0.0 - 8.7 2.56 i

8.7 - 76.2 1.14 76.2-500 0.768 120000_

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-TIME AFTER CONTACT, SECONDS 8-17

Figure 8.2-9 CALVERT CLIFFS UNIT 1 CYCLE 11 i

0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PEAK CLADDING TEMPERATURE 2400_

2000~

^-

5 s

b-o 1600 7 wa u.i t

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r e

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t Figuro 8.2-10 t

CALVERT CLIFFS UNIT 1 CYCLE 11

'O,6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG GAP CONDUCTANCE AT LOCATION OF PCT v

i I

t 1800 o

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8-19

e Figure 8.211 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MAXIMUM LOCAL CLADDING OXIDATION 18_

15

~

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o Figuro 8.212 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG TEMPERATURE OF FUEL CENTERLINE, FUEL AVERAGE, CLADDING AND COOLANT AT LOCATION OF PCT

- - - Fuel Centerline Fuel Average

-- - - - Cladding

......... Coolant 30001 c

t t

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200 300 400 500 TIME, SECONDS 8-21

Figure 8.2-13 CALVERT CLIFFS UNIT 1 CYCLE 11 i

0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG l

- HEATTRANSFER COEFFICIENT AT LOCATION OF PCT t

i i

I 180 L

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o Figure 8.2-14 CALVERT CLIFFS UNIT 1 CYCLE 11 0.6 DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOY ROD INTERNAL GAS PRESSURE INITIAL PRESSURE = 1227 PSIA CLADDING RUPTURES AT 26.2 SEC 1800_

1500 :

1200

-4m o-Ld 900

~

s, m

m tu oc n_

600 :

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~

0'''20 40 60 80 100 O

TIME, SECONDS 8-23

3..,.

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Figure 8.3-1

CALVERT CLIFFS UNIT 1-CYCLE 11-0.1 FT2 BREAK IN PUMP DISCHARGE LEG CORE POWER 1.2 4

1.0 ~

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Figure 8,3-2 CALVERT CLIFFS UNIT 1 CYCLE 11 0.1 FT2 BREAK IN PUMP DISCHARGE LEG INNER VESSEL PRESSURE 2400 -

4 2000 !

F f

<1600 ;

w c.

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TIME, SECONDS 8-25

41 Figure 8.3-3

~ CALVERT CLIFFS UNIT 1 CYCLE 11

^

0.'1-FT2 BREAK IN PUMP DISCHARGE LEG LEAK FLOW RATE 2400_

.2000 :

o-w-

m 1600 N

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TIME, SdCONDS 8-26

Figure 8.3-4 cat., dRT CLIFFS UNIT 1-CYCLE'11 0.1 FT2 BREAK IN PUMP DISCHARGE LEG

. lNNER VESSEL INLET FLOW RATE

'35000;.

l 28000 owm 21000 N

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8-27

o Figure 8.3-5 CALVERT CLIFFS UNIT 1 CYCLE 11 0.1 FT2 BREAK IN PUMP DISCHARGE LEG INNER VESSEL TWO-PHASE MIXTURE HEIGHT 60

)

-50 :

40 i t-J I

5:2 30

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BO T T O fA OF C O R E' Of' ' ' ' ' ' ' 'g' ' ' ' ' 'q' Q ' ' ' ' ' 3 Q ' ' ' ' 'g g ' ' ' ' 'g g TIME, SECONDS 8-28

Figuro 8.3-6

.CALVERT. CLIFFS UNIT 1 CYCLE 1'i O.1' FT2 BREAK IN PUMP DISCHARGE LEG HEATTRANSFER COEFFICIENT AT LOCATION OF PCT 100000 i

10000 y

'w N.

lZ f

,1000 I

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100

.g.x

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10 tff f f fff f fffffff ff I ffffffif Ifff ff fff ffff f f fif 0-500 1000 1500 2000 2500 TIME, SECONDS 8-29 i

Figure 8.3-7 CALVERT CLIFFS UNIT 1 CYCLE 11 0.1 FT2 BREAK IN PUMP DISCHARGE LEG COOLANT TEMPERATURE AT LOCATION OF PCT 1200_

i

[

]

1000 ::

u_

8 800 ~

f i

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400

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200 :

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'$ BOO 0

TIME, SECONDS 8-30

Figure 8.3-8 CALVERT CLIFFS UNIT 1 CYCLE.11 0.1 FT2 BREAK IN PUMP DISCHARGE LEG

^

PEAK CLADDING TEMPERATURE 2200_::

E A,

1900 f

Lt.

.h d1600 [

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Q 1300 e.

w a.

