ML20002A990

Proposed Changes to STS Re Cycle 4

ML20002A990
Person / Time
Site:	Calvert Cliffs
Issue date:	12/04/1980
From:	BOSTON EDISON CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20002A984	List: ML20002A982 ML20002A990
References
NUDOCS 8012090291
Download: ML20002A990 (48)

Text

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A-1 i

4 ENCLOSURE A I

Calvert Cliffs Unit II Cycle 4 i

Refueling License Amendment o

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.801209o

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o A-2 CALVERT CLIFFS UNIT II CYCLE 4 REFUELING LICENSE AMENDMENT Table of Contents Section Page Number 1.

Introduction and Summary 3

2.

Operating History of Calvert Cliffs II Cycle 3 4

3.

General Description 5

4.

Fuel System Design 10 5.

Nuclear Design 11 6.

Thermal-Hydraulic Design 22 7.

Transient Analysis 23 8.-

ECCS Analysis 35 9.

Technical Specifications 36

10. Startup Testing 46

11. References 47 l

s e

A~3 1.

INTRODUCTION AND

SUMMARY

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

All planned operating conditions remain the same as those for Cycle 3.

The core will consist of presently operating 0 and E assemblies and fresh Batch F assemblies.

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Plant operating requirements have created a need for flexibility in the Cycle 3 terminatios point, ranging from 10,000 MWD /T to 11,000 MWD /T.

In performing analyses of postulated accidents, determining limiting safety settings and establishing limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 4 conditions are enveloped, provided the Cycle 3 termination point falls within the above burnup range.

The evaluations of the reload core characteristics have been examined with respect to the Calvert Cliffs Unit I Cycle 5 safet/ analysis described in References 1 and 2, hereafter referred to as the " reference cycle" in this report unless otherwise noted.

This is an appropriate reference cycle because of the similarity in the basic system characteristics of the two reload cores.

Specific core differences have been accounted for in thf present analysis.

In all cases, it has been concluded that either the reference cycle analyses envelope the new conditions or the revised analyses presented herein continue to show acceptable results.

Where dictated by variations from Cycle 3, proposed modifications to the plant Technical Specifications are provided.

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A-4 2.

OPERATING HISTORY OF CALVERT CLIFFS II CYCLE 3 Calvert Cliffs Unit II is presently operating in its third fuel cycle utilizing Batch, B, C, D and E fuel assemblies. Calvert Cliffs Unit II Cycle 3 began operation on December 6,1979 and reached full power on December 12. The Cycle 3 startup testing was reported to the NRC in Reference 3.

Cycle 3 is presently scheduled to terminate on about January 2,1981 with a cycle burnup of approximately 10,800 MWD /T. However, flexibility in this endpoint burnup is necessary because of uncertainties in the Unit capacity factor during the remainder of Cycle 3.

The Cycle 3 termination point can vary between 10,000 MWD /T and 11,000 MWD /T to accommodate the plant schedule and still be within the assumptions of the Cycle 4 analyses. As of mid-November 1980, the Cycle 3 burnup had reached 9850 MWD /T, Initial criticality of Cycle 4 is expected to occur on or about February 12, 1981.

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A-5 3.

GENERAL DESCRIPTION The Cycle 4 core will consist of the number and types of assemblies and fuel batches as described ~in Table 3-1.

The primary change to the core in Cycle 4 is the removal of 1 Batch B. assembly, 68 Batch C assemblies, and 59 Batch D assemblies.

These assemblies will be replaced by 40 Batch F (3.65 w/o enrichment) and 88 Batch F/ (3.03 w/o enrichment) assemblies.

The 88 low enrichment Batch F/ assemblies contain 8, burnable poison pins per assembly.

Figure 3-1 shows the fuel management pattern to be employed in Cycle 4.

Figure 3-2 shows the locations of the fuel and poison pins within the lactice of the Batch F/ assemblies and the fuel pin locations in the unshimmed Batch F assemblies.

This pattern will accommodate Cycle 3 termination burnups from 10,000 MWD /T to 11,000 MWD /T.

The Cycle 4 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 fuel rods, initial enrichment and burnup.

Figure 3-3 shows the beginning of Cycle 4 assembly burnup distribution for a Cycle 3 termination burnup of 10,500 MWD /T.

The initial enrichment of the. fuel assemblies is also shown in Figure 3-3.

TABLE 3-1

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CALVERT CLIFFS UNIT II CYCLE 4 CORE LOADING Batch' Average Initial Total Total Initial Burnup Poison Poison Number Number Assembly Number of Enrichment E0C3 =

Rods per Loading Poison Fuel Designation Assemblies Wt% U-235 10,500 Assembly Wt% B4C Rods Rods D

25

-3.03 20,300 0

0 0

4400 E

48 3.03 9000 0

0 0

8448 E/

16 2.73.

12,000-0 0

0 2816 F

40 3.65 0.

0' 0

0 7040 F/

88 3.03 0

8 3.03 704 14,784 TOTALS 217 704 37,488 Note:

Shim B10 concentration equals.02685.gms B10/ inch cn

A-7 F

F F

F Fl E

FI F

Fl E

Fl E

El F

FI E

Fl D

Fl 9

F Fl E

Fl D

Fl E

Fl F

E FI D

Fl E

- Fl El Fl Fl D

FI E

FI D

FI F

~

E E

Fl E

FI D

FI El F

Fl El FI El FI 9

D

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CALVERT CLIFFS UNIT II CYCLE 4 Figure GAS L

T CCO.

coivert Clins CORE MAP 3-1 Nuclear Power Flont

UNSHIMMED ASSEMBLY

-a 4

8 POISON R0D ASSEMBLY X

X 1

X

!X X

X

' FUEL R0D LOCATION

{

POIS0N R00 LOCATION CALVERT CLIFFS UNIT II CYCLE 4 Figure GAS El.E T IC CO.

carvert ci;rrs ASSEMBLY FUEL AND OTHER ROD LOCATIONS 32 Nuclear Power Plant

A-9 o

INITIAL ENRICHMENT, w/o U-235 3.65 3.65 B0C4 BURNUP (MWDIT) EOC3 = 10,500 MWDIT 0

0 3.65 3.65 3.03 3.03 3.03 0

0 0

8,400 0

3.65 3.03 3.03 3.03 3.03 2.73 0

0 7,600 0

9,000 12,000 3.65 3.03 3.03 3.03 3.03 3.03 2.73 0

0 7,300 0

20,200 0

12,000 3.65 3.03 3.03 3.03 3.03 3.03 3.03 3.03 0

0 7,300 0

20,100 0

11,600 0

3.65 3.03 3.03 3.03 3.03 3.03 3.03 2.73 0

7,600 0

20,100 0

10,100 0

12,400 3.03 3.03 3.03 3.03 3.03 3.03 3.03 3.03 3.65 0

0 20,200 0

10,100 0

20,300 0

0 3.03, 3.03 3.03 3.03 3.03 3.03 3.03 2.73 3.65 8,400 9,000 0

11,600 0

20,300 0

11,800 0

3.03 2.73 2.73 3.03 2.73 3.03 2.73 3.03 0

12,000 12,000 0

12,400 0

11,800 21,800 i

BALTIMORE CALVERT CLIFFS II CYCLE 4 scure GAS & ELECTRIC CO.

