Text

Proposed Tech Specs B2-7 3/4.4 Limiting Safety Sys Settings, Reactivity Control Sys & RCS Reactor Coolant Loops
	ML20032E929
Person / Time
Site:	Crystal River
Issue date:	11/16/1981
From:	; FLORIDA POWER CORP.
To:	;
Shared Package
ML20032E919	List: ML20032E918; ML20032E922; ML20032E929;
References
	TAC-46962, NUDOCS 8111230506
	Download: ML20032E929 (31)
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,I BNd-1684 June 1981 I

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CRYSTAL RIVER UNIT 3

- Cycle 4 Reload Raport -

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N BABC0CK & WILCOX Nuclear Power Group Nuclear Power Generation Divisica P. O. Box 1260 Lynchburg, Virginia 24505 DR,b ht$$$$$h2 Babcock & Wilcox PDR

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CONTENTS Page 1.

INTR 07UCTION AND

SUMMARY

1-1 2.

OPERATING HISTORY........................

2-1 3.

GENERAL DESCRIPTION 3-1 4.

FUEL SYSTEM DESIGN.

4-1 I

4.1.

Fuel Assembl: Mechanical Design 4-1 4.2.

Fuel Rod Design 4-1 4.2.1.

Cladding Collapse 4-1 1

4.2.2.

Cladding Stress 4-1 4.2.3.

Cladding Strain 4-2 4.3.

Fuel Thermal Design 4-2 4.4.

Operating Experience.

4-2 5.

NUCLEAR DESIGN.

5-1 3

5.1.

Physics Characteristics 5-1 g

5.2.

Changes in Nuclear Design 5-2 6.

THERMAL-HYDRAULIC DESIGN.

6-1 7.*. ACCIDENT AND TRANSIENT ANALYSIS 7-1 i

7.1.

General Safety Analysis 7-1 1

7.2.

Accident Evaluation 7-1 7.3.

Dose Consequences of Accidents.

7-2 8.

PR0l'OSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS.

8-1 9.

STARTUP PROGRAM - PHYSICS TESTING 9-1 1

9.1.

Precritical Tests 9-1 9.1.1.

Control Rod Trip Tent 9-1 9.1.2.

RC Flow 9-1 1

9.2.

Zero Power Physics Tests.

9-2 9.2.1.

Critical Boron Concentration.

9-2 9.2.2.

Temperature Reactivity Coefficient 9-2 g

9.2.3.

Control Rod Group Reactivity Worth.

9-2 g

9.2.4.

Ejected Control Rod Reactivity Worth.

9-3 1

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Babcock & Wilcox

I CONTENTS (Cont'd)

Page 9.3.

Power Escalation Tests.

9-3 9.3.1.

Core Power Distribution Verification at

%40, 75, and 100% FP With Nominal l

Control Rod Position.

9-3 9.3.2.

Incore Vs Excore Detector Imbalance Correla-tion Verification at 440% FP.

9-5 g

9.3.3.

Temperature Reactivity Coefficient at %100% FP.

9-5 5

9.3.4.

Power Doppler Reactivity Coefficient at N100% FP.

9-5 9.4.

Procedure for Failure to Meet Acceptance criteria 9-6 REFERENCES.

A-1 I

List of Tables Table 4-1.

Fuel Rod Design Parameters 4-3 g

4-2.

Fuel Thermal Analysis Parameters 4-4 g

5-1.

Physics Parameters, Crystal River 3, Cycles 3 and 4......

5-3 5-2.

Shutdown Margin Calculation for Crystal River 3, Cycle 4 5-5 6-1.

Thermal-Hydraulic Design Conditions.

6-2 7-1.

Comparison of Key Parameters for Accident Analysis 7-3 7-2.

LOCA Limits, CR-3, Cycle 4 After 50 EFPD 7-4 7-3.

LOCA Limits, CR-3, Cycle 4, 0-50 EFPD.

7-4 g

8-1.

Technical Specification Changes.

8-2 5

8-2.

RPS Trip Setpoints 8-6 8-3.

Reactor Protection System Instrumentation.

8-32 8-4.

Reactor Protection System Instrumentation Response Times 8-33 3-5.

Reactor Protection System Instrumentation Surveillance Requirements.

8-34 I

List of Figures Figure 3-1.

Core Loading Diagram for Crystal River 3 Cycle 4...

3-2 3-2.

Enrichment and BOC Burnup Distribution for Crystal River 3. Cycle 4 3-3 3-3.

Control Rod Locations and Group Designations for l

Crystal River 3, Cycle 4 3-4 5

3-4.

Crystal River 3. Cycle 4 BPRA Enrichment and Distribution.

3-5 5-1.

BOC (4 EFPD), Cycle 4 Two-Dimensional Relative Power g

Distribution - HFP, Equilibrium Xenon, Banks 7 and g

8 Inserted 5-6 8-1.

Reactor Core Safety Limits 8-4 8-2.

Reactor Core Safety Limits 8-5 Babcock & Wilcox

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Figures (Cont'd)

F Figure Page 8-3.

Reactor Trip Setpoints E-7 8-4.

Pressure / Temperature Limits.

8-11 8-5.

Regulating Rod Group Insertion Limits for Four-Pump Operation From 0 to 50 EFPD.

8-18 8-6.

Regulating Pod Group Insertion Limits for Four-Pump Operation From 50 ! 10 to 270 ! 10 EFPD.

8-19 8-7.

Regulating Rod Group Insertion Limits for Four-Pump Operation After 270 ! 10 EFPD.

8-20 8-8.

Regulating Rod Group Insertion Limits for Three-Pump Operation From 0 to 50 EFPD.

8-21 8-9.

Regulating Rod Group Insertion Limits for Three-Pump Operation From 50 to 270 ! 10 EFPD 8-22 I

8-10.

Regulating Rod Group Insertion Limits for Three-Pump Operation After 270 ! 10 EFPD.

8-23 8-11.

APSR Position Limits for 0 to 50 EFPD.

8-25 8-12.

APSR Position Limits for 50 ! 10 to 270 10 EFPD.

8-26 I

8-13.

APSR Position Limits After 270 ! 10 EFPD 8-27 8-14.

Axial Power Imbalance Envelope for Operation From 0 to 50 ! 10 EFPD.

8-29 I

8-15.

Axial Power Imbalance Envelope for Operation From 50 ! 10 to 270 10 EFPD 8-30 8-16.

Axial Power Imbalance Envclape for Operation After 270 10 EFPD.

8-31 I

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lI II 1.

INTRODUCTION AND

SUMMARY

I Th '.s report justifies the operation of Crystal River Unit 3 (cycle 4) at a

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rated core power level of 2544 MWt.

Included are the required analyses to support cycle 4 operation; these analyses employ analytical techniques and design bases established in reports that have received technical approval by the USNRC (see references).

The design for cycle 3 raised the rated thermal power from 24S2 to 2544 MWt which was the ultimate core power level identified in the Crystal River Unit 3 FSAR.1 The cycle 4 core has been designed with an increased cycle lifetime cd 350 ef fective full power days (EFFDs) and the incorporation of burnable prison rod assemblies (BPRAs) to aid in reactivity control.

The Technical Specifications have been reviewed, and the modifications for cycle 4 are justified in this report.

Based on the analyses performed, which take into account the postulated ef-fects of fuel densification and the Final Acceptance Criteria for emergency core cooling (ECCS), it has been concluded that Crystal River 3, cycle 4, can be safely operated at a core power level of 2544 MWt.

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2.

OPERATING HISTORY I

Cycle 3, the current Crystal River Unit 3 operating cycle, is the reference fuel cycle for the nuclear and thermal-hydraulic analysis performed for cycle 4 operation.

Cycle 3 achieved criticality on August 8,1980, and is scheduled for completion on September 19, 1981, after approximately 320 EFPD. During cycle 3 the core power level was upgraded to 2544 MWt from 2452 MWt.

No op-erating anomalies have occurred during previous cycle operations that would adversely affeet fuel performance in cycle 4.

Cycle 4 is scheduled to start operation in December 1981 at a rated power level of 2544 MWt.

The design cycle length is 350 EFPD.

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I 3.

GENERAL DESCRIPTION I

The Crystal River Unit 3 reactor core is described in detail in Chapter 3 of the Final Safety Analysis Report for the unit.1 The cycle 4 core consists of 177 fuel assemblies (FAs), each of which is a 15-by-15 array containing 208 fuel rods; 16 control rod guide tubes; and one incore instrument gt.ide tube.