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700

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40 0 ' ' ' ' ' ' ' '50 0 1000

-1500 2000 2500 O

TIME, SECONDS 8-31

3

_9.0~I TECHNICAL SPECIFICATIONS Technical-Specification changes are being requested in order to make the Calvert-Cliffs Unit 1 Technical Specifications consistent with either-the> reference cycle (Reference 1) analyses which have been

+

verified as-applicable to Unit 1 Cycle 11 or the analyses presented in th_is: License Submittal. - These requested changes are presented in this section in three sets.

The first set -involves only those changes which have already been implemented in the Unit 2 Technical _ Specifications via the requests contained in the Unit 2 Cycle 9 License Submittal (Reference 1).

A summary of.- these changes is presented in Table 9-1 in the form of:

1) an action statement - for each change, 2) the reasons for each-change, and-.3) a reference to -the supporting Unit 1 Cycle-11 analyses. which continue to demonstrate. acceptable safety analyses results _ for each change (The original analyses supporting these requests were performed for Unit 2 Cycle 9 with one exception).

Following Table 9-1 the existing Unit 1 Technical Specification page with the intended modification is provided for each rechnical

. Specification for which-a change is being requested.

Some of' the _ Technical Specification changes discussed in Table 91 are being requested to accommodate the use of the new on-line incore LC0. moni toring system consisting of the CECOR 3.3/BASSS computer network. - This new monitoring-system was approved for Unit 2 Cycle 9 (reference cycle) and is presently being used for that cycle.

See Section 5.2 and Appendix B of Referance-1 for a description of this new system.

The changes in the first set are identical to tnose proposed and approved for Unit 2 Cycle 9 with two exceptions.

First. the changes

-for Unit 2 Cycle 9 concerning..the implementation of CECOR 3.3/BASSS supported ' the alternative of continuing to monitor with the octant solution. code, INCA.

The changes proposed for Unit 1 Cycle 11 no longer support that alternative.

Second, several of the Unit 1 Technical. Specifications being changed in the first set contain the exemption on alignment / operation of the center CEA which was added during ~ Cycle 10 operation due to concerns about swelling of the

" weak" Al 0 fingers of that CEA (see Section 4.1.3).-

The Unit 2 Technical dpecifications never contained such an exemption.

a The changes proposed.for Cycle 11 include the_ removal of this exemption.

The second set consists of most of the changes which are associated with the removal of the special exemption granted for Cycle 10 that permits continued operation should the center CEA become inoperable.

The remaining changes needed to remove this. special exemption are contained in the first set.- This exemption is being removed because the center CEA is being replaced (see Section 4.1.3).

A = summary of the second set of changes-is contained in Table 9-2 in

.the form of: 1) a listing of each Technical Specification, and 2) an action statement, explanation, and supporting statement _ which applies to all listed Technical Specifications.

Following Table 9-2 the existing Technical Specification page with the intended 9-1

i modification is provided for each-Technical Specification for which a change is being requested.

The third set consists of two requested changes. The Unit 1 Cycle 11 analyses provide the original support for these new changes.

A summary of these changes is presented in Table 9-3 in the form of:

1) an action statement for each change, 2) the reason for each change, and 3) a reference to the Unit 1 Cycle 11 analysis or analyses which support each change.

Following ~lable 9-3 the existing Technical Specification page with the intended modification is provided for each Technical Specification for which a change is being requested.

l i

1 9-2

-TABLE 9-1

'CALVERT CLIFFS UillT 1 CYCLE 11 TEClifilCAL SPECIFICATION CilAllGES ALREADY-lMPLEMEllTED FOR UNIT 2

-Tech. Spec No, and Paae Action Exrdanat ion Support /Use-Figure 2.2-1 Modify Figure ~2.2-1, as The AOR on the negativa

' side was The setpoint analysis.

page 2-11 indicated, to reduce the reduced for Unit 2 Cy e a performed for Unit 1 Acceptable.0peration Region accommodate the Cycle 11 takes. credit (AOR) between 100% and 40%

Average. Linear

.~.a t e for this modification power on the negative ASI (CAlllGR) resu' 5:

in demonstrating side, such that the negative increased r

?- a that acceptable results for ASI limit at.100% power is are present

< nonth Unit 1 Cycle 11.

reduced from.20 to.18.

cycle.

Tht:

, m,so being made for Uni l' as it will have even im ar shims than Unit 2 Cy. o

-*n se note that ea this slight U4 cease in the number of shims d70s not require further tightening o the AOR as sufficient r

margin is available to cover the corresponding, small increase in CAL!IGR.

3.1.3.1 Increase f/, as inslicated, The f/ value was raised for Unit 2 The setpoint. analysis page 3/4 1-18 from 1.65 to 1.70.