ASSEMBLY AVERAGE BURNUP AND INITIAL

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Coivert ciirrs ENRICHMENT DISTRIBUTION 3-3 Nuclear Power Plant

A-10 4.0 FUEL SYSTEM DESIGN.

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.4.1 MECHANICAL DESIGN

. The-mechanical design for the' standard Batch F reload fuel is identical to that of the standard Batch E_ fuel used in Calvert Cliffs 2 -(Reference 9) 1 and ofLthe Calvert Cliffs 1 standard Batch G fuel described in the reference cycle submittal (Reference 1).

Details of.the standard Batch D fuel-design parameters can be found in Reference 4.

C-E has performed analytical predictions of cladding creep-collapse time for all Calvert Cliffs Unit 2 fuel batches that will be irradiated in 2

-Cycle 4 and has concluded that the collapse resistance.of all standard fuel rods is sufficient to preclude collapse' during their design lifetime, This. lifetime will not be exceeded by the Cycle 4 duration (Table 4-1).

t These analyses utilized the CEPAN computer code (Reference 5) and included-as input conservative. values of internal pressure, cladding dir2nsions, cladding temperature and neutron flux.

Table 4-1 E0C4 Minimum.

Batch Collapse Time Exposure D

>29,500 Hours 28,433 Hours E

>22,200 Hours 21,236 Hours F

>22,200 Hours 12,965 Hours The metallurgical ~ requirements of the fuel cladding and the fuel assembly i

structural members for the Batch F fuel are identical to those of the i

Batch D and E fuel from Cycle 3.

Thus, the chemical or metallurgical l

performance of the Batch F fuel will remain unchanged from the performance i

of the-Cycle 3 fuel.

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-4.2 HARDWARE MODIFICATIONS TO MITIGATE GUIDE TUBE WEAR All standard fuel assemblies which will be placed in CEA locations in Cycle 4 will-have stainless steel sleeves installed in the guide tubes to prevent guide tube wear.' A detailed discussion of the' design of the sleeves and their effect on reactor operation is contained in Reference 6.

4.3.

THERMAL DESIGN Using the FATES fuel evaluation model (Reference 7), the thermal perfomance -

of the various~ fuel' assemblies (fuel Batches D, E, and'F).has been.

evaluated with respect to prior burnup, the proposed burnup during Cycle 4, their respe~ctive fuel characteristics, and expected flux level during i

- Cycle 4, The' fresh fuel, Batch F, has been' determined to be.the limiting 'aei batch with respect.to stored energy.

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A-11 5.0 '

NUCLEAR DESIGN 5.1 PHYSICS CHARACTERISTICS 5.1.1 Fuel Manacement The Cycle 4 fuel management employs a mixed central region as c3 scribed. in Section 3, Figure 3-1.

The fresh Batch F is comprised of two sets of assemblies, each having a unique enrichment in order to minimize radial power peaking.

There are 40 assemblies with an enrichment of 3.65 wt% U-235, 88 assemblies with an enrichment of 3,03 wt% U-235 and 8 poison shims per assembly.

With this loading, the Cycle 4 burnup capacity for full power operation is expected to be between 17,100 MWD /T and 17,600 MWD /T, depending on the final Cycle 3 termination point.

The Cycle 4 core characteristics have been examined for Cycle 3 terminations between 10,000 and 11,000 MWD /T and limiting values established for the safety analyses.

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

Physics characteristics including reactivity coefficients for Cycle 4 are listed in Table 5-1 along with the corresponding values from the reference cycle.

Please note that the values of parameters a tually employed in safety analyses are different from those displayed in Table 5-1 and are typically chosen to conservatively bound predicted values n th accommodation for appropriate uncertainties and allowances.

1 Table 5-2 presents a summcry of CEA shutdown worths and reactivity allowances for the end of Cycle 4 zero power steam line break accident with a comparison to reference cycle data.

The EOC zero power steam line break was selected. since it is the most limiting zero power steam line break accident, ud thus provides the basis for establishing the Technical Specification-shutdown worth.

4 A 12 Table 5-3 shows the reactivity worths of various CEA groups calculated at full power conditions for Cycle 4 and the reference cycle.

5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (AR0) planar radial power distributions at 80C4, MOC4 and EOC4 that are character-

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istic of the high burnap end of the Cycle 3 shutdown window.

These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80 percent of the fuel height.

The higher burnup end of Cycle 3 shutdown window tends to increase the power peaking in this axial central region of the core for Cycle 4 The planar radial power distributions for the above region with CEA Group 5 fully inserted at beginning and end of Cycle 4 are shown in Figures 5-4 and 5-5, respectively, for the high burnup end of the Cycle 3 shutdown window.

The maximum planar radial pin peak of 1.48 occurs at beginning of cycle and decreases over the cycle.

It is characteristic of both AR0 and Bank 5 inserted conditions that the Cycle 4 peaks are highest near 80C.

The radial power distributions described in this sction are calculat1d data without uncertainties or other allowances.

However, single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice. 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 4 These conservative values, which are used in Section 7 of this document, establish the allowable lin :ts for power peaking to be observed during operation.

The range of allowable axial peaking is defined by the limiting conditions for operation covering axial shape index (ASI).

Within these ASI limits, the necessary DNBR and kw/ft margins are maintained

A 13 for a wide range of possible axial shapes.

The maximum three-dimensional or total peaking factor anticipated in Cycle 4 during normal base load, all rods out operation at full power is 1.85, not including uncertainty allowances and augmentation factors.

5.1.3 Safety Related Data The safety related data for Cycle 4 is identical to the safety related data used in the reference cycle analysis as presented in Section 5.1.3 of Reference 1.

5.2 ANALYTICAL INPUT TO IN-CORE MEASUREMENTS In-core detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the manner described in' Reference 8, which is the same method used for the reference cycle.

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

5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties which are applied to Cycle 4 are the same as those applied to the reference cycle (Reference 1).

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A-14 TABLE 5 <

CALVERT Cliffs UNIT II CYCLE 4-NOMINAL PHYSICS CHARACTERISTICS REFERENCE UNITS CYCLE 4 CYCLE Dissolved Boron Dissolved Boron Content for Criticality, CEAs Withdrawn Hot Full Power,.

PPM 1150 1010 Equilibrium Xenon, BOC Boron Worth Hot Full Power B0C PPM /%Ap 105 101

[_

Hot Full Power EOC PPM /%ap 83 83 Reactivity Coefficients.