The fuel assemblies in batches 5 and 6 have an average nominal fuel loading of 463.6 kg of uranium; whereas the batch 4 assemblies maintain an average nomi-nal fuel loading of 468.6 kg of uranium. The cladding is cold-worked Zircaloy-4 with an outside diameter (OD) of 0.430 inch and a wall thickness of 0.0265 inch.

The fuel consists of dished-end, cylindrical pellets of uranium dioxide (see I

Table 4-2 for data).

Figure 3-1 is the core loading diagram for cycle 4 of Crystal River 3.

The I

initial enrichments of batches 4 and 5 were 2.64 and 2.62 wt % uranium-235, respectively. Nine of the batch 2B assemblies, sixty of the batch 3 assem-blies, and three of the batch 4B assemblies will be discharged at the end of cycle 3.

The design enrichments of batches 6A and 6B are 2.62 and 2.95 wt %

uranium-235, respectively.

Four batch 4A assemblies were reinserted. These were prematurely discharged at the end of cycle 2 after one cycle of operation because assembly NJ018E had a broken holddown spring. The spring is rurrently scheduled to be replaced prior to cycle 4 fuel loading. The batch 4C and 5 I

assemblies will be shuffled to new locations with batch 5 on the periphery.

The fresh batch 6A and 6B assemblies will be loaded into the core interior in a symmetric checkerboard pattern.

Figure 3-2 is an eighth-core map showing the burnup of each assembly at the beginning of cycle 4 and its initial en-richment.

Core reactivity is controlled by 61 full-length Ag-In-Cd control rod assemblies (CRAs), 64 burnable poison rod assemblies (BPRAs), and soluble boron shim.

In addition to the full-length CRAs, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The cycle 4 locations of the 69 control rods and the group designations are indicated 3-1 Babcock & Wilcox

I in Figure 3-3.

The cycle 4 locations and enrichments of the BPRA clusters are shown in Figure 3-4.

Control cod group 7 will be withdrawn at 270 10 EFPD of operation.

I Figure 3-1.

Core Loading Diagram for Crystal River 3, Cycle 4 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 A

5 5

5 l

R9 N2 R8 N14 R7 5

6A 5

6A 4C 6A 5

6A 5

M14 RIO C9 R6 M2 l

5 6B 5

6A 4C 6B 4C 6A 5

6B 5

C 013 L2 E7 E9 L14 03 g

5 6B 4C 6A 4C 6A 4C 6A 4C 6A 4C 6B 5 ~

D Pil M12 L3 F8 L13 M4 P5 E

6A 5

6A 4C 6A 4C 4A 4C 6A 4C 6A 5

6A E

E B10 C7 D6 D(

D10 G13 B6 g

'C 6A 4C 6A 4C 6A 5

5 4

F 5

5 6A 4C 6A 4C 6A K15 L15 C10 D5 D8 D11 C6 L1 K1 g

5 6A 4C 6A 4C 6A 5

5 5

6A 4C 6A 4C 6A 5

B12 G5 F4 C4 C12 013 F12 Gil B4 5

4C 6B 4C 4A 4C 5

4C 5

4C 4A 4C 6B 4C 5

q," _

H10 K13 HI

-Y g

HIS G3 H6 42, H4 D3 K9 N13 H12 5

6A 4C 6A 4C 6A 5

5 5

6A 4C 6A 4C 6A 5

g K

P12 K5 L4 N3 04 012 L12 K11 P4 5

5 5

6A 4C 6A 4C 6A 4C 6A 4C 6A 4C 6A 5

5 t

GIS FIS 010 N5 N8 N11 06 F1 G1 6A 5

6A 4C 6A 4C 4A 4C 6A 4C 6A 5

6A M

PIO K3 N6 N14 N10 09 P6 (Cv2) 5 6B 4C 6A 4C 6A 4C 6A 4C 6A 4C 63 5

Bil E12 F3 L8 F13 E4 B5 5

6B 5

6A 4C 6B 4C 6A 5

68 5

0 C13 F2 M7 M9 F14 C3 5

6A 5

6A 4C 6A 5

6A 5

p E14 A10 07 A6 E2 5

5 5

g R

g A9 D2 A8 D14 A7 l

I Batch Previous core location l

3-2 Babcock & Wilcox

I Figure 3-3.

Control Rod Locations and Group Designations for Crystal River 3, Cycle 4 I

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2 3

4 5

6 7

8 9

10 11 12 13 14 15 X

A B

4 7

4 C

2 5

5 2

I D

6 8

3 8

6 E

2 3

1 1

3 2

F 4

8 i 7 6

7 8

4 G

5 1

3 3

1 5

H b'---

7 3

6 4

6 3

7

-Y K

5 1

3 3

1 5

L 4

8 7

6 7

8 4

M 2

3 1

1 3

2 N

6 8

3 8

6 0

2 5

5 2

P 4

7 4

R I

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X Group Number No. of Group rods Function 1

8 Safety 2

8 Safety 3

12 Safety 4

9 Safety 5

8 Control 6

8 Control 7

8 Control 8

8 APSRs Total 69 I

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Figure 3-4.

Crystal River 3, Cycle 4 BPRA Enrichment and Distribution l

8 9

10 11 12 13 14 15 l

H 0.2 0.5 I

K 0.5 0.5 0.2 L

0.5 0.2 0.2 I

M 0.2 0.2 0.5 I

~

N 0.5 0.5 0.2 0

0.5 0.2 0.2 P

0.2 R

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X.X HPRA Concentration, wt % bgg in Al203 1

3-5 Babcock & Wilcox l

_------.-----.------_---_-------------J

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l 4.

FUEL SYSTEM DESIGN l

8 4.1.

Fuel Assembly Mcchanical Design The types of fuel assemblies (FAs) and pertinent fuel design parameters for Crystal River Unit 3, cycle 4 are listed in Table 4-1.

All FAs are identical in concept and are interchangeable.

i Retainer assemblies will be on 64 assemblies which contain BPRAs and two as-semblies which contain regenerative neutron sources. The justification for the design and use of the retainers is described in references 3 and 4.

=

I 4.2.

Fuel Rod Design The fuel rod design batch 6 fuel is identical to the batch 5 fuel used in cycle 3 (see Table 4-1).

The mechanical evaluation of the fuel rod is dis-cussed below.

4.2.1.

Cladding Collapse The batch 4 fuel is more limiting than batches 5 and 6 due to previous incore exposure time.

The batch 4 assembly power histories were analyzed to deter-mine the most limiting three-cycle power history for creep-collapse. The worst case power history was then compared against a generic analyses to ensure that l

creep-ovalization will not affect fuel performance during Crystal River 3, cycle 4.

The generic analyses were performed based on reference 5 and are applicable to batch 4 design.

The creep-collapse analyses predict a collapse time greater than 35,000 effec-tive full power hours (EFPH), which is greater than the maximum expected resi-dence time of 25,200 EFPH (Table 4-1).

4.2.2.

Cladding Stress The Crystal River 3, cycle 4 stress parameters are enveloped by a conservative fuel rod strees analysis. There was no new method used for the analysis of this cycle than that used on previous cycles.

4-1 Babcock & Wilcox

I 4.2.3 Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding plastic tensile circumferential strain. The pellet is designed to assure that cladding plas-tic strain is less than 1% at design local pellet burnup and heat generation rate. The design burnup and heat generation rate are higher than the worst case values cycle 4 is expected to see.

The strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density and the 5

lower colerance value for the cladding inside diameter (ID).

4.3.

Fuel Thermal Design All FAs in this core are thermally similar. The fresh batch 6 fuel inserted I

for cycle 4 operation introduces no significant differences in fuel thermal 5

performance relative to the fuel remaining in the core. The design minimum linear heat rate (LHR) capability and the average fuel temperature for each batch in cycle 4 are shown in Table 4-2.

The maximum fuel assembly burnup at end of cycle 4 (EOC-4) is predictei to be 33,458 mwd /mtU. Fuel rod internal 13 pressure has been evaluated with TAFY-3 for the highest burnup fuel rod and was predicted to be less than 2200 psia.

4.4.

Operating Experience Operating experience with the Mark-B 15-by-15 FA has verified the adequacy of its design. As of April 30, 1981, the following experience has been accumu-lated for the eight B&W 177-FA plants using the Mark-B fuel assembly:

Max.

burnup' Cumulative net Current electric output, Reactor cycle Incore Discharged MWh Oconee 1 6

28,250 40,000 35,591,479 Oconee 2 5

30,800 33,700 30,343,260 Oconee 3 6

26,440 32,061 30,007,244 TMI-1 5

32,400 32,200 23,840,053 ANO-1 5

22,300 33,222 26,633,490 Rancho Seco 5

27,684 37,730 25,165,850 Crp tal River 3 3

23,122 22,389 14,948,242 Davis L~sse 2

20,785 13,252 9,274,871

(^ As of April 30, 1981.