Cycle 9 to accommodate the increased performed for Unit I radial peaking associated with a Cycle 11 supports this second 24-month cycle.

This change change.

is also being made for Unit 1 Cycle 11 as its radial peaking will be even slightly higher than that for Unit 2 Cycle 9.

Please note that there remains a sufficient differer.ce between the predicted F/

peak for Unit 1 C Unit 2 Cycle 9F,ycle11andthe Tech. Spec. value that a further increase in this value is not necessary for Unit 1 Cycle 11.

TABLE 9-1 (Cont'd)

CALVERT CLIFFS UNIT 1 CYCLE 11 TECliNICAL SPECIFICATION CHANGES ALREADY IMPLEMENTED FOR UNIT 2 Tech. Spec.

Exolanation Sucoort/Use tio. and Pace Action 3.1.3.1 Hodify the text, as The text concerning allowable The new network calculates page 3/4 1-18 indicated, to support the use time to realign a CEA is the allowable time to realign (cont'd) of BASSS.

being changed to accommodate a CEA. The Unit I Cycle 11 the use of the new CECOR setpoint analysis has been 3.3/BASSS network which performed to provide that computes allowable time network with proper automatically. The modified coefficients to make that text continues to support the calculation.

use of Tech. Spec. Fig. 3.1-3 when BASSS is unavailable.

{

The footnote and associated Section 4.1.3 discusses the Modify the text, as indicated, to remove the referencing are being removed replacement of the center exclusion concerning the as they are no longer needed CEA.

center CEA during Cycle 10.

due to the replacement of the center CEA.

3.1.3.1 Modify the text, as The text is being changed to The new network calculates page 3/4 1-19 indicated, to support the use recognize that the time the allowable time to align a allowance to realign a CEA CEA. The Unit 1 Cycle 11 of BASSS.

can come from either the new setpoint analysis has been CECOR 3.3/BASSS network or performed to provide that figure 3.1-3.

Figure 3.1-3 network with the proper would be employed if BASSS is coefficients to make that unavailable.

calculation.

Modify the text, as See the same change on Page See the same change on Page indicated, to remove the 3/4 1-18.

3/4 1-18.

exclusion concerning the center CEA.

TABLE 9-1 (Cont'd)

CALVERI CLIFFS UfflT.1 CYCLE 11 IECllillCAL SPECIFICATI0ft CilAtlGES ALREADY IMPLEMEtiTED FOR Ull!T 2 Tech. Spec.

~

flo. and Page Action Explanation Support /Use figure 3,1-3 Modify figure 3.1-3, as The f,' values of Figure 3.1-The setpoint analysis page 3/4 1-198 indicated, to increase the 3 were raised for Unit 2 performed for Unit 1 Cycle 11 F,' values.

Cycle 9, in conjunction with supports this change.

the. int.rease in the f ' value of Tecn. Spec. 3.1.3.}, to accommodate the increased radial peaking associated with a second 24-month cycle.

Similarly, since the F,'

value of Tech. Spec. 3.1.3.1 y

is also being raised for Unit

! Cycle 11 (as discussed ui previously), the corresponding changes in figure 3.1-3 are likewise being made for Unit I Cycle 11.

Figure 3.2-3b Modify Figure 3.2-3b, as This "f1" factor curve was lhe setpoint analysis takes page 3/4 2-4a indicated, to reduce the reduced for Unit 2 Cycle 9, credit for this modification Acceptable Value region, such partially to accommodate the in demonstrating acceptable that the F,,' value below increased Core Average Linear results for Unit 1 Cycle 11.

which 100% power operation is lleat Generation Rate (CALHGR) acceptable is reduced from resulting from the increased 1.54 to 1.50 number of B C shims in a 4

second 24-month cycle.

This change is being made for Unit 1 Cycle 11 for the same reason (see discussion concerning change in Figure 2.2-1).

=

~

1 TABLE 9-1,(Cont'd)-

CALVERT CLIFFS UNIT 1 CYCLE 11

'TECilNICAL SPECIFICATION CilANGES

ALRfADY IMPLEMENTED FOR UNIT 2

. Tech. Spec.'

No. a'nd Page' Action'

' Explanation Support /Use:

byafullcore,[powermapping-Eliminate the'e This change is being made to The'value of F calculated ;

3.2.2.1 calculating':f,,'quation for page3/42-6"

', as-

. accommodate the full' core

' indicated.

measured power; distribution code'already includes the.

which will be calculated by effect of azimuthal tilting.-.

the new CECOR 3.3 code.