(CEAs Withdrawn)

-Moderator Temperature

' Coefficients, Hot Full Power, Equilibrium Xenon Beginning of Cycle 10-4ap/*F 0.0-

-0.1 End of Cycle 10-4ap/*F

-1.9

-1.9

. Doppler Coefficient I

Hot Zero Power BOC 10-5ap/*F

-1.55 '

-1.55 Hot Full Power BOC 53pfer

. l.16

-1.21 Hot Full Power EOC 10-53pfog

_j,40 1,40 Total Delayed Neutron Fraction, Seff s

BOC

.00662

.0bC?d

-EOC

.00517

.00521 Neutron Generation' Time, 1*

BOC 10-6sec-23.8 24.4 j

f E0C 10-6sec 29.8

'29.7

A-15 TABLE 5-2 Calvert Cliffs Unit II Cycle 4 Limiting Values of

' Reactivity Worths and A11cwances for Hot Zero Power Steam Line Break, %op End-of-Cycle (EOC) 4 Reference Cycle Cycle 4 1.

Worth of All CEA's Inserted 9.4 9.2 2.

Stuck CEA Allowance 2.2 2.0 3.

Worth of All CEA's Less Highest Worth CEA Stuck Out 7.2 7.2 4.

Zero Power Dependent Insertion Limit CEA Bite 2.0 2.2 5.

Calculated Scram Worth 5.2 5.0 6.

Physics Uncertainty (10%ofItem5)

<5

.5 7.

Net Available Scram Worth (Item 5 minus Item 6) 4.7 4.5 8.

Technical Specification Shutdown Worth 4.3 4.3 9.

Margin in Excess of Technical Specification Shutdown Worth

+0.4

+0.2 i

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A-16 TABLE 5-3 CALVERT CLIFFS UNIT'I! CYCLE.4 REACTIVITY WORTH 0F CEA REGULATING GROUPS AT HOT FULL POWER, %ao Beginning of Cycle End of Cycle Regulating Reference Reference CEAs Cycle 4 Cycle Cycle 4 Cycle Group 5 0.46 0.49 0.53 0.57 Group 4 0.28 0.32 0.41

. 0.39 Group 3 0.89 0.97 1.01 0.93 j

Note:

Values shown assume sequential group insertion.

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A-17 0.72 0.98 x

0.70 0.98 0.97 1.10 1.20 0.77 1.04 1.14 1.18 1.14 1.04 0.77 1.05 1.15 1.16 0.92 1.16 1.02 0.70 1.04 1.15 1.15 0.90 1.12 1.05 1.14 0.98 1.14 1.16 0.90 1.10 1.04

~1.06 0.89 0.97 1.18 0.92 1.12 1.04 1.02 0.77 0.93 0.72 1.10 1.14 1.16 1.05 1.06 0.77 0.88 0.74 0.98 1.20 1.04 1.02 1 14 0.89 0.93 0.74 0.63 NOTE: x = MAXIMUM 1 - PIN PEAK = 1.aR BALTIMORE CALVERT CLIFFS II CYCLE 4 Figure GAS & ELECTRIC CO.

ASSEMBLY RELATIVE POWER DENSITY AT BOC, Coiver Clins EQUILIBRIUM XENON 5-1 Nuclear Power Plant

A-18 0.62 0.80 0.65 0.87 0.92 0.93 1.07 O.74 1.02 1.03 1.15 1.02 0.92 0.74 1.05 1.08 1.19 0.91 1.19 0.97 0.65 1.02 1.08 1.21 0.93 1.22 1.07 1.23 0.87 1.03 1.19 0.93 1.22 1.10 1.23 0.99 x-0.92 1.15 0.91 1.22 1.10 1.21 0.91 1.16 0.62 0.93 1.02 1.19 1.07 1.23 0.91 1.13 0.91 L

1.07 0.92 0.97 1.23 0.99 1.16 0.91 0.80 NOTE: x = MAXIMUM 1 - PIN PEAK = 1.40 i

BALTIMORE CALVERT CLIFFS II CYCLE 4 ngure GAS & ELECTRIC CO.

ASSEMBLY RELATIVE POWER DENSITY AT 8 GWDlT, Calvert clirrs EQUILIBRIUM XENON 5-2 Nuclear Power Plant

A-19 1

t 0.68 0.84 0.70 0.90 0.99 0.95 1.10 0.79 1.06 1.02 1.16 1.00 0.92 4

0.79 1.08 1.05 1.18 0.91 1.16 0.94 4

4 0.70 1.06 1.05 1.18 0.92 1.17 1.01 1.18 x

i 0.90 1.02 1.18 0.92 1.18 1.03

-1.18 0.96 l

0.99 1.16 0.91 1.17 1.03 1.18 0.91 1.17 l

0.68 0.95 1.00 1.16 1.01 1.18 0.91 1.17 0.94 0.84 1.10 0.92 0.94 1.18 0.96 1.17 0.94 0.84 NOTE: x = MAXIMUM 1 - PIN PEAK = 1.32 BALTIMORE CALVERT CLIFFS II CYCLE 4 rigure GAS & ELECTRIC CO.

ASSEMBLY RELATIVE POWER DENSITY ATE 0C Coiveri clirrs EQUILIBRIUM XENON 5-3 Nuclear Power Plant

A-20 CEA BANK 5 0.71 0.96 LOCATIONS 0.70 0.98 0.98 1.06 1.12

//b7 0.66 1.01 1.13 1.19 1.08 7

77'

'. 07 1.18 0.95 1.17 0.99 0.66 0

1

//

0.70 1.01 1.07 1.15 0.94 1.19 1.11 1.21 x

0.98 1.13 1.18 0.94 1.19 1.12 1.16 0.99 0.98 1.19 0.95 1.19 1.12 1.12 0.83 0.99 0.71 1.%

1.08 1.17 1.11 1.16 0.83 0.90 0.70

/ //,

/ //,

1.12 0.79 0.99 1.21 0.99 0.99 0.70 0.38

//

l//

NOTE: x = MAXIMUM 1 - PIN PEAK = 1.45 BALTIMORE CALVERT CLIFFS II CYCLE 4 r;gure GAS & ELECTRIC CO.

ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5 i Nuc e r w

Plant

A-21 CEA BANK 5 LOCATIONS 0.69 0.85 0.70 0.92 1.01 0.95 1.05

' //

/

0.66 1.02 1.04 1.19 0.97

/0. 7 If //

0.66 0.98 1.19 0.95 1.17 0.93

// /

0.70 1.02 0.98 1.16 0.95 1.24 1.07 1.23 0.92 1.04 1.19 0.95 1.24 1.11

-1.26 1.03 x

1.01 1.19 0.95 1.24 1.11 1.25 0.96 1.20 0.69 0.95 0.97 1.17 1.07 1.26 0.96 1.13 0.86 0'85

/'//

/',0.47/l//

1.05

/0.7 0.93 1.23 1.03 1.20 0.86

// /

'//b NOTE: x = MAXIMUM 1 - PIN PEAK = 1.40 BALMAORE CALVERT CLIFFS II CYCLE 4 Ficure GAS & ELECTRIC CO.

ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5

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INSERTED, HFP, E0C 5-5 Nuc e r w

Plant

A-22 6.0 Thermal Hydraulic Design 6.1 DNBR Analysis The thermal hydraulic models and pertinent design parameters used for Calvert Cliffs II Cycle 4 are the same as those used in the reference cycle as reported in Reference 2 and corrected in Reference 10.