}As of March 31, 1981.

4-2 Babcock & Wilcox

Table 4-1.

Fuel Rod Design Parameters Batch 4A/4C 5

6A 6B Fuel assembly type Mark-B4 Mark-B4 Mark-B4 Mark-B4 Number of assemblies 4/49 56 56 12 Fuel rod OD, in.

0.430 0.430 0.430 0.430 Fuel rod ID, in.

0.377 0.377 0.377 0.377 Flexible spacer type Spring Spring Spring Spring

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Rigid spacer type Zirc-4 Zirc-4 Zirc-4 Zirc-4 Undensified active fuel 143.6 141.79 141.80 141.80 length, in.

Fuel pellet (mean specified) 0.3697 0.3686 0.3686 0.3686 diameter, in.

Fuel pellec initial density 94.0 95.0 95.0 95.0

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(mean specified), % AT Initial fuel enrichment, 2.64 2.62 2.62 2.95 v % * 'U Estimated residence time, 19,920 24,200 25,200 25,200 EFPH Cladding collapse time,

>35,000

>35,000 EFPH 4-3 Babcock & )Milcox

Table 4-2.

Fuel Thermal Analysis Parameters Batch 4A/4C 5

6A/6B No. of asuemblies 4/49 56 56/12 Nominal pellet density, % TD 94.0 95.0 95.0 Pellet diameter, in.

0.3697 0.3686 0.3686 Stack height, in.

143.6 141.8 141.8 Densified Fuel Parameters ("

Pellet diameter, in.

0.3648 0.3649 0.3649 Fuel stack height, in.

141.8 140.7 140.7 W

Nominal LHR at 2568 MWt, kW/ft 5.7 5.8 5.8 Avg fuel temper ature at 1280 1310 1310 nominal LHR, F LHR to centerline fuel melt, 20.1 20.1 20.1 g

kW/ft g

Core average densified LHR at 2544 MWt is 5.7 kW/ft

" Densification to 96.5% TD assumed.

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!:I Table 4-1.

Fuel Rod Design Parameters Batch 4A/4C 5

6A 6B Fuel assembly type Mark-B4 Mark-B4 Mark-B4 Mark-B4 Number of assemblies 4/49 56 56 12 Fuel rod OD, in.

0.430 0.430 0.430 0.430 Fuel rod ID, in.

0.377 C.377 0.377 0.377 Flexible spacer type Spring Spring Spring Spring Rigid spacer type Zirc-4 Zirc-4 Zirc-4 Zirc-4

!g Undensified active fuel 143.6 141.79 141.80 141.80

!,g length, in.

j Fuel pellet (mean specified) 0.3697 0.3686 0.3586 0.3686 j

diameter, in.

Fuel pellet initial density 94.0 95.0 95.0 95.0 (mean specified), % AT Initial fuel enrichment, 2.64 2.62 2.62 2.95 23s i

wt %

U Estimated residence time, 19,920 24,200 25,200 25,200

I EFPH Cladding collapse t'

>35,000

.I EFPH T

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Table 4-2.

Fuel Thermal Analysis Parametecs Batch 3 /4C 5

6A/6B No. of assemblies 4/49 56 56/12 Nominal pellet density, % TD 94.0 95.0 95.0 Pellet diameter, in.

0.3697 0.3686 0.3686 Stack height, in.

143.6 141.8 141.8 Densified Fuel Parameters Pellet diameter, in.

0.3648 0.3649 0.3649 Fuel stack height, in.

141.8 140.7 140.7 el Nominal LHR at 2568 MWt, kW/ft 5.7 5.8 5.8 Avg fuel temperature at 1280 1310 1310 nominal LHR, F LHR to centerline fuel melt, 20.1 20.1 20.1 g

kW/ft g

Core average densified LHR at 2544 MWt is 5.7 kW/ft l

( }Densification to 96.5% TD assumed.

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5.

NUCLEAR DESIGN 5.1.

Phyaics Characteristics Table 5-1 compares the core physics parameters of cycles 3 and 4; these values were generated using PDQ07 for both cycles. The differential cycle burnup 7

will be larger for cycle 4 than for cycle 3 because of the longer cycle 4 length.

Figure 5-1 illustrates a representative relative power distribution for the be-ginning of cycle 4 at full power with equilibrium xenon and nominal rod posi-tions.

Although both cyples 3 and 4 are rodded cycles with a transient bank withdrawal near EOC, the initial BPRA loading, longer design life, and different shuffle pattern for cycle 4 make it difficult to compare the physics parameters with those of cycle 3.

The critical boron concentrations for cycles 3 and 4 are given in Table 5-1.

The cycle 4 design length was set slightly beyond predicted depletion of reactivity (17 ppm) to ensure the cycle would be licensed to actual depletion of reactivity. The control rod worths dif fer between cycles due to changes in radial flux and burnup distributions. Calculated ejected rod worths and their adherence to criteria are considered at all times in life ar.d at all power 1cvels in the development of the rod position limits presented in section 8.

The maxf2num stuck rod worth for cycle 4 is less than that for design cycle

[

3 at BOC and greater at EOC. The adequacy of the shutdown margin with cycle 4 stuck rod worths is demonstrated in Table 5-2.

The following conservatisms

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were applied for the shutdown calculations:

1.

Poison material depletion allowance.

2.

10% uneartainty on net rod worth.

3.

Flux redistribution penalty.

Flux redistributicn was accounted for since the shutdown analysis was calculated using a two-dimensional model. The shutdown calculation at the end of cycle 3 was analyzed at 270 EFPD. This is the latest time ( 10 EFPD) in core life at which the transient bank is nearly fully inserted. After 270 EFPD, the

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5-1 Babcock & Wilcox

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transient bank will be almost fully withdrawn, thus, the available shutdown margin will be increased.

The cycle 4 power deficits, differential boron worths, and effective delayed neutron fractions differ from those for cycle 3 due to the presence of burnable poison and the longer cycle length.

5.2.

Changes in Nuclear Design 1

There is only one significant core design change between the reference and re-u load cycles. The change is the increase in cycle lifetime to 350 EFPD and the accompanying use of BPRAs to aid in reactivity control. The calculational methods and design information used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle.

No significant operational or procedural changes exist with regard to axial or radial power shape control, xenon control, or tilt cont ol.

The operational limita and RPS limits (Technical Specification changes) for cycle 4 are pre-sented in section 8.

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Table 5-1.

Physics Parameters, Crystal River 3 Cycles 3 and 4 Cycle 3 Cycle 4("}

Design cycle length, EFPD 335 350 Design cycle burnup, mwd /mtU 10,345 10,814 Design average core burnup - EOC, mwd /mtU 17,922 17,737 I

Design initial core loading, mtU 82.3 82.3 Critical boron - BOC, ppm (no Xe)

I HZP(b), group 8 inserted 1,430 1,374 HZP, groups 7 and 8 inserted 1,351 1,277 HFP (D ), groups 7 and 8 inserted 1,185 1,090 I

Critical boron - EOC, ppm (eq Xe)

HZP) 287 298 HFPJgrup8 inserted 27 2

Control rod worths - HFP, BOC, % Ak/k Group 6 1.09 1.11 Group 7 0.83 0.99 Group 8 0.48 0.41 Control rod worths - HFP, EOC, % Ak/k Group 7 1.07 (c) 1.23(d)

Croup 8 0.48(")

0.44 (d)

Max. ejected rod worth (*

- HZP, % Ak/k BOC (N-12) 0.53 0.51 EOC (N-12) 0.59(")

0.54 (d)

Max. stuck rod worth - HZP, % Ak/k BOC (M-13) 1.76 1.54 EOC (H-14) 1.84( )

2.14 (d)

Power deficit, HZP to HFP, % Ak/k BOC

-1.30

-1.44 I

EOC

-2.12

-2.36 Dopph r coef f s (Ak/k/*F)

BOC, 100% power, no Xe

-1.52

-1.55 I

EOC, 100% power, eq Xe

-1.61

-1.71 d

Moderator coeff - HFP, 10-" (Ak/k/*F)

I BOC (0 Xe, critical ppm, group 8 inserted)

-0.30

-0.52 EOC (eq Xe, 17 ppm, group 8 inserted)

-2.63

-2.89 Boron worth - HFP, ppm /% Ak/k I

BOC, 1150 ppab 108 111 EOC, 17 ppmb 94 99 Xenon worth - HFP, % A k/k I

BOC (4 EFPD) 2.63 2.63 EOC (equilibrium) 2.74 2.72 5

5-3 Babcock & Wilcox

-I Table 5-1.