Modify the text, as'

.The text concerning the The new network calculates '

.the allowed' THERM indicated,.to support the use reduction in' TilERMAL POWEft

'withregardtof,qlPOWER of BASSS.-

when the f limit is

.,The' Unit-exceeded.is being changed to

. i Cycle ll setpolnt analysis -

J y

accommodate the use of tha has computed the coefficients 1

o new CECOR 3.3/BASSS' networ k for that~ calculation.

which will compute the IIIERMAL POWER.1imi1-automatically.

Modify the text, as-See same change for Tech.

.See same change for Tech.

indicated, to remove the

.. Spec. 3.1.3.1.

Spec. 3.1.3.1.

exclusion concerning the-center CEA.

4.2.7.1.2 Modify the text, as lhe' text concerning the.

See first change for' Tech.

page 3/4 2-6 indicated, to support the use calculation of f,,' is being Spec. 3.2.2.1.

l of CECOR 3.3.

modified to accommodate the full core measured power distribution which will be calculated by the new CECOR-

-3.3 code.

4.2.2.1.3-.

Modify the text, as.

See change for Tech. Spec.

See first change for Tech.

page 3/4 24 indicated, to support the use 4.2.2.1.2.

Spec. 3.2.2.1.

of CECOR 3.3.

,=

TABLE _9-1 (Cont'd)

CALVERT CLIFFS UNIT 1 CYCLE 11 TECHNICAL SPECIFICATION CHANGES ALREADY IMPLEMENTED FOR UNIT 2 lech. Spec.

No. and Page Action Explanation Support /Use 4.2.2.1.3 Modify the text, as See same change for Tech.

See same change for Tech.

page 3/4 2-7 indicated, to remove the Spec. 3.1.3.1 Spec. 3.1.3.1.

(cont'd) exclusion concerning the center CEA.

4.2.2.1.4 Modify the text, as The text concerning the value See first change for Tech.

to be used in the Spec. 3.2.2.1.

of T,lation of F,/ is being page 3/4 2-7 indicated, to eliminate the calcu determination of T.

y eliminated as it is no longer needed due to the f

implementation of the CECOR 3.3 code.

u 4.2.2.2.2 Modify the text, as See change for Tech. Spec.

See first change for Tech.

page 3/4 2-8 indicated.

4.2.2.1.2.

Spec. 3.2.2.1.

4.2.2.2.3 Modify the text, as See change for Tech. Spec.

See first change for Tech.

page 3/4 2-8 indicated.

4.2.2.1.2.

Spec. 3.2.2.1.

Modify the text, as See same change for Tech.

See same change for Tech.

' indicated, to remove the Spec. 3.1.3.1 Spec. 3.1.3.1.

exclusion concerning the center CEA.

4.2.2.2.4 Modify the text, as See change for Tech. Spec.

See first change for Tech.

page 3/4 2-8 indicated.

4.2.2.1.4.

Spec. 3.2.2.1.

Eliminate the e This change is being made to The value of F,' calculated calculatingF/,quationfor 3.2.3 as accommodate the full core by a full core power mapping page 3/4 2-9 indicated.

measured power distribution code already includes the which will be calculated by effect of azimuthal tilting the new CECOR 3.3 code.

TABLE 9'-1'(Cont'd)

CALVERT CLIFFS UNIT >l CYCLE 11-TECilNICAL SPECIFICATION CilANGES-ALREADY IMPLEMENTED FOR UNIT 2 Tech. Spec.

No. and Pace

' Action Explanation Support /Use 3.2.3 Increase F' as indicated, TheLF ' value was raised for The setpoint analysis page 3/4 2-9

'froml.65Io,1.70.

Uniti Cycle 9 to accomodate. performed for Unit 1 Cycle 11- :

(cont'd) the increased radial peaking supports this change.

associated with a'second 24-month cycle. As noted; earlier, this' change is also being made' for' Unit.1 Cycle -

11 for the'same reason (see j'

discussion concerning change to.. Tech. Spec. 3.1.3.1).

i Modify the text, as The text concerning the The new network calculates indicated, to support the use reduction in TilERMAL POWER theallowedTHERyLPOWER-i of BASSS.

when the F ' limit is with regard to F.. The Unit l

exceeded is being changed to 1 Cycle 11 setpoInt analysis accommodate the use of the has computed the coefficients i

new CECOR 3.3/BASSS network for that calculation.

which will compute the TilERMAL POWER 1imit automatically.

r l

Modify the text, as See same change for Tech.

See same change for Tech.

i

-indicated, to remove the Spec. 3.1.3.1.

Spec. 3.1.3.1.

exclusion concerning the l

-center CEA.

-t 4.2.3.2 Modify the. text, as The text concerning the-See first change for. Tech.

l page 3/4 2-9

' indicated, to support the use calculation of F[ is being Spec. 3.2.3.-

of CECOR.3.3.

modified to' accommodate the full core measured power distribution which will be calculated by the new CECOR.