6.2 Effects of Fuel Rod Bowing on DNBR Margin The fuel rod bowing effects on DNB margin for Calvert Cliffs Unit II have been evaluated within the guidelines set forth in Reference 11.

A total. of 89 fuel assemblies will exceed the NRC-specified DNB penalty threshold burnup of 24,000 MWD /T, as established in Reference 11, during Cycle 4.

At the end of Cycle 4, the maximum burnup attained by any of these assemblies will be 37,100 MWD /T.

From Reference 11, the corresponding DNB penalty for 37,100 MWD /T is 4.4 percent.

An examination of power distributions for. Cycle 4 shows that there exists at least 6.0 percent DNB margin for assemblies exceeding 24,000 MWD /T relative to the DNB limits established by other assemblies in the core. This margin is considerably greater than the Reference 11 reduction penalty of'4.4' percent imposed upon fuel assemblies exceeding 24,000 MWD /T in Cycle 4.

Therefore, no power penalty for fuel rod bowing is required in Cycle 4.

7, ;.7 - + -

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.'7.0~

TRANSIENT ANALYSIS The purpose ofJ his'sdction'is to present.the~ result's of the 3nitimore t

Gas' & Electric Calvert ' Cliffs Unit 'II, Cyc1'e 4 Non-LOCA safety analysis ~

lat-2700 Mut. _The Design: Bases Events (DBEs) considered in;the safety analysescare_ listed in Table.7-1.

Each of'the events listed in Table 7-1 has been reviewed for Cycle 4 to_ determine if;an explicit reanalysis vas required.-. Table 7-1 indicates-the analysis status of each transient.

Each DBE_was reviewed by com-

. paring'all the current and reference cycle key transient _ parameters that significantly impact the results of the event.

The reference cycle is one for.which a'DBE in question;has been shown to meet required safety-criteria.

If:all the-current cycle values of key parameters for a i~

'particular event _ are bounded' by (conservative' uith respect to, or the

.same as) the reference cycle,ino reanalysis is required.

2 The reference cycle-for this. analysis is Calvert Cliffs Unit I,. Cycle 5

-(Reference I as-amended'by Reference 2).

The results of the review were that the key. input _ parameters to all the

- DBCs for Unit II Cycle 4 operation are'the same as or less_ limiting than the specified reference cycle input' parameters (see Table 7-2) except.

for the Loss of Flow (LOF)' event.

The Loss of Flow (LOF)' event was reanalyzed to account for the' fact that-the flow coastdown for Unit II is different from, and more adverse than, the coastdown for Unit I.

Therefore'as indicated in Table.7-1,-only.the Loss of Flow transient has been reanal; zed for Unit II Cycle 4.

1 1

For all DDEs other than'the LOF event, the reference cycle safety analyses bound the results that would be obtained for Cycle 4 and demonstrate safe operation of the Calvert Cliffs Unit II Cycle 4 at'2700 MWt.

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y i...esr-9--

i TADI.E ' 7-1 A-24 b

D i @D

[D

" Chb'ERT CLIFFS UillT II, CYCLE 4

+.

us IliCIDENTS CONSIDERED Ill TRidiSIEliT A!:D ACCJDElli A:ALYSIS s

Analysis Status Anticipated Operational;0ccurrences for which the RPS Assures no Violation of SAFDLs:

Control Element Assembly !!ithdrawal liot Reanalyzed

,Doron Dilution flot Reanalyzed Startup of an Inuctive. React'r Coolant Pump Not' Reanalyzed o

. Excess Load

~..,

!!ot Reanalyzed

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4 Loss of Load Not Reanalyzed s

Loss of Feedwater Flow Not Reanalyze'd Excess lleat Removal due to Feedwater lialfunction Not Reanalyzed Reactor Coolant System Depressurization Not Reanalyzed 1

Loss of Coolant Flow

.r

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Reanalyzed LohsofACPower Not Reanalyzed Anticipated Operational Occurrences which are Dependent on Initial Overpower.largin for Piotectica Against Violation of SAFDLs:

3 Loss of Coolant Flow Reanalyzed Loss. of AC Power Not Reanalyzed

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Full Length CEA Drop Not Reanalyzed Transients Resulting _ from 14alfunction of One

Not Reanalyzed Steam Generator 2 Postulated Accidents:.

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CEA E,iection Not Reanalyzed Steam Line Rupture Not Reanalyzed Steam Generator Tube Rupture Not Reanalyzed Seized Rotor Not Reanalyzed

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Requires low Flow Trip.

2R: quires Asymetric Steam Generator' Protective Trip. Function

t, A-25 TABL'E 7-2 CALVERT. CLIFFS U!!IT II CYCLE 4 CORE PAR /4;ETERS ":PUT TO SAFETY AtlALYSES s

^

FOR Of(B At?D CTM (CEliTERLIliE-T0 MELT) DESIGil LIMITS

~

Reference Unit II

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' Cycle Values Cycle 4 Units Unit I Cycle 5 Values Physics Parameters Radial Peaking Factors t'

• For DI!B'Marein' Analyses (F )

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r 1,62 1.62 Unrodded Region ll'.

Bank 5 Inserted 1.78 1.78 For Planar Radial Component (Fxy)'

a of 3-0 Peak (CTM Limit Analyses)

~

1.62 1.62 Unrodded Region

,1.78 1.78 Bank 5 Inserted l.055 1.~055

. Maximum Augmentation Factor

~10-4dp/ F.

-2. 5*-* +. 5

-2. 5*-* +. 5

' Moderator Temperature Coefficient

-4.3 Shutdown Margin (Value assumed

%6p

-4.3 in Limiting EOC Zero Power SLS)

Tilt Allowance

.J 3.0 3.0 Safety Parameters Power Level

MWt,

, 2754 2754 Maximum Steady State Core Inlet

• F

~

550 550 Temperature Minimum Steady State RCS Pressure" psia 2200 2200 6

Reactor Ccolant Flow '(550*F, 2200 psia) 101b/hr 133.9 133.9 Hegative Axial Shape Index LCO I

.16

.16 P

ext;reme assumed at Full Power

' Maximum CEA Insertion at Full Power

% of Insertion of 25 25 '

Bank 5

', 'Maxir.WInitial Linear Heat Ra'tE,for KW/ft 16.0 16.0 Transient Other Than LOCA Steedy State Linear Heat Rate to Fuel DI/ft 21.0 21'.0 Centerline ' alt-Assumed in the Safety

.' Analyses CEA Drop Time from P.enoval of sec 3.1 3.1 ~

Power to Holding Coils to 90%'

)

Insertion Minimum D:tBR (CE-1) 1.195 1.195

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• The effective initic.1 MTC for the SLB ever.t is -2.?X10 ap/*F.

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-A-26 7.1 l Loss of' Coolant Fist Event

-t The Loss of Coolant Flou event uns reanaly cd for Cycle 4 to determine the minimum' initial margin that must be maintained-by the Limiting Conditions for Ope' ration (LCOs)lsuch that in conjunction uith the RPS

-(low flow trip),: the DNBR limit vill not be execeded.

-The methods used to analyze this event are the some as those'used in the reference cycle analysis.