(Cont'd)

Cycle 3 Cycle 4("}

Effective delayed neutron fraction - HFP BOC 0.00597 0.00628 g

EOC 0.00519 0.00523 3

(") Cycle 4 data are for the conditions stated in this report; the cycle 3 values given are at the core conditions identified in reference 2.

(b)11ZP denotes hot zero power (532F T

); HFP denotes hot full power g

(579F Tavg)*

(

Rod worths for EOC-3 are calculated at 250 EFPD, the latest time in core life at which the transient bank is nearly full-in.

(

Ro3 uorths for E0C-4 are calculated at 270 EFPD, the latest time in core life at which the transient bank is nearly full-in.

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Ejected rod worth for groups 5 through 8 inserted.

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Table 5-2.

Shutdown Margin Calculation for Crystal River 3, Cycle 4 BOC, % Ak/k EOC(a), % Ak/k Available Rod Worth Total rod worth, HZP 9.58 9.75 I

Worth reduction due to burnup of poison material

-0.42

-0.42 Maximum stuck rod worth, HZP

-1.54

-2.14 I

Net worth 7.62 7.19 Less 10% uncertainty

-0.76

-0.72 Total available worth 6.86 6.47 Required Rod Worth Power deficit, HFP to HZP 1.44 2.23 Ma:t allowable inserted rod worth 1.17 1.50 t

Flux redistribution 0.50 1.02 Total required worth 3.11 4.75 Shutdown Margin l

Total ava13able minus total required 3.75 1.72 I

Note: Required shutdown margin is 1.00% Ak/k.

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(# For shutdown margin calculations, this is defined as 270 EFPD, the latest time in core life at which the transient bath is nearly full-in.

I HZP: hot zero power, HFP: hot full power.

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/igure 5-1.

BOC (4 EFPD), Cycle 4 Two-Dimensio6..il Relative i

Power Distribution - HFP, Equilibrium Xenon, Banks 7 a'r! 8 Inserted 8

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12 13 14 15 X

I H

1.18 1.28 1.23 1.28 1.14 1.16 0.51 0.45 I

K 1.28 1.22 1.18 1.22 1.08 0.93 0.52 M

L 0.72 1.19 1.01 1.24 1.02 0.46 M

1.16 1.24 1.16 0.91 I

N 1.13 1.10 0.54 I

O 0.63 P

R I

l l

I j

X Inserted rod group Fo.

X.XX Relative power density I

I I

5-6 Babcock & Wilcox

I I

I 6.

THERMAL-HYDRAULIC DESIGN The thermal-hydraulic design evaluation supporting cycle 4 operation used the methods ind models described in references 1, 2, and 6.

The incoming batch 6 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The cycle 3 and 4 maximtr.: design conditions and significant parameters are shown in Table 6-1.

The minimum DNBR for cycle 4 is based on 106.5% of reactor coolant design flow, 8.1% maximum core bypass flow, a 1.71 reference design radial x local peaking factor, and includes the I

ef fects of in-core fuel densification.

A red bow penalty has been calculated according to the procedure approved in reference 8.

The burnup used is the maximum fuel assembly burnap of the batch that contains the limiting (maximum radial x local peak) fuel assembly. For I

cycle 4, t.his burnup is 33,458 mwd /mtU in a batch 4C assembly.

The resulting net rod bow penalty af ter inclusion of the 1% flow area reduction factor credit is 3.5% DNBR reduction.

I I

I 6-1 Babcock & Wilcox

I Hble 6-1.

Thermal-Hydraulic Design Conditions Cycle 3 Cycle 4 2544 MWt 2544 MW:

Design power level, MWt("}

2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 Reference design radial x local power peaking factor, FAH 1.71 1.71 Reference design axial flow shape 1.5 cosine 1.5 cosine Hot channel factors Enthalpy rise 1.011 1.011 Eeat flux 1.014 1.014 Flow area 0.98 0.98 Densified active length, in.(")

140.2 140.2 Average heat flux at 100% power, 2

3 Btu /h-ft 176 x 10 176 x 10' Maximum heat flux at 100% power, 2

8 3

Btu /h-ft 452 x 10 432 x 10 CHF correlation BAW-2 BAW-2 Minimum DNBR, % power 1.98(112) 2.05(112)

(" Used in analysis.

I I

I l

I 1

I I

I 6-2 Babcock & Wilcox

I I

7.

ACCIDENT AND TRANSIENT ANALYSIS I

7.1.

General Safety Analysis 1

Each FSAR accident analysis has been examined with respect to changes in cy-I cle 4 parameters to determine the ef fect of the cycle 4 reload and to ensure that thermal performance during hypothetical transients is not degraded.

The effects of fuel densification on the FSAR accident results have been eval-uated and are reported in reference 6.

Since batch 6 reload FAs do not contain I

fuel rods whose theoretical density is lower than those considered in the ref-crence 6 report, the conclusions in that reference are still valid with the exception of the four-pump coastdown and locked-rotor accident which were re-evaluated at 102% of 2568 MWt for the previc e cycle of operation and remain valid for cycle 4.

7.2.

Accident Evaluation The key parameters that have the greatest ef fect on dotarmining t he outcome of a transient can typical.ly be classified in three :.njor areas: core thermal parameters, thermal-hydraulic parameters, and kine t to parameters, including the reactivity feedback coefficients and control rod worths.

Core therma ~ properties used in the FSAR accident analysis were design operat-ing values based on esN1ational values plus uncertainties. The cycle 4 thermal-hydraulic maximum design conditions are compared to the previous cy-cle 3 values in Table 6-1.

These parameters are common to all the accidents considered in this report. A comparison of the key kinetics parameters from the FSAR and cycle 4 is provided in Table 7-1.

Generic LOCA analyses for B&W 177-FA lowered-loop NSSs have been performed using the Final Acceptance Criteria ECCS Evaluation Model. The large-break analysis is presented in a topical report', and is further substantiated in I

a letter report The small break analysis is presented in a letter report.ll l8 These analyses used the limiting values of key parameters for all plants in 7-1 BabC0Ck & Wilcox

I I

the category.

Furthermore, the average fuel temperature as a function of LHR and lifetime pin pressure data used in the LOCA limits analysis' are conser-vative compared to those calculated for this reload. Thus, these analyses and LOCA limits provide conservative results for the operation of Crystal m

River Unit 3, cycle 4.

Crystal River Unit 3's proposed long-term ECCS modification for small break g

W LOCA is presented in reference 12.

The LOCA analysis uced a power level of 2772 Wt, which is conservative rela-I

=

tive to the 2544 Wt rating. Table 7-2 shows the bounding values for allow-able LOCA peak LHRs for Crystal River Unit 3, cycle 4 fuel after 50 EFPDs.

The LOCA kWift limits have been reduced for the first 50 EFPDs. The reduction g

will ensure that conservative limits are maintained while a transition is being made in the fuel performance codes that provide input to the ECCS analy-sis" in order to account for mechanistic fuel densification. The limits for the first 50 EFPDs are shown in Table 7-3.

It is concluded f rom the examination of cycle 4 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that thic core re-load will not adversely affect the Crystal River Unit 3 plant's ability to operate safely during cycle 4.

Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cy-cle 4 is bounded by previously accepted analyses. The initial conditions for the transients in cycle 4 are bounded by the FSAR with the exception of the four-pump coastdown and locked-rotor accidents which were re-evaluated for I

cycle 3 operation at a core power of 102% of 2568 Wt.

l h3. Dose Consequences of Accidents 2

A complete dose evaluation was presented in the cycle 3 reload report based on the ':pgraded power level of 2544 Wt.

A complete dose evaluation for cy-cle 4 was not perfcrmed because the cycle 4 doses should be essentially the 5

l same ac t b doses presented in the cycle 3 reload report. This conclusion is based on a comparison of the fission product inventory in the cycle 3 and 4 cores, which were calculated using the appropriate fuel data for the respec-tive cycles. The activities of the xenon, krypton, and iodine nuclides (which are the nuclides that control the accident doses) in both cycles 3 and 4 were essentially the same.

The ratio of cycle 4 to cycle 3 activities ranged from 7-2 Babcock s.Wilcox t

0.985 to 1.005 for various nuclides. Thus, the cycle 4 doses are essentially the same as the doses indicated in Tables 7-7 and 7-8 of the cycle 3 reload report.2 I

Table 7-1.