3.3 code.

i i

L;

[

II TABLE 9-1.(Cont'd)

CALVERT CLIFFS UNIT 1 CYCLE 11

- TECilNICAL SPECIFICATION CHANGES ALREADY IMPLEMENTED FOR UNIT'2

' Tech. Spec.

No. and Pace Action Explanation Support /Use 4.2.3.3:

Modify the text, as See change for Tech. Spec.

See first change for Tech.

page 3/4 2-10 indicated, to support the use 4.2.3.2.

Spec. 3.2.3..

of~ CECOR 3.3.

Modify the text,-as See same change for' Tech.

See!same change for Tech.

indicated, to remove the Spec. 3.1.3.1.

Spec. 3.1.3.1.

exclusion concerning'the center CEA.

4.2.3.4-Modify the text, as The text concerning the value-See first' change for Tech.

to be used

' Spec. 3.2.3.

of T,lation of F,jn the page 3/4 2-10 indicated -to eliminate the

?

determination of T*.

calcu

'Is being eliminated as it is no longer needed due to the implementation of the CECOR 3.3 code.

Figure 3.2-3c Modify figure 3.2-3c, as The F ' limit curve was The setpoint analysis page 3/4 2-10a indicated, to increase the raiseb for Unit 2 Cycle 9, in performed for Unit 1 Cycle 11 F/ values.

conjunction with the increase supports this change.

in the F/ value, to accommodate the increased radial peaking associated with a second 24-month cycle.

Again, as noted earlier, this change is also being made for Unit 1 Cycle 11 for the same reason (see discussion concerning change to Tech.

Spec.~3.1.3.1).

TABLE 9-1 (Cont'd)

CALVERT CLIFFS UNIT 1 CYCLE 11 TECHNICAL SPECIFICATION CHANGES ALREADY IMPLEMENTED FOR UNIT 2 Tech. Spec.

No. and Paae Action Explanation Support /Use 3.2.5 Change " Core Power" to The term " Core Power" was This change is supported by page 3/4 2-13

" THERMAL POWER", as changed to " THERMAL POWER" the logic of consistency.

indicated.

for Unit 2 Cycle 9 to be consistent with all other Tech. Specs. This change is being made for Unit 1 Cycle 11 for the same reason.

4.2.5.3 (new)

Add the new Tech. Spec.

This new surveillance Tech.

The setpoint analysis y

page 3/4 2-13 4.2.5.3, as indicated.

Spec. is being added to performed for Unit 1 Cycle 11 accommodate the use of the supports the use of the new new CECOR 3.3/BASSS network.

CECOR 3.3/BASSS network.

Table 3.2-1 Change " Core Power" to See Change for Tech. Spec.

See change for Tech. Spec.

page 3/4 2-14

" THERMAL POWER", as 3.2.5.

3.2.5.

indicated.

B 3/4.1.3 Modify the text, as See second change for Tech.

See second change for Tech.

page B 3/4 1-4 indicated.

Spec. 3.1.3.1.

Spec. 3.1.3.1.

B 3/4.7.1.2 Increase the maximum The maximum Auxiliary An evaluation of the effect page B 3/4 7-2a Auxiliary Feedwater Flow, as Feedwater Flow was increased of this increase in flow on indicated, from 1300 to 1550 for Unit 2 Cycle 9 to the Steam Line Rupture event gpm.

accommodate potentially was made for Unit 2 Cycle 9 higher flows. This change is (see Section 7.0 of Reference also being made for Unit 1 1). That evaluatfon Cycle 11 for the same reason.

determined that previously reported results remained bounding. The Unit 1 Cycle 11 analyses of Secticn 7.0 continue to show acceptable results for this change.

"[

e..

TABLE 9-1 (Cont'.d)-

CALVERT CLIFFS UNIT 1 CYCLE'll-TECHNICAL SPECIFICATION CHANGES I

ALREADY IMPLEMENTED FOR UNIT 2'

~

Tech. Spec.

'No. and Pace Action Explanation

.Succort/Use 1

5'. 3.1

. Increase the maximum

' The specification of the -

All physics, performance, page 5-4 enrichment for.a reload core,- enrichment limit for reload safety,;and setpoint. analyses as' indicated, from'4.1 to fuel assemblies was. Increased perfonned for Unit 1 Cycle 11 4.35 w/o U-235.

for Unit 2 Cycle 9 to permit took the higher enrichment of:

the use of higher enrichment the fresh' fuel into account-fuel. This change..is also (see Section 3.0.for.a being made for Unit 1 Cycle description of the Unit 1' 11 for the'same reason.

Cycle 11 core).