The 4-pump Loss of Coolant Flow produces a rapid approach to the DNBR limit due to the rapid decrease-in the core-coolant flow.

Protection

' against~ exceeding the DMBR-limit for this transient is provided by the initial steady state thermal' margin which is maintained by adhering to

-the' Technical Specifications LCOs on DNB and by the response of the RPS which provides cu: automatic reactor trip on low recetor coolant.

flow as ceasured by the-steam generator differential pressure transmitters.

The transient is characterized by the flow'coastdown curve given in

' Figure 7.1-1.

Table 7.1.-1 also lists the key transient parameters used Lin-the present analysis and compares them with comparable reference cycle values.

Tablef. 7.1-2 presents the'NSSS and RPS responses during a four pump loss of flow initiated at a 0.0' shape index.

The low flow trip setpoint is reached at 0.90 seconds and the scram rods start dropping into the core at 1.9 seconds.

A minimum CE-1 DNBR of 1.195 is reached at 3.22 seconds.

Figures 7.1-2 to'7.1-6 present the core pouer, heat flux, RCS pressure,-

core coolant temperatures and the DNBR as a function of time'.

The analysis shows that a Loss of Flou event mitigated by the action of the Low Flow Trip will ensure that DNBR limit vill not be exceeded when the initial conditions are no more severe than those permitted by adherence

-to the Technical Specification LCO's.

)

e fu WJ JU(

w> u l

j I

f 1

e D

D

/

4 h-

+

n

- A-27 TAPLE 7.1-1

~

- KEY PARNIETERS: ASSUMED Ill TliE LOSS OF C00LAllT FLO;f AllALYSIS

~

~

Paremeter Units Reference Cycle

• Cycle 4 Initial Core Power Level Milt 2754

~

.I5i54 Initial Core Inlet Coolant

• F 550 550 Temperature 6

Initial Core'Itass Flow Rate 10 lbm/hr

-133.9 133.9

^ 22 00 Reactor Coolant System Pressure psia 2200 Moderator Temperature Coefficient 10~44p/F

+. '5

+.5 1.00 1.00**

Poppler Coefficient-Multiplier LFT Response Time-sec

.5

.5 CEA Holding coil Delay sec

.5

.5'

~

.CEA Time to 90" Insertion sec 3.1 3.1 (Including Holding coil Delay)

CEA Worth at Trip (all rods ou't) 10-2ap

-5.60'

-5.60 Unrodded Radial Peaking Factor 1.62 1.62

.(FJ) 4-Pump RCS Flow Coastdown Figure 7.2.1-1 Figure of Reference Cycle 7.1-1 (see Reference 2)

J Un#t I Cycle 5 C*

Since this is a second order effect and the most limiting doppler nultiplier varies during the transient, a ncminal value is used.

9 e

e g

e e

e m

~

' A-28' TABLE 7.1-2 SEQ'JEl:CE OF EVE!!TS FOR LOSS".0F FL0tl a

- Time (sec.) -

J-

-hvent-Setpoint or Value O.0 Loss of Power to all Four Reactor Coolant Pumps-

-0.90

. Low Flow Trip Signal. Generated 93% of initial 4-Pu:-

Fl e.

1.40 Trip Breakers Open 1.90 Shutdown CEAs Begin to Drop Into Core 1.195 3.22

, Minimum,CE-1 D!lBR.

2307 6.00 Maximum RCS Pressure, psia O

r

.v y hp

.opw o

7 onsk 5

a.

6 4

s m

-)

A -

e

1. 0 4-PUMPC0ASTD0UN

0. 8 O

Ht; 0. 6 of W

?:

S b 0.4 cc Oo 0.2 0;

0 4

8 12 16 20 TIME, SECONDS BALTNOS:

I

~

GAS !. ELECTP.12 CO.

LOSS OF C00Lo.NT FLOW EVENT coiven cuits 7.1-1 Nuclect Pcwcr Plant (hEFLCliFRACT10iiVSTIME

A-30 120 IbO

~

g 80 N

y b

60 g

6 y

W 40 8

20 0

I I

I I

O 2

4 6

8 10

TIME, SECONDS LOSS OF COOLANT FLOV/ EVENT GAS i

C O.

cc.1 veri cufh CORE POWER vs TIME 7.1-2 Nuclec.r Power Plant w

w

A-31 a

~

120 d

~

100 w

d 8

80 M

165 e

-g 60 40

~

20 0

I I

I I

O 2

4 6

8 10

TIl1E, SECONOS cAsI^tttIISt!co.

LOSS OF COOLANT FLOW EVENT 7'

coiveri cntr5 CORE HEAT l-LUX vs TIME h!vclec.r Pawer Plant

..-e.

A-32.

~ '-

~'

2400 t

~

2350 x

m ea-2300 LLI Eam m

2250 to E

mu 2200 2

~

2150 2100 I

I I

I 0

2 4

6 8

10

TIl1E, SECONOS

^

LOSS OF C001. ANT FLOW EVENT 7,1_ q GAS EEi C O, cois.e,i c!;ns RCS PRESSURE vs TIME I

Nuclear Power Plant j

A-33 G20

~

TOUTLET 600 TAVG 580 u_

N 560

[5 T

ig E

~

f2 8

540 520 500 I

I I

I O

2 4

6 8

10

.TIl1E, SECONOS

. 0^LT CE LOSS OF COOLAilT Fi.0Y! E\\!ENT GAf E. ELECTP.lC CO.

coiveri citris RCS TEMPER ATURES vs TIME 7' 1. g Nucicer Power Picnt e

A-34 hk hkh 2.0 iii i

i i

i i.

i 1,9 ASI=0,0 1,'8 _.

1,7 1.6

~

~

c2 g

1,5 5y

.t,1; E

1,3

/

1,2 1.1 CE-1 RER LIIIIT OF 1,195 1.0 i

i i

i i

i i

i i

0,0 1.0 2,0 3,0 ti,~0 5.0

~

TIE, SEC

~

^

LOSS OF COOLANT FLOW EVENT GAS LL I C O.

ceivcit cims RSR(CE-1)VSTifE 7.1-6 Nucleo: Power Plant -

A-35

' 8.0 ECCS ANALYSIS' r

An ECCS performance analysis was perfonned for Calvert Cliffs Unit 2 Cycle 4 to demonstrate compliance with 10CFR50.46 which presents:the NRC Acceptance Criteria' for Emergency Core Cooling Systems' forl Light-Water-Cooled reactors (12) i

. The analysis justifies an allowable peak linear heat generation rate (PLHGR)

~

]

of.15.5 kw/ft which is equal-to. the existing limit for Unit 2.

The ECCS performance analysis for Calvert Cliffs Unit 1. Cycle 5 operation (13) was used as the reference cycle analysis'for the Unit 2 Cycle 4 evaluation.

That analysis used f0e1 performance data which bound both_ Unit 1 Cycle 5 2

and Unit 2,' Cycle 4.

Therefore. the -results reported in Reference 13 are -

-t applicable to Unit 2 Cycle 4.

i The results of that ' analysis identified-the peak clad. temperature as 1987*F

~

4, as opposed to the acceptance limit-of 2200 F.