Comparison of Key Parameters for Accident Analys g 1

FSAR,

Predicted I

densif'n cycle 4 Parameter value' Cycle 1 value 15 I

EOL Doppler coeff, 10-5 ak/k/ F

-1.17

-1.47

-1.55 (268 EFPD)

EOL Doppler coeff, 10-s ak/k/*F

-1.30

-1.66

-1.71 (510 EFPD)

BOL moderator coeff, 10-" ak/k/*F 0(^)

-0.75

-0.52 (268 EFPD)

I

-4.0(b)

-2.42

-2.89 EOL moderator coeff, 10-' ak/k/*F (510 EFPD)

I All-rod bank worth at BOL, HZP, 12.9 9.12 9.583

% ak/k (268 EFPD)

Boron reactivity worth (HFP),

100 101 111 ppm /1% Ak/k Max. ejected rod worth (HFP), % ok/k 0.65 0.55 0.564 I

Dropped rod worth (HFP), % Ak/k 0.40 0.20 0.20 Initial boron conc'n (HFP), ppm 1150 795 1090 I

+0.50 x 10-" ak/k/*F was used for the moderator dilution accident.

-3.0 x 10-4 Ak/k/*F was used for the steam line failure analysis and dropped rod accident analysis.

I I

7-3 Babcock & Wilcox

I Table 7-2.

LOCA Limits, CR-3, Cycle 4 After 50 EFPD Elevation, LHR limits, ft kW/ft 2

15.5 4

16.6 6

18.0 8

17.0 10 16.0 I

Table 7-3.

LOCA Limits, CR-3, Cycle 4, g

0-50 EFPD E

Elevation, LHR limits, g

ft kW/ft g

2 14.5 4

16.1 6

17.5 8

17.0 10 16.0 I

I I

I i

I I

I 8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS I

All technical specifications have been reviewed by Fbrida Power Corpcration and B&W and revisions were made to accommodate cycle 4 operation and r1 vised RPS instrument errors.

Table 8-1 lists the Technical Specif k ?* ion thv.;es and cross-references this reload report numbers with the Teci nical spec 3 fits _-

tion numbers.

The re-analysis of Technical Specification for cycle 4 operation used tue same analytical techniques as the cycle 3 design.2 The 0-50 EFPD operating limits on rod index, APSR position, and axial power imbalance were established based I

on the interim LOCA linear heat rate limits which account for mechanistic fuel densification. " After 50 EFPD, the FAC LOCA LHR limits were used. The Tech-nical Specifications also provide protection for the over power condition that could occur during an overcoaling transient because of nuclear instrumentation errors.

The review of the Technical Specifications based on the analysis presented in thir. report and the proposed modifications contained in this section, ensure that the Final Acceptance Criteria ECCS limits will not be exceeded nor will the thermal design criteria be violated.

I I

I 8-1 Babcock & Wilcox

I Table 8-1.

Technical Specification Changes Tech Spec No.

Repcrt page Nos.

l W

(figure, table Nos.)

Reason g r change (Figure 2.1-1)

S-4 Revised for cycle 4 opera-(Figure 8-1) tion.

(Figure 2.1-2) 8-5 Revised for cycle 4 opera-(Figure 8-2) tion.

(Table 2.2-1) 8-6 Added nuclear overpc,wer (Table 8-2) based on RCPPM trip, revised setpoints to include RPS instrument error and cycle 4 reload.

(Figure 2.2-1) 8-7 Revised for cycle 4 opera-(Figure 8-3) tion.

2.1.1, 2.1.2 8-8 Revised text to indicate three-pump operation is more i

restrictive on pressure / temp g

limit curve (Figure 2-1) 3 2.2.1 8-9 Revised nuclear overpower 8-10 trip value to include RPS instrument error. Added nuclear overpower trip based on RCPPMs.

(Figure 2.1) 8-11 Revised press / temp limit (Figure 8-4) curve.

3.1.1.1.2 8-12 Shutdown margin was revised 4.1.1.1.2.1 to reflect the required mode 4 and 5 margin for cycle 4 operation to account for the inadvertant dilution by sodium hydroxide addition.

3.1.2.2 8-13 Revised mode 4 shutdawn margin as above.

3.1.2.4.2 8-14 Revised mode 4 shutdown margin above.

3.1.2.7 8-15 Revised mode 4 shutdown margin as above.

3.1.2.9 8-16 Revised borated storage water volume and mode 4 shutdown margin as above 8-2

I I

13.ble 8-1.

(Cont'd)

I Tech Spec No.

Rep'irt page Nos.

(figure, table Nos.)

Rea soit for change 3.1.3.6 8-17 Revise figure numbers.

8-18 thru 8-23 Revised regulating rod group (Figure 3.1-1)

(Figure 8-5) limits for three-and four-I (Figure 3.1-la)

(Figure 8-6) pump operation for cycle 4.

(Figure 3.1-2)

(Figure 8-7)

(Figure 3.1-3)

(Figure 8-8)

I (Figure 3.1-3a)

(Figure 8-9)

(Figure 3.1-4)

(Figure 8-10) 3.1.3.9 8-24 Revised figure numbers.

8-25 thru 8-27 Revised APSR limits for cycle (Figure 3.1-9)

(Figure 8-11) 4 operacion.

I (Figure 3.1-9a)

(Figure 8-12)

(Figure 3.1-10)

(Figure 8-13) 3.2.1 8-28 Revised figure number.

8-29 thru G-31 Revised axial power imbalance (Figure 3.2-1)

(Figure 8-14) envelope for cycle 4 opera-I (Figure 3.2-la)

(Figure 8-15) t ion.

(Figure 3.2-2)

(Figure 8-16)

I (Table 3.3-1) 8-32 Added nuclear overpower (Table 3.3-1)

(Table 8-3) based on RCPPMs trip.

I 8-33 Added nuclear overpower (Table 3.3-2)

(Table 8-4) based on RCPPMs trip.

8-34 Added nuclear overpower I

(Table 4.3-1)

(Table 8-5) based on RCPPMs trip.

Bases 3/4.1.2 8-35 Revised shutdown margin and I

8-36 boron requirements to ac-count for inadvertent dilu-tion by sodiur. hydroxide.

I I

lI Babcock & Wilcox 8-3

I Figure 8-1.

Reactor Core Satety Limits (Tech Spec Figure 2.1-1) 1 2400 Il RCS PRESSURE-HIGH TRIP RC OUTLET TEMP __

g 2200 HIGH TRIP g

ACCEPTABLE y

OPERATION

=

m

=

2000 4

SAFETY LIMIT e

/

4 UNACCEPTABLE OPERATION RCS PRESSURE 1800 i

LOW TRIP 3

580 600 620 640 Reactor Outlet Temperature, F I

I I

I 8-4 Babcock & Wilcox

lI Figure 8-2.

Reactor Core Safety Limits (Tech Spec Figure 2.1-2)

- 120

(-26.88,112)

(20.16.112) m m

EPTABLE 4 PUMP

(-39,105) #

OPERAT!0N l

- - 100 (39,100)

(20.I6,89.71)

(-26.88,89.71)

_ _ 90 n

m

(-39,82.71) (

ACCEPTABLE 3 & 4

- - 80 g.

I PUMP OPERATION

- - 70

~

- - 60 E'

U- - 50 i;g

,c

]E

- - 40 e

jI

- - 30 e

t j

$- - 20

5-10 E

1 f

f 1

f f

I l 50

-40

-30

-20 10 0

10 20 30 40 50 60 Reactor Power imoalance, %

I i

lI

!I

!I 8-5 Babcock & Wilcox

]

1

_ Table 8-2.

RPS Trip Setpoints Table 2.2-1.

Reactor Protection System Instrumentation Trip Setpoints Functional unit Trip setpoint Allowable values

1. Manual reactor trip Not applicable Not applicable

2. Nuclear overpower 5 104.88% of RATED THERMAL POWER 5 104.88% ci RATED 'HERMAL POWER with four pumps operating with four pump". creating 5 79.92% of RATED THERMAL POWER 5 79.92% of RATED THERMAL POWER with three pumps operating with three pumps operating

3. RCS outlet temp-high 5 618 F 5 618 F

4. Nuclear overpower Trip setpoint not to excr o the Allowattle values not to exceed based on RCS flow and limit line of Figure 2.< -1 the limit line of Figure 2.2-1 AXIAL POWER IMBALANCEa a

5. RCS pressure-low 2 1800 psig 2 1800 psig 7

6. RCS pressure-high 5 2300 psig 5 2300 psig

7. RCS pressure-variable-2 (11.59 Tout F-5037.8) psig i: (11.59 T lowa out F-5037.8) psig

8. Nuclear overpower More than one pump inoperable.

More than one pump inoperable.

based on RCPPMsa

9. Reactor containment 5 4 psig s 4 psig vessel 8

Trip may be manually bypassed when RCS pressure s 1720 psig by actuating the shutdown bynass, provided that (1) the nuclear overpower trip setpoint is _ 5% of RATED THERMAL POWER, (2) the m

g.

shutdown bypass RCS pressure-high trip setpoint of s 1720 psig is imposed, and (3) the shut-g down bypass is removed when RCS pressure > 1800 psig.