Y' 4

i i

i c,

2-TABLE 9-2

.CALVERT CLIFFS UNIT 1 CYCLE 11 TECHNICAL SPECIFICATION CHANGES TO' REMOVE CENTER CEA EXEMPTION (EXCLUDING THOSE ALREADY.. COVERED'IN, TABLE 9-1)

. Tech. Spec. No. and Pace-Action

' Explanation' Support /Use-4.1.1.1.1 (Page 3/4.1-1)'

t 4.1.1.2

.(Page 3/4 1-3)'..

4.1.3.1.I' (Page 3/4 1-17)

3. l'.3.1 4.1.3.1.2 -(Page 3/4 1-19A) 4.1.3.1.3 3.1.3.3,(Pages 3/4 1-21 and 3/4 1-22) 4' l.3.3.l}(Page ~ 3/4 'l-22) f 4.1.3.3.2)(Page 3/4 1-23) a U

3.1.3.4; '-

(Modifythetext,as

'The footnote and associated-Section 4.1.3 discusses lindicated, to remove referencing are being removed the replacement of the -

(Page3/41-23)htheexclusion-4.1.3.4.

(Page'3/4 1-24) as they are no longer needed center CEA.

4.1.3.5 t

3.1.3.6

.(Pages 3/4 1-25)i concerning the center due to the replacement of the and 3/4 1-26) fEA during Cycle 10.

center CFA.

4.1.3.6 (Page'3/4 1-26) 3.10.l' ' (Page 3/4 2-1) 4.2.1.3 4.10.1.I'-(Page 3/4 10-1) )

l 4.10.1.2,

/

l i

t i

I l

a

~

~ * "

TABLE 9-3 CALVERT CLIFFS UNIT 1 CYCLE 11 NEW TECHNICAL SPECIFICATION CHANGES Tech. Spec..

Eo. and Paae Action Explanation Support /Use Figure 3.1-lb Modify the figure, as The. Shutdown Margin is being The results of the revised page 3/4 1-2a indicated, to reduce Shutdown decreased at EOC from 5.0% ak/k Steam Line Rupture analysis Margin.

to 4.5% Ak/k to accommodate the (Section.7.0) were less reduced scram worth available limiting than those:

for Unit 1 Cycle 11 due previously reported. Since principally to its low-fluence Lthe results were bounded, fuel management.

text describing this analysis was not included in Section 7.0.

Y' S

Figure 3.2-1 Modify the figure, as Cycle lengths have been All safety /setpoint/

fndicated.f o eliminate increased beginning with the performance analyses support t

page 3/4 2-3 specific cycle times.

implementation of 24-month a constant value lof Allowable cycles for Cycle 10, causing Peak Linear Heat Rate the cycle times on this' figure throughout the cycle.

to be out-of-date. Since time-in-cycle dependent Allowable Peak Linear Heat Rates are not being utilized at this time, specific cycle times are not required.

v

-r

-10.0 STARTUP' TESTING The' startup -testing program proposed for Unit-1 Cycle 11 is identical'to the program used for the reference cycle-(Unit 2 Cycle 9, Reference 1).

The potential effect of low fluence fuel management upon that part of. the --startup testing program concerned-with the confirmation of core symetry was evaluated.

That evaluation determined that the 4

+

core symmetry verification procedure recomended-for Unit 2 Cycle 9, which consisted of full core power distribution monitoring along with - minimal CEA symetry testing, remains applicable to Unit 1 Cycle 11.

1 6

I 1

10-1 t--,

A:

1;'

11.0l REFERENCES References -' Section 1

-1.

Letter, G.

C.

Creel (BG&E) to Document Control Desk (NRC),

"Calvert~ Cliffs Nuclear - Power P'iant' Unit No. 2; Docket No.

50-318, Request for Amtnhen, Unit 2 Ninth Cycle License

~ Application," February 7, 1989.

2.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit Nos. I and 2; Docket.

Nos. 50-317_ and 50-318, Requast for Amendment, Unit One Tenth-j Cycle License Application; Unit Two Axial Shape Index Region -

Enlargement," February 12, 1988.

3.

Letter, S.

A.

McNeil (NRC) to G.

C.

Creel (BG&E),

Docket No. 50-318, '(Safety Evaluation Report of Ninth Cycle

. License Application), January 10, 1990.

4.

CENPD-269 P, Revision 1-P,

" Extended 8urnup Operation of Combustion Engineering PWR Fuel," July 1984.

-5.

. Letter,- E. J.

Butcher (NRC) to A.

E. Lundvall, Jr. (BG&E),

Docket No1, 50-317 and 50 318, " Safety Evaluation for Topical Report CENPD-269-P, Revision 1-P," (Extended Burnup Operation),

October 10, 1985.