The peak local' c?tJ -oxidation-was 9.7% versus the acceptance -limit of 17% 'and the. peak core wide clad' oxidation was less 'than.51%.versus the acceptance limit of-l.0%.

Hence,

~

Unit 2 Cycle 4 operation.at a peak-linear heat generation rate of 15.5 kw/ft 4

t (102% of 2700 Mw ) will result in acceptable and 'at a power level of 2754 Mw t

ECCS performance.

4 g

j I

9 d

?

s b, '

t w

~

n-v

-3 r

-gy,-,y y.

,~

eL A-36 9.0 Technical Specifications The Technical Specification changes which must be made in order to make the Calvert Cliffs II Technical Specifications valid for the operation of Cycle 4 are nearly identical to the changes made in the reference cycle as ' reported in References 1 and 2.

Specific differences are:

1.

Present Unit II Technical Specifications contain a most negative MTC

-2.5 x 10 j.3 x 10'4 ak/k/*F as compared to the former Unit I' limit of limit of -

. For both units the most negative MTC limit is ak/k/*F.

being changed to -2.2 'x 10-4 ak/k/*F.

2.

The present Unit II peak linear heat rate limit is 15.5 kw/ft and, therefore, no change is 'needed. -The' Unit'I limit was raised from 4

14.2 kw/ft to.15.5 kw/ft.

3.

The present radial peaking factor limits for Unit II are different than

'the former Unit I limits. As for Unit I,these peaking factor limits must be changed to-l.62.

t Table.9-1 presents & summary of the Technical Specification changes required for_ Unit II.

For your convenience, the-items in this table are presented in the same order as the changes presented in References 1 and 2.

1 Specific pages from the ~ Unit II Technical ' Specifications showing the required modifications are.not included since the corresponding Unit I pages can be found in the reference indicated in Table 9-1 for each change. Table.9-2 presents the explanations for the changes summarized in Table 9-1.

i f-W uw T'77 v

r T'7-W W

M

Rf A, 0 A-37

[C'

.gf \\a.

m m-f D

• ~

c

,a TACLE 9-1 Calvert Cliffs !! Cycle 4-Technical Specifict. tion Changes

+

Chance a Tech Snec 8 Action Reference 1

Figure 2.1-1 page 2-2 Replace Figure 2.1 2 2

Table 2.2-l page_2-9 Change stean generator pressure-les setpoint fror.: 1500 psia to 1570 psia 1

~

3 Table 2.2-1 page 2-10 Add stcan generator pressure difference -

I high setpoint 4

Table 2.2-1 page 2-10 Change steam generator orcssure-lcw trip bypass frca bclu.: 600 psia ~to below 635 psia 1

ft Figure 2.2-1 page 2-11 Replace Figure 2.2-1 2

6 Figure 2.2-2 page 2 !!o changes from Cycle 3

~'

~

7 Figure 2.2-3 pg,e 2-13 Nochanges'frimCycle'3.

~

8 B.2 1.1 page B2-l Remove numerical specification of LHG2 to centerline melt 1

9 8.2 1.1 page 02-1 tio changes from Cycle 3 10 B.2.1, B.2.2 Change ninimun D';CR value from 1.19 to 1.105 2

pages 02-1, B2-3, B2-5, B2-6 t

11 8.2.2.1 page 82-4 fio changes from cycle. 3 12 B.2.2.1 page B2-5 Change stesa generator crcsserc-lew setpoint fron 500 psia to 570 psia 1-13 C.2.2.1 page D2-7 Revise Arcription of T"/LP trip and add asynN ric stc3; g r.erater transier.L ;irotec-

!tive trip function description I

U.

[

t :.

A-38~

L f.'_

1 7

g g PD g d

La L

l TABLE 9-1 (continued)

Channe (

Tech Spec f Action Reference

~

14 3.1.1.1 page 3/4 1-1

- Change Shutdown Margin Tavg >200 F from

>3.4%:k/k to >4.3%:k/k and enance liiinimum boration concentration from.1720 1

ppm to 2300 ppm 15 3.1.1.2 DaQe 3/4 l-3, Change Shutdown Margin Tavg 1200 F

' ~

from y1.0%ak/k to 23.0%a k/k and change j

minimu_m boration concentration from 1720 ppm to 2300 ppm 16 3.1.1.4 page 3/4 1-5 Change HTC less negative than ;2.3x10'#ak/k/*F to less negative than -2.2x10-'ak/k/cF 1

17 3.1.2.2 page 3/4 1-9 Change Shutdown Hargin equivalent from at least 1%Ak/k at 2000F to at least 3%ak/k 1

IS 3.1.2.4 page 3/4 1-11 Change Shutdown Hargin equivalent from at least 1%ak/k at 200 F to at least 3%ak/k 1

19 3.1.2.6 page 3/4 1-13

. Change Shutdown Margin equivalent from at least 1%ak/k at 200cF to at least 3".ak/k 1

20 3.1.2.7 page 3/4 1-14 Change refueling water tank minimum I

borated water volume from 9.978 gallons I

l to 9,844 gallons 21 351.'2.'7page3/41-14 Change refueling water tank boron concentration from 1720 ppm to between 2300 and 2800 ppm.

1

' 22 Figure 3.1-1 Change minimun' boric acid storage tank -

1 page 3/41-15 volume functien 23 3.1.2.8 page 3/4 1-16 Change refueling water tank baron concentration

-from between 1720 and 2200 ppo to between 2300 and 2800ppo and Shutdown Margin equivalent frem 1%ak/k at 200'F to 3%ak/k at 2000F 1

?4 Figure 3.2-1 No change from Cycle 3.

page 3/4 2-3 25' Figure 3.2-2 Replace Fiaure 3.2-2 2

l page 3/4 2-4

.I l

1"'T I

_y

~_

c 7

I i

iL

' A-39 TABLE 9-1 (continued)

Change a Tech Soce A Action Reference 26 Figure 4.2-1, Replace Figure 4.2-1

'l page 3/4 2-5 27 J.2.2 page 3/4 2-6 Change calculated valus f Fxy from s.l.610 to 11.620 and Fxy >1.610 to 2

Fxy >l.620 28 Figure 3.2-3 Replace Figure 3.2-3 2

page 3/4 2-8 T

29

'3.2.3 page 3/4 2-9 Chance calculated value of Fr from

~

<l.540 to <l.620 and change T

FrT>1.540 to Fr >l 42 0 2

2 30 Figure 3.2-4 Replace Figure 3.2-4 2

page 3/4 2-11

31' Table 3.3-1, Add steam generator pressure difference -

page 3/4 3-2 high description to table 1

32 Tsble 3.3-1, Change steam generator pressure-low trip i

page 3/4 3-4 bypass from below 600 psia to belcw 685 psia I

33 Table 3.3-2, Add steam generator pressure difference-l page 3/4 3-6 high response time 1

~34 Table 4.3-1 Add steam generator pressure difference-page 3/4 3-7 high surveillance 1

35 Table 3.3-3, Change Main Steam Line Isolation steam page 3/4 3-15 generator pressure-low trip bypass frcm below 600 psia to below 685 psia 1

36 Table 3.3-4, Change Main Steam Line Isolation steam page 3/4 3-17 generator pressure-low setpoint frcm 1478 psia to 1570 psia

_1 37 Table 3.3-5, Change Containment Purge Isolatfori Valve page 3/4 3-20 Response time from 1 to <5 see 1

6 38 3.5.1 page 3/4 5-1 Change safety injection tank boron concentration from between 172n and 2200 ppm to between 2300 ppm and 2800. ppm.