9-e.

a t

g

Figure 8-3.

Reactor Trip Setpoints (Tech

(

Spec Figure 2.2-1)

- 120 lI

(-7.107)

- - 110 (+,107 )

2 m

m M i

+ 0 6800

- - 100

=

M2 = -0.7391 l

ACCEPTABLE 4

(-32.90)

PUMP OPERATl0ft - 90

(+25.90)

(-7.79.G2) P O 80 U2,79.92)

-- 70

(-32,62.92) 4 a

) (+25,62. 92)

ACCEPTABLE 5- - 60 I

3 & 4 PUMP OPERAT10t; 3 -- - 50 N

E- - 40 I

-- 30 m

0 5

1

- - 20 a

tj - - 10 i

t t

i I

t t

f f

f

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 Reactor Power imoalance, %

8-7 Babcock & Wilcox

I SAFETY LIMITS BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core out-let pressure, providing a more c.onservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two ther-mal limits and account for the effects of potential fuel densification and E

potential fuel rod bow:

W l.

The 1.30 DNBR limit produced by a nuclear power peaking factor of F" = 2.57 or the combination of the radial peak, axial peak ar.dhositionoftheaxialpeakthatyieldsnolessthana1.30 DNBR.

l

?

The combination of radial and axial peak that causes central fuel melting at the hot spot.

The limit is 19.7 kW/ft.

Power peaking is not a directly oi: servable quantioy and therefore limits iuve been established on the basis

'f tne reactor power imbalance produced by 4

the sower peakir.g.

l The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond j

to the expected minimum flow rates with four pumps and three pumps respective-I lv.

Il ~

ll The curve of Figure 2.1-1 is the most restrictive if all possible reactor coolant pump-maximum thermal power t.ombinations shown in BASES Figure 2.1.

The curves of 3ASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the num-l' ber of reactor coolant pumps in operation.

These curves include the potential effects of fuel rod 50; and +uel densification.

i I

The DNBR as calculated by the BAW-2 DNB correlation continually increases l

from point of minimum DNBR, so that the exit DNBR is always higher.

Extrapo-lation of the correlation beyond its published quality range of 22% is justi-fied on the basis of experimental data.

b each curve of BASES Figure 2.1, a pressure-temperature point above and - o th 14t of the curve would result in a DNBR greater than 1.30 or a loca: 4'it at the point of minimum DNBR less than 22% for that particular reactor coolant pump situation.

The 1.30 DNBR curve for three pump operation is more restri;tive than any other reactor coolant pump combination because any pressure / temperature point above and to the left of the three pump curve will be above and to the left of the other curves.

I Babcock & Wilcox 8-8 l

I 2.2 LIMITING SAFETY SYSTEM SETTINGS I

)

BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor Trips are set for each param-eter.

The Trio Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

Operation with a trip setpoint less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

The Shutdown Bypass provides for bypassing certain functions of the I

Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures.

The' purpose of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated.

This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass :s initiated. The Nuclear Overpower Trip Setpoint of s 5.0%

prevents any significant reactor power from being produced.

Sufficient natu-I ral circulation would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating.

Manual Reactor Trip The Manual Reactor Trip is a red e dant channel to the automatic Reactor Protection System instrumentation channels and provides manual reactor trip capability.

Nuclear Overpower I

A Nuclear Overpower trip at high power level (neutron flux) provides re-actor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure prctective circuitry.

During normal station operation, reactor trip is initiated when the re-l+

actor power level reaches 104.88% of rated power. Due to calibration and in-h strument errors, the maximum actual power at which a trip would be actuated could be 112%, which was used in the safety analysis.

8-9 Babcock 8 Wilcox

I l

L M TING SAFETY SYSTEM SETTINGS BASES

=

I P actor Containment Vessel Pressure - High 3

The Reactor Containment Vessel Pressure - High Trip Setpoint s 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a ' mss-of-cool?.nt accident, even in the absence of a RCS Pressure - Low trip.

Nuclear Overpower Based on RC?PMs In connection with the Flu.y-Delta Flux-Flow Trip the Nuclear Overpower Based on RCPPM Trip prevents the minimum core DNBR from decreasing below 1.30 by tripping the reactor due to the loss of reactor coolant pumps. The pump monitors also restrict the power level for the number of pumps in operation.

I I

I

' ' CRYSTAL RIVER - UNIT 3 Babcock & Wilcox 8-10

I I

Figure 8-4.

Pressure / Temperature I. in ir s (Tech Spec ILises rigure 2.1) 2400

.I CURVE 2 3 PUMP N

J 2200 E

E S

I

[

CURVE 1 y

4 PU. '

g 2000 5

o I

I 1800 -

I f

580 600 620 640 Reactor Outlet Temperature, F I

l CURVE FLOW

(".

DESIGN)

POWER (RTP)

PUMPS OPERATING (TYPE OF LIMIT) 1 106.5 113.055 4 PUMPS (DNBR) 2 79.6 90.84%

3 PUMPS (DNBR)

I I

I 8-11 Babcock & Wilcox

3/4.1 REACTIVITY CONTROL SYSTEMS Ii 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be 2 3.5% Ak/k.

APPLICABILITY: MODES 4 and 5.

ACTION:

With the SHUTDOWN MARGIN < 3.5% Ak/k, immediately initiate and continue bora-tion at 2 10 gpm of 11,600 ppm boric acid solution or its equivalent, until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be 2 3.5% Ak/k:

l a.

Within one hour after detection of an inoperable control rod (s)

E and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is E

inoperable.

If the inoperable control rod is immovable or un-trippable, the above-required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immov-able or untrippable control rod (s).

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1.

Reactor coolant system baron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and l

6.

Samarium concentration.

l l

I CRYSTAL RIVER - UNIT 3 I

Babcock &Wilcox 8-12

L E

REACTIVITY CONTROL SYSTEtis l

l FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE:

l a.

A flow path from the concentrated boric acid storage system via a I

boric acid pump and makeup or decay heat removal (DHR) pump to the Reactor Coolant System, and I

b.

A flow path from the borated water storage tank via makeup or DHR pump to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

MODES 1, 2, and 3:

a.

With the flow path from the concentrated boric acid storage system inoperable, restore the inoperable flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a I

SHUTDOWN MARGIN equivalent to 1% ak/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , I

b.

With the flow path from the borated water storage tank inoperable, restore the flow path to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN I

within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODE 4:

a.

With the flow path from the concentrated boric acid storage system inoperable, restore che inoperable flow path to OPERABLE status e

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be borated to a SHUTDOWN MARGIN equivalent to I

3.5% t.k/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 CRYSTAL RIVER - UNIT 3 8-13 Babcock & Wilcox

REACTIVITY CONTROL SYSTEMS MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4.2.

At least one makeup pump shall be OPERABLE.

APPLICABILITY: MODE 4*

ACTION:

With no makeup pump OPERABLE, restore at least one makeup pump to OPERABLE status within one hour or be borated to a SHUTDOWN MARGIN equivalent to 3.5% ak/k at 200 F and be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l SURVEILLANCE REQUIREMENTS 4.1.2.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

With RCS pressure 2 150 psig.

I I

I CRYSTAL RIVER - UNIT 3 8-14 Babcock & Wilcox

l l

REACTIVITY CONTROL SYSTEMS 1

BORIC ACID PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.7 At least one boric acid pump in the boron injection flow path re-quired by Specification 3.1.2.2a shall be OPERABLE and capable of being pow-ered from an OPERABLE emergency bus if the flow path through the boric acid pump in Specification 3.1.2.2a is OPEPABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

MODES 1, 2, and 3:

With no boric acid pump OPERABLE. restore at least one boric acid pump to OPERAPLE status v!ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% ak/k at 200F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; re-store at least one boric acid pump to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODE 4:

I With no boric acid pump OPEPABLE, restore at least one boric acid pump to lOPERABLEstatuswithin72hoursorbeboratedtoaSHUTDOWNMARGINequivalen l

to 3.5% ek/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least one boric acid pump to OPERABLE status within the next 7 days or be in COLD StiJTDOWN I

l viithin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.7 No additional Surveillance Requirements other than those required I

by Specificaticn 4.0.5.

I I

CRYSTAL RIVER - UNIT 3 Babcock & Wilcox 8-15

I J REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCES -- OPERATING

3. '. 2. 9 Each of the following borated water sources shall be OPERABLE:

a. The concentrated boric acid storage system and associated heat tracing with:

1.