6.

-CEN-382(B)-P, " Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD /kg: for Calvert Cliffs Units 1 and 2,"

January 1989.

4 i

~

)

11-1

a n

Reference - Section 2 1.a. Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

. "Calvert Cliffs Nuclear Power Plant Unit No.

1, Docket No.

50-317 Report of Startup Testing for Unit 1 Cycle 10,"

September 9, 1988,

b. Letter, J. A. Mihalcik (BGLE) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No.

50-317, Correction to Report of Startup Testing for Unit 1 Cycle 10," NEU 88-308, October 13, 1988, 11-2

a Referpnce - Section 3

~1.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Hus;. lear Power Plant Unit Nos. I and 2; Docket Nos. 50-317 and 50-318, Request for Amendment, Unit One Tenth Cycle License Application; Unit Two Axial Shape Index Region Enlargement," February 12, 1988.

11-3

)

T*

Q Peferences - Section 4 1.-

Letter, G.

C. Creel (BG&E) to Document' Control Desk (NRC),

Calvert Cliffs Nuclear-Power Plant Unit No. 2; Docket No. 50-318,. Request for Amendment,. Unit 2 Ninth Cycle License Application," February 7, 1989.

~2.

Calvert Cliffs Nuclear Power Plants Units'1 and 2 Updated Final

~

Safety Analysis Report, Chapter 3, Revision 11, January 1991.

3.

CENPD-269-P, Revision 1-P,

" Extended Burnup Operation o f --

Combustion ' Engineering' PWR Fuel,"' July 1984.

4. __

July 1974.-

CENPD-139-P-A, "C-E Fuel Evaluation-Model Topical Report,"

5.

CEN-161(B)-P-A, " Improvements to Fuel Evaluation Model," August 1989.=

6.

CEN-161(B)-P, Supplement 1-P, " Improvements to Fuel Evaluation-.

Model," April.-1985.

7. -

CEN-372-P-A, " Fuel Rod Maximum Allowable Gas Pressure," May 1990.

'8.

Letter, S.-A. McNeil (NRC) to J. A. Tiernan (BG&E), Docket Nos.-

50-317 and 50-318, " Safety Evaluation of Topical Report CEN-M1 (B)-P Supplement 1-P, ' Improvements - to Fuel Evaluation Model',"

February 4, 1987.

f

,_+

11-4

?

h-t.

References - Section 5-1.

Letter, G.

C. Creel (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant' Unit No. 2; Docket No. 50-318.

Request for Amendment, Unit 2 Ninth Cycle License Appt; cation," February 7, 1989.-

2.

CENPD-266-P-A, "The ROCS ' and DIT Computer Code for Nuclear Design," April 1983.

3.

CENPD-275-P-A, "C-E Methodology _ for Core Designs Containing

.Gadolinia - Urania Burnable Absorbers," May 1988.

4.

CENPD-153-P, - Rev.

1-P-A,

" Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed In-

-Core Detector System," May 1980.

l 11-5 l

4

'I j

y

' References - Section 6 1.-

CENPD-161-P-A, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986.

L 2.

- CENPD-162-P-A,." Critical Heat ~ Flux Correlation for C-E Fuel Assemblies.with Standard Spacer Grids Part 2, Uniform Axial-Power Distribution,' April 1975.

3.

_ CENPD-206-P-A,. " TORC Code, Verification and Simplified Modeling

= Methods," June 1981.

1 4..

CEN-191(B)-P, 'CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981.

.5. -

Letter, D.' H..Jaffe (NRC) to A. E.

Lundvall, Jr.

(BG&E),

l Regarding Unit 1 Cycle-6 License Approval (Amendment #71 to

l DPR-53 and SER), June 24, 1982.

j 6.

CEN-348(B)-P,

" Extended Statistical Combination of i

Uncertainties," January 1987.

7.

Letter, S. A. McNeil -(NRC) to J. A. Tiernan (BG&E), Docket Nos..

50-317 and 318, " Safety Evaluation of Topical Report CEN-348(B)-P,-' Extended-Statistical Combination of Uncertainties',"

October 21, 1987.

! 8.

CENPD-225-P-A, " Fuel and Poison Rod -Bowing," June _1983.

9.

Letter, G.

C.

Creel (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318, -Request for Amendment, Unit Two Ninth Cycle License Application," February 7, 1989.

t:'

?

i 11-6

c

'r B.tffCfnces - Section 7 1.

Letter, G.

C. Creel (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit No.

Docket No.

50-318, Request for Amendment, Unit 2 Ninth Cycle License Applic6 tion,' February 7,1989.

2.

CEN-124[B).P.