2 39 3.5.4 page 3/4 5-7 Change refueling water tank boron concentration from between 1720 and 2200 ppm to to between 2300'and 2800 ppa 1

1 if 1

,~

A-40 O

y TABLE 9-1 (continued) -

Channe 3 Tech Srm #

Action

. Reference 40 3.9.1 page 3/4 9-1 Change refueling horon cor. centration of >l720 ppn tv >2300 pp:t ard !. oration -

at [40 spn of 1770 ppn to boration at 3,40 gpu of 2300 ppa and shutdcwn cargiti

^

from In k/k to 3 nk/L i

41..

3.10.1 page 3/ 10-1 Change boration at 340 gpn of 1720 ppo

.)

to horation at >40 spa of 2300 ppm 42 B 3/4.1.1.1 and B 3/4.1.1.2, Chanc,e minimur.1 Shutdown Margin with TecJ Page B 3/4 1-1 y_200 F from 1 2 k/k to 32 k/k and revise basis 1

5 43 8! 3/4.1.2, pa;es Change Shutdo.:n I-targin of 1.02k/k af ter B 3/41-2, B 3/4 1-3 xenon decay and cooldoun to 2000F to 3.0:.:t.k/k af ter xenen decay and ceoidcun to 200 F and the refueline water tank boron concentration from 1720 ppn to

.1 2300 ppa

~ ~.

Change 3313 gallens of 7.25" boric acid solution to 6500 gallons and 47,23a galhns of borated water to 55,627 gallons.

j

i 44 8 3/4.1.2 -page Change 9,978 gallons of borated cater to

.s acid gallons and 439 gallons of 7.253 boric -

9844

-B 3/4 1-3 o 737 gallons.

1 45 B 3/4.2.5, page Change niniren DilBR of 1.19 to minimum 2

B 3/4 2-2 DtWR of 1.195 46 B 3/4.o.1, page Chanqe ninitun baron concentration

~

_j l

B 3/4 9 1 (1720 ppm) to'(2300 ppm)

-47 3.4.1 rage 3/4.4-2 Include spect.fic operation of reactor.

.)

coolant purps for '. ode 3 4

43 3.1.1.2. page 3/4 1-3, Replace paces 3/4 1-3 and B 3/4 l'-1 2

and B 3f4 1.1.1 6 3/4 1.1.2 page 3 J/4;.1-1 49 4.5.2, e.3 and e.4-Change minirun volune of TSP frem 75 rubic.

pg. 3/4 5-5

_ feet to 100 cubic feet and ch.ince sanple volume to 4.010.1 cas-in 3.5,+.1 liters of RWT wicr.

2 w

A-4 TABLE 2 Explanations.for Cycle 4 Tech Spec Changes Change #,

Tech Spec #

Explanatio_n Thermal limit. lines have been changed to 1

Figure 2.1-1 reflect different radial peaking factors.

2 Table 2.2-1 The steam generator pressure-low setpoint is being increased to minimize the consequences of a Steam Line Break Event.

3 Table 2.2-1 A trip for Asymmetric Steam Generator pressure has been added to minimize the consequence of the Loss of Load to One Steam Generator Event.

~

4 Table 2.2-1 The steam generator pressure-low trip bypass has been increased to be consistent with the new trip value.

5

_ ' Figure 2.2-1 The LHR LSSS has been changed to reflect different radial peaking factors.

' ~ ~ ' ~

6 Figure 2.2-2 Reanalysis for Cycle 4 has produced ~no

. changes in TM/LP trip

'7 Figure 2.2-3

,, Reanalysis for Cycle 4 has produced'nb,'

... changes,inTM/LPtryp,

,,e 8

B.2.1.1 The numerical specification of centerline melt limit is being deleted to standardize spec to other C-E plants.

9 B.2.1.1 No changes from Cycle 3.

10 B.2.1, B.2.2' The minimum DNBR has been changed to 1.195.

11 B.2.2.1 No change from Cycle 3.

A-42

~

~~

TABLE 9-2 (continued)

Change #

Tech Soec'#

Explanation 12 B.2.2.1 The basis of the steam generator pressure-low trip setpoint has been ' changed to be

consistent with Table 2.2-1.

13 -

B.2.2.1 The TM/LP basis has been streamlined for clarity and a description of the asymmetric steam generator pressure trip has been added to the bases.

14 3.1.1.1 The shutdown margin has been increased to yield acceptable consequences from a

~

Steam Line Break Event. The new boron concentration is consistent with the new re-fueling water tank concentration for cycle' 4.

15 3.1 l.2 The shutdown margin has been increased to lengthen the operator action time required in'a baron dilution event. The new boron concentration is consistent with the new refueling water tank concentration for Cycle 4 16 3.1.1.'4 -

The most negative MTC permitted for Cycle 4 has been made less negative

~

to yie'id acceptable consequences from a Steam Line Break event.

17 3.1.2.2 The required shutdown margin has been increased to be consistent with Tech Spec 3. l.1. 2.

18 3.1.2.4 The required shutdown margin has been increased to be consistent with Tech Spec 3.l.1.2.

19 3.1.2.6 The required shutdown margin has been increased to be consistent with Tech Spec 3.l.1.2 20 3.1.2.7 The volume of borated water has been decreased due to the higher soluble boron' concentrations.

21 3.1.2.7 The refueling water tank boron concentration has been changed to be consistent with Tech Spec 3.9.1 22 Figure.3.1-l' The volume of borated water has been increased to allow a higher shutdown boron insertion due to the higher core average enrichments of future cycles.

'A-43 TABLE 9-2 (continued)

Change #-

Tech Spec'#'

Explanation 23 3.1.2.8' The refueling water tank boron concentration has been chanced to be 4

consistent with Tech Spec 3'.9.1 and.

the required s_hutdown margin has been increased to.be consistent with Tech Spec 3.1.1.2.-

3 1

24 Figure 3.2-1

.No change from Cycle 3.

-25 Figure 3.2-2 The LHR LCO is,being changed as a result of higher radial peaks.

I i-26 Figure 4.2-1 Augme,ntation factors have been increased to envelgpe; future, cycles p.

27 3.2.2 Radial jeaking factors, both FxyT T

and Fr, are being raised for Cycle 4.

and FrT, peaking factors,"both FxyT Radial 28 Figure 3.2-3 are being' raised for Cycle 4.

2? '

3.2.3 Radial pe'aking factors, both FxyT-

~

T and Fr, are~ be'ing raised for Cycle 4.