A minimum contained borated water volume of 6730 gallons, 2.

Between11,600 and 14,000 ppm of boron, and 3.

A minimum solution temperature of 105F.

b. The borated water storage tank (BWST) with:

1. A contained borated water volume of between 415,200 and 449,000 gallons,

2. Between 2270 and 2450 ppm of boron, and

3. A minimum solution temperature of 40F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

MODES 1, 2, and 3:

a.

With the concentrated bcric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at 3

least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 5

1% ak/k at 200F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the concentrated boric acid ' storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODE 4:

a.

With the concentrated boric acid storage system inoperable, restore the 5torage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be borated to a SHUTDOWN MARGIN equivalent to 3.5% ak/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;l restore the concentrated boric acid storage system to OPERABLE status g

within the next 7 days or be in cold shutdown within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

E b.

With borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.9 Each borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the boren concentration in each water source, l

2.

Verifying the contained borated water volume of each water source, and 3.

Verifying the concentrated boric acid storage system solution temperature.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the BWST temperature when the outside air temperature is < 40F.

CRYSTAL RIVER - UNIT 3 8-16 Babcock & Wilcox l

1

L

[

REACTIVITY CONTROL SYSTEMS E

REGULATING R0D INSERTION LIMITS

{

LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-1, 3.1-la, 3.1-2, 3.1-3, 3.1-3a, and 3.1-4 with a rod l

[

group overlap of 25 5% between sequential withdrawn groups 5 and 6, and 6 and 7.

I APPLICABILITY: MODES 1* and 2*#.

ACTION:

With the regulating rod groups inserted beyond the above insertion limits, or with any group sequence or overlap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

I a.

Restore the regulating groups to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or g

b.

Reduce THERMAL POWER to less than or equal to that fraction of g

RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I

See Special Test Exceptions 3.10.1 and 3.10.2 With Keff a 1.0.

I I

I l

l ll CRYSTAL RIVER - UNIT 3 8-17 Babcock & Wilcox

Figure 8-5.

Pegulating Rod Group Insertion Limits for Four-Pump Operation From 0 to 50 (+10/-0) EFPD (Tech Spec Bases Figure 3.1-1) 110 100

( 8, Op g ] 2)

POWER LEVEL 90 (180,92)

(209,92) 00 80 UNACCEPTABLE (171,80)

(213,80) g OPERATION E

p 70 OJ E

60 I

50 (300,50)

]

(102,50) 3 40 d'

30 20 -

ACCEPTABLE OPERATION 10 -

(0,6) 0 0

50 100 150 200 250 300 Rod Index, ! Witnarawn 0

25 50 75 100 0

25 50 75 100 i

f I

I I

I Group 5 Group 7 E

O 25 50 75 100 E

1 1

I I

I OfDup 6 I

I Dabcock & Wilcox 8-18

I Figure 8-6.

Regulating Rod Group Insertion Limits for Four-Pump Operation From 50 (+10/-0) to 270 10 EFPD (Tech Spec Bases Figure 3.1-la) l 109 (182,102),

, (209,102)

POWER LEVEL W 2, m (C 90 (180,02) 100: FP) l 80 (171.80)

(219.80) y UNACCEPTABLE l

f 70 OPERATION 2

m 60 B

a 33 (102.50)

(300.50)

I e

g 40 I

3 m

30 -

I 20 -

(35,15) 0 - (0,1.9) 0 50 100 150 200 250 300 Rod Index, % Witnarawn 0

25 50 75 100 0

25 50 75 100 1

I I

I i

i l

i Group 5 0

25 50 75 100 Group 7 I

I I

GIOUp 6 I

8-19 Babcock & Wilcox w -

I i

Figure 8-7.

Regulating Rod Group Insertion Limits for Four-Pump Operation Af ter 270 : 10 EFPD (Tech Spec Bases Figure 3.1-2) 110 100 POWER LEVEL (300,10?)

CUT 0FF 90 (261,92) g 5-f 80

(

I UNACCEPTABLE OPERATION E

7G E

60 E

50 (156,50)

+,

I j

40 30 20 (35.15)

ACCEPTABLE OPERATION 0

0 50 100 150 200 250 300 Rod Index, t Witnarann E

O 25 50 75 100 0

25 50 75 100 l

i 1

I I

I f

I Group 5 Group 7 I

I I

f f

Group 6 I

8-20 Babcock & Wilcox

L r

L Figure 8-8.

Regulating Rod Group Insertion Limits for Three-Pump

[

Operation From 0 to 50 (+10/-0) EFPD (Tech Spec H

Bases Figure 3.1-3) 110 100 90 L

y 80 (164,77)

(222.77)

E

)

70 UNACCEPTABLE

{

OPERATION

~

[

60 l

5 50 (102,50)

(300,50) o l

40 30 20 ACCEPTABLE CPERATION 10 (0,6) 0 i

f I

I i

0 50 100 150 200 250 300 Rod Index, % Witndrawn 0

25 50 75 100 0

25 50 75 100 l

I l

l 1

1 1

l l

l Group 5 0

25 50 75 100 Group 7 1

1 i

i I

l Group 6 8-21 Babcock & Wilcox I

)

I Figure 8-9.

Regulating Rod Group Insertion Limits for Three Pump Operation From 50 (+10/-0) to 270 2 10 EFPD (Tech Spec Bases Figure 3.1-3a) 110 100 90 UNACCEPTABLE OPERATION 80 (227.77) y (165,77) 5 70 E

60 I

[

50 (300.50) i i

g 40 (90.38) a 30 -

ACCEPTABLE 20 -

OPERATION 10 -

(35,11.8) 0 I

0 50 100 150 200 250 300 Rod incax, $ Witncrawn 0

25 50 75 100 0

25 50 75 100 i

t f

f i

1 I

f f

I l

Group 5 Group 7 0

25 50 75 100 1

i f

f I

Group 6 l

I I

Babcock & Wilcox 8-22

I Figure 8-10.

Regulating Rod Group Insertion Limits for Three-Pump Operation Af ter 270 t 10 ETPD (Tech Spec Bases Figure 3.1-4) 110 100 90 80 UNACCEPTABLE OPERATION (241,77) j (300,77) g 70 I

C SD 5

50 (156.50) d a

40

.g 3

30

~

20 -

ACCEPTABLE OPERATICN 10 -

'I'O' 0,1.9) 0 0

50 100 150 200 250 300 i

Roa index, 5 Witnoraan 0

25 50 75 100 0

25 50 75 100 I

t I

I I

Group 5 Group 7 0

25 50 75 100 I

i i

l I

l Group 6 t

1I lI i

!I l

l 8-23 Babcock & Wilcox l

l

I REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION' 3.1.3.9 The axial power shaping rod group shall be limited in physical in-sertion as shown on Figures 3.1-9, 3.1-9a, and 3.1-10.

I l

APPLICABILITY: MODES 1 and 2*.

ACTION:

With the axial power shaping rod group outside the above insertion limits, either:

a.

Restore the axial power shaping rod group to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allcwed by the rod group position using the above figure within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I SURVEILLANCE REQUIREMENTS I

4.1.3.9 The position of the axial power shaping rod group shall be determined to be within the insertion limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K 2 1.0.

eff I

I l

I I

I CRYSTAL RIVER - UNIT 3 8-24 Babcock & Wilcox

I Figure 8-11.

APSR Position Limits for 0 to 50 (+10/-0) EFPD (Tech Spec Bases Figure 3.1-9) i 110 (8,102)

(32.102) 90 (8,92)

(38,92) g 80 (0.80)

(38,80)

UNACCEPTABLE I

j OPERATION E

70 I

E E

60

~

50 (100,50)

I o

40 -

m E

I O

30 I

20 ACCEPTABLE OPERATION g

0 10 20 30 40 50 60 70 80 90 100 Rod Po sition, % Witndrawn I

I I

I 8-25 Babcock & Wilcox l

i I

Figure 8-12.

APSR Position Limits for 50 (+10/-0) to 270 210 EFPD (Tech Spec Bases Figure 3.1-9a) l 110 (8.102)

(32,102) 100 g

90 (8,92)

(41,92)

E l

UNACCEPTABLE (51.80) 80 (0,80) p I

5 70 I

80 e

3 g

l 5

E 50 (100,50) g l

n 40 m

E f

E 30 -

E 20 -

ACCEPTABLE OPERATif;il 10 -

l 0

I i

i i

0 10 20 30 40 50 60 70 80 30 100 Rod Position, % Witnarawn I

I I

I 8-26 Babcock & Wilcox

I I

Figure 8-13.