"Stettstical Combination of Uncertainties Methodo:ogy Part 1: C-E Calculated Local Power Density and Thermal Margin /l,w Pressure LSSS for Calvert Cliffs Units I and

I,' December 19lS.

3.

CEN-174(B)-P,

" Statistical ComLination of Uncertainties Methodology Part 2:

Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units 1 & 2,' January 1980.

4.

CEN-124(B)-P,

' Statistical Combination of Uncertainties Methodology Part 3:

C-E-Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Calvert Cliffs Units 1 and 2,* March 1980.

5.

' N-348(B)-P,

" Extended Statistical Combination of Uncertainties," January 1987.

6.

Letter, D. H.

Jaffe (NRC) to A. E.

Lundvall, Jr.

(BG&E),

Reg 6rding Unit 1 Cycle 6 License Approval (Amendment #71 to DPR-53 and SER), June 24, 1982.

7.

Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E), Docket Nos.

50-317 and 50-318

" Safety Evaluation of Topical Report CEN-348(B)-P, ' Extended Statistical Combination of Uncertainties','

October 21, 1987.

l l

11-7

.-,.,,e

~, -

p.

r References - Section 8 i

1.

Acceptance Criteria for Emergcacy Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3, Friday, January 4,1974.

2.

CENPD 132, Supplement 3 P A, ' Calculative Methods for the C E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS " June 1985.

3.

Letter, D. M. Crutchfield (NRC) to A. E. Scherer (C E), ' Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related

. Licensing Topical Reports," July 31, 1986.

4.

Letter, J. A. Tiernan (BG1E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit Nas. I and 2; Docket Nos. 50 317 and 50 318, Request for Amendment, Unit one Tenth Cycle Application; Unit Two Axial Shape Index Region Enlargement," february 12. 1988.

5.

CENPD 133, Supplement 5,

'CEFLASH 4A, A FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis,' June 1985.

6.

CENPD 134, Supplement 2

'COMPERC-II, A Program for Emergency Refill Reflood of the Core," June 1985.

7.

CENPD 135, Supplement 5 P, 'STRIKIN-!!, A Cylindrical Gtometry Fuel Rod Heat Transfer Program,' April 1977.

8.

Letter, A. E. Scherer (C-E) to J. R. Hiller (NRC)

LD 81095, Enclosure 1 P,

'C E ECCS Evaluation Model Flow Blockage Analysis," December 15, 1981.

9.

CENPD 139 P A, 'C-E fuel Evaluation Model Topical Report,* July 1974.

10.

CEN-161(B)-P-A, ' Improvements to Fuel Evaluation Model," August 1989.

11.

CEN-161(B)-P, Supplement 1-P, ' Improvement to fuel Evaluation Model," April 1986.

12.

Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E), Docket Nos, 50 3)/ and 50 318, " Safety Evaluation of Topical Report CEN-161(B) D Supplement 1-P,

' Improvements to Fuel Evaluation Model',' February 4,1987.

13.

Letter, G.

C.

Creel (BG&E) to Document Control _ Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit No. 2: Docket No. 50-318 Request for Amendment, Unit 2 Ninth Cycle License Application," February 7, 1989.

11-8 w

h References - Section 8 (Cont'd)

14. CENPD 137, Supplement 1 P,

' Calculative Methods for the C E Small Break LOCA Evaluation Model," January 1977.

15.

Letter, K. Kniel (NRC) to A. E. Scherer (C-E), " Evaluation of Topical Reports CENPD-133, Supplement 3P and CENPD 137, Supplement 1-P,' September 27, 1977.

16.

Letter, A.

E. Lundvall, Jr. (BGLE) to J.

R. Hiller (NRC),

'Calvert Cliffs Nuclear Powt.r Plant Unit 1; Docket No. 50-317, Amendment to Operating License DPR-53, Eighth Cycle License Appilcation," February 22, 1985.

17. CENPD-133, Supplement 3 P, "CEFLASH 4AL A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident,' January 1977.

' 18. CENPD 138, Supplement 2-P, ' PARCH, A FORTRAN IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup,"

January 1977.

11-9

_____m..___

/

Reference - Section 9 1.

Letter, G. C.

Creel (BG&E) to Document Control Desk (NRC),

'Calvert Cliffs Nuclear Power Plant Unit No. 2 Docket No. 50-

318, Request for Amendment, Unit 2 Ninth Cycle License Application," February 7,1989.

O 11-10

_______2--.-.____-_

A

^

Reference - Section 10 Letter, G. C. Creel (BGLE) to Document Control Desk (NRC),

1.

"Calvert Cliffs Nuclear Power Plant Unit No. 2 Docket No. 50-

318, Request for Amendment, _ Unit 2 Ninth Cycle License Application," February 7, 1989, 1

4

'l,::

l 11-11

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