30 Figure 3.2-4 The DNB LCO limits are changing due to higher radial peaks.3

' 31, Table-3.3-1 The asymetric steam generator pressure trip has been added to the table.

32' Table 3.3-1 The steam generator pressure-low trip bypass has been increased to be~ consistent with _the new trip value.

33 Table 3.3-2 The asymmetric steam generator pressure trip has been added to the table.

34 Table 4.3-1 The asymmetric steam generator pressure l

trip has been added to the table.

35 Table 3.3-3 The Main Steam Line Isolation steam generator pressure-low trip bypass has been ircreased to be consistent with the new trip value..

36, Ta bl e. 3. 3-4 '.

The Main Steam _ Line' Isolation _ steam

' generator pressure-low trip _setpoint i

has been increased to be. consistent.

J

-with th'e reactor trip setpoint.'

i

A-44

-TABLE 9-2 (continued)

Change #

-Tech Spec #

Explanation

~

Containment isolation value response time 37 Table 3.3-5 is being reduced from 6 seconds to 5 seconds to satisfy NRC requirements. (flRC ' Branch Technical Position CSB 6 41

-38 3.5.1 The safety injection tank boron concentration has been increased to assure a uniform boron concentration in all coolants that have access to the reactor vessel.

39 3.5.4 The' refueling water tank boron concentra-tion has been increased to be consistent with Tech Spec 3.9.1 40 3.9.1 The refueling boron concentrations.have been increased due to the higher core

_,, average enrichment of future cycles an,d the' shutdown margin has increased to be consistent with 3.1.1.2.

41 3.10.1 The boration concentrations have been increased to be consistent with the new boron concentration of the refueling water tank.

42 B 3/4.1.1.1 and The shutdown margins in the bases have B 3/4.1.1.2 been increased to be consistent with

~

~

.those in Tech Specs 3.1.1.1 and 3.1.1.2 and explain applicability of shutdown margin for steam line break accident.

The number of gallons of PPli boron has increased to accommodate increased boron insertion requirements for future cycles.

43 B 3/4.1.2 The shutdown margin has been increased in the bases -to be consistent with Tech Spec 3.1.1.2.

The refueling water tank boron concentretion in the bases has been increased to be consistent' with Tech' Spec 3.9.1 44 8 3/4.1.2 page The volume of barated water.in BAST has B 3/4 1-3 been decreased due to the higher-soluble boron concentration and increa:;ed in'R'.:T due to increased boron. insertion require-ments.

~. _,

- A-4'5 TABLE 9-2 (continued).

Change #

Tech Spec #

Explanation The minimum DNBR has been changed to 1.195.

~45 -

B 3/4.2.5 46 B 3/4.9.1 The refueling water concentration in

-the bases has been increased to be consistent with Tech Spec 3.9.1 47 3.4.1 One-loop no load conditions have not been analyzed for cycle 4 48 3.1.1.2 and Additional requirements to the pressurizer B 3/4 1.1.1 level have been included to increase the i

time to criticality during a boron dilution event.

49 4.5.2 The minimum volume of TSP needed to raise e.3 and e.4 the PH of the borated water of the ECCS

~ to 7.0 is 100 cubic feet.

In order to 1

test the ability of the TSP to raise the i

PH of the borated water of the ECCS, the ratio of. the volume of TSP'to the volume of ECCS borated water must be the same in containment as it is in the laboratory.

I h

9 9

e e

1 y--

ur my-.

r, 4

u-,c y,

,,.e--ww--*

+

ty

-,--w--

~

k A-46 I

' 10.0.Startup Testing 1

' The startup testing program proposed for Calvert. Cliffs II Cycle 4 is identical-to the program proposed for the reference cycle in References j

1 and 14.

~

Y I

4 i

i l

J t

i f.

4 t

.-~

m

-a.-

m r,

-w++

c-e p

y ap..

_.g

=my

..e5-y

p++/

'*4 &'

_ E E__

TEST TARGET (MT-3) 1.0 g a gag y l[l EE I.l

? '= lM l.8 1.25 1.4 1.6 i

[

6"

/

t

+ sh i@f/,N #

1. )e+4

.;h /%

4 7/

n' o

4

A> *?'

N

?/

TEST TARGET (MT-3)

+

1.0 Mmna E [i EE I.I i '* ll%

i 1.8 i

i.25 IA 1.6 4

gn I

W@

%.y@%

s,

s A-47 11.0 References 1.

Letter from A. E. Lundvall to R. A. Clark, September 22, 1980, "Calvert Cliffs Nuclear Power Plant Unit I, Docket 50-317 Amendment to Operating License DPR-53 Sth Cycle License Application" 2.

Letter fran A. E. Lundvall to R. A. Clark, November 4,1980, "Calvert Cliffs Nuclear Power Plant Unit I, Docket 50-317 Amendment to Operating License DPR-53 Supplement 1 to 5th Cycle License Application" 3.

Letter from A. E. Lundvall to R. A. Clark, June 12, 1980, "Calvert Cliffs Nuclear Power Plant, Unit No. 2, Docket No. 50-318 Report of Startup Testing for Cycle Three" 4

Lotter, A. E. Lundvall to R. W. Reid, " Request for Amendment to Operating License, Unit 2, Cycle 2 License Application", dated July 26, 1978 5.

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

dated June 1975 6.

CEN-83(B)-P,"CalvertCliffsUnit1ReactorOperationWithModified CEA Guide Tubes", dated February 8,1979 and letter, A. E. Lundvall, Jr. to V. Stello, Jr., " Reactor Operation With Modified CEA Guide Tubes", dated February 17, 1978 7.

CENPD-139, "C-E Fuel Evaluation Model Topical Report", dated July 1974 8.

CENPD-153-P, Revision 1, " Evaluation of Uncertainty in the Nuclear Poar Peaking Measured by the Self-Powered Fixed In-core Detector System", dated May 1980 9.

Letter from A. E. Lundvall to R. W. Reid, July 11, 1979, " Proposed Finding of No Unreviewed Safety Questions on Unit 2, Cycle 3 Reload Core Design" 10.

Letter A. E. Lundvall to R. A. Clark, December 3, 1980, "Calvert

~

Cliffs Nuclear Power Plant, Unit No. 1, Docket No. 50-317 Amendment to Operating License DPR-53, Fifth Cycle License Application, Responses to NRC Staff Ouestions" 11.

"The Interim SER on Effects of Fuel Rod Bowing in Themal Margin Calculation for Light Water keactors," Rev. 1, Feb. 16, 1977.

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

i

9 o

A-48

13. Letter A. E. Lundvall to R..A. Clark, November'25, 1980, "Calvert Cliffs' Nuclear Power Plant, Unit No. 1, Docket No. 50-317 Amendment

- to Operating License DPR-53, Supplement 2 to Fifth Cycle License Application"

- 14. Letter-A. E. Lundvall to R. A. Clark, November 19 -1980, "Calvert Cliffs Nuclear Power Plant, Unit No. 1, Docket No. 50-317 Amendment to' Operating License DPR-53, Fifth cycle License Application, Responses to NRC Staff. Questions" 4

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