APSR Position Limits Af ter 270 ! 10 EFPD (Tech Spec Bases Figure 3.1-10) 110 100 (35 102) 90 (8.92)

(47 92)

UNACCEPTA8l.E I

OPERATION 80 (0.80)

(55,80)

I 70 5

o 5

60 I

a E

50 (100,50)

E l

'[

40 --

h 30 I

j 20 -

ACCEPTABLE OPERATION 10 -

l 0

10 20 30 40 50 60 70 80 90 100 E

Rod Position, f, Witndrawn l

ll ll l

8-27 Babcock & Wilcox

I 3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1, 3.2-la, and 3.2-2.

l APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.*

ACTION:

With AXIAL POWER IMBALANCE exceeding the limits specified above, either:

a.

Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or b.

Be in at least HOT STANDBY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REQUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits in each core quadrant at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above 40% of RATED THERMAL POWER except when an AXIAL POWER IMBALANCE monitor is inoperable, then calculate the AXIAL POWER IMBALANCE in each core quadrant with an inoperable monitor at least once por hour.

See Special Test Exception 3.10.1.

I I

l CRYSTAL RIVER - UNIT 3 g

Babcock & Wilcox 8-28

I l

Figure 8-14.

Axial Power Imbalance Fnvelope for Operarion Frcm 0 to 50 (+10/-0) EFPD (Tech Spec Bases Figure 2-1)

- - 110

( -14.4,102)

(+10.4,102)

( -18. 5,92 )

- - 90 y- - 80

(+17,80)

( -19,80 )

I f

g- - 70

=

I

- - 60 E

a:

I

-- 50 E

o 40 I

O

-- 30 UNACCEPTABLE OPERATION

-- 20 I

-- 10 ACCEPTABLE OPERATION I

I f

1 1

I I

I i 40

-30

-20

-10 0

10 20 30 40 50 Axial Power Imualance, %

I 8-29 Babcock & Wilcox

I I

Figure 8-15.

Axial Power Imbalance Envelope for Operation From 50 (+10/-0) to 270 1 10 EFPD (Tech Spec Bases Figure 3.2-la)

- - 110 l

( -15.3,102)

(+15.3,102)

)

(+25,92)

-- 90 l

( - 3,80) 5-- 80

( +30. 80 )

E

- - 70 5

i I

o

-- 60

-- 50 I

i

-- 40 5

a.

- - 30

- 20 UNACCEPTABLE ACCEPTABLE OPERATION E

-- 10 CPERATION g

-50 30

-20 10 0

10 20 30 40 50 Axial Power imoalance, 5 I

I I

8-30 Babcock 8.Wilcox I

[

Figure 8-16.

Axial Power Imbalance Envelope for Operation After 270 ! 10 EFPD (Tech Spec Bases Figure 3.2-2)

- - 110

(-19,102)

[

-- 100

(-31.92)

-- 90

[

(-32,83)

-- 80

( +30,80 )

5 a.

g

-- 70 b

E

-- 60

-- 50 E

[

a

-- 40

[

-- 30 UNACCEPTABLE CPERATION

[

-- 20 ACCEPTABLE

[

-- 10 OPERATION

{

i i

e i

i i

i I

i

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Power Imualance, f6

[

Babcock & Wilcox 8-31

Table 8-3.

Reactor Protection Systen Instrumentation (Tech Spec Table 3.3-1)

Minimum Total No.

Channels channels Applicable Functional Unit of channels to trip operable modes Act ion 1.

Manual Reactor Trip 1

1 1

1, 2, and

8 2.

Nuclear Overpower 4

2 3

1, 2 2#

3.

RCS Outlet Temperature - High 4

2 3

1, 2 3G 4.

Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE 4

2(a) 3 1, 2 2#

5.

RCS Pressure - Low 4

2(a) 3 1, 2 3#

6.

RCS Pressure - High 4

2 3

1, 2 3#

oo 7.

Variable Lou RC S Precsure 4

2 (a) 3 1, 2 3#

8.

Nuclear Overpower Based on RCPPMs 4

2 (a) (b) 3 1, 2 3#

9.

Reactor Containment Pressure - High 4

2 3

1, 2 3#

10.

Intermediate Range, Neutron Flux and 2

0 2

1, 2 and

4 11.

Source Range, Neutron Flux and Rate A.

Startup 2

0 2

2## and

5 B.

Shutdown 2

0 1

3, 4 and 5 6

12.

Control Rod Drive Trip Breakers 2/ trip 1/ trip 2/ trip 1, 2 and

7#

system system system fD 13.

Reactor Trip Module 2/ trip 1/ trip 2/ trip 1, 2 and

7#

c[

system system system N

14.

Shutdown Bypass RCS Pressure - High 4

2 3

2**,3**

6#

l[

4**,

5**

o-aus aus as sus sua sum um nun aus muu em uma sus em aus aus mas sus num

Tabt.c 8-4.

Reactor Protection System Instrumentation Response Times (Tech Spec Table 3.3-2)

Functional Unit Response Times 1.

Manual Reactor Trip Not Applicable 2.

Nuclear Overpower

5 0.3 seconds 3.

RCS Outlet Temperature - High Not Applicable 4.

Nuclear Overpower Based on RCS Flcw and AXIAL POWER IMBALANCE

5 1.4 seconds 5.

RCS Pressure - Low 5 0.5 seconds 6.

RCS Pressure - High s 0.5 seconds

?

O 7.

Variable Low RCS Pressure Not Applicable 8.

Nuclear Overpower Based on RCPPMs*

5 0.47 9.

Reactor Containment Pressure - High Not Applicable

Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

?

cr O

R-P hw

Table 8-5.

Reactor Protection System Instrumentation Surveillance Requirements (Tech Spec Table 4.3-1)

Channel Modes in which Channel Channel functional surveillance Functional Unit check calibrat ion test required 1.

Manual Reactor Trip N.A.

N.A.

S/U(1)

N.A.

2.

Nuclear Overpower S

D(2) and Q(7)

M 1, 2 3.

RCS Outlet Temperature - High S

R M

1, 2 4.

Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE S(4)

M(3) and Q(7,8)

M 1, 2 5.

RCS Pressure - Low S

R M

1, 2 6.

RCS Pressure - High S

R M

1, 2 m

7.

Variable Low RCS Pressure S

R M

1, 2 5

8.

Nuclear Overpower Based on RCPPMs S

R M

1, 2 9.

Reactor Containment Pressure - High S

R M

1, 2 10.

Intermediate Range, Neutron Flux and Rate S

R(7)

S/U(1)(5) 1, 2 and

11.

Source Range, Neutron Flux and Rate S

R(7)

S/U(1)(5) 2, 3, 4 and 5 12.

Control Rod Drive Trip Breaker N.A.

N.A.

M and S/U(1) 1, 2, and

13.

Reactor Trip Module N.A.

N.A.

M 1, 2, and

14.

Shutdown Bypass RCS Pressure - High S

R M

2**,

3**,

4**,

5**

??

8x to av 1

M M

M

I REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the reactor coolant system average temperature less than 525F, This limita-tion is required to ensure that (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range, (3) the pressurizer is capable of being in I

an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4.1.2 B0 RATION SYSTEMS The boron injection system ensures that negative reactivity control is avail-able during each mode of facility operation.

The components required to per-I form this function include (1) borated water sources, (2) makeup or DHR pumps, (3) separate flow paths, (4) boric acid pumps, (5) associated heat tracing systems, and (6) an emergency power supply from OPERABLE emergency busses.

With the RCS average temperature above 200F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoper-able. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 3.5% ak/k after xenon decay and cool-I down to 200F.

The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires either 6730 gallons of 11,600 ppm boric acid solution from the boric acid storage tanks or 47,698 gallons of 2270 ppm borated water from the borated water storage tank.

The requirements for a minimum contained volume of 415,200 gallons of borated water in the borated water storage tank ensures the capability for borating u

the RCS to the desired level.

The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.

Therefore, the r

larger volume of borated water is specified.

With the RCS temperature below 200F, one injection system is acceptable with-out single failure consideration on the basis of the stable reactivity condi-L tion of the reactor and the additional restrictions prohibiting CORE ALTERA-TIONS and positive reactivity chance in the event the single injection system becomes inoperable.

I The boron capability required below 200F is sufficient to provide a SHUTDOWN MARGIN of 3.5% ak/k after xenon decay and cooldown from 200F to 140F. This condition requires either 300 gallons of 11,600 ppm boric acid solution from the boric acid storage system or 1608 gallons of 2270 ppm borated water from the borated water storage tank. To envelop future cycle BWST contained bor-ated water volume requirements, a minimum volume of 13,500 gallons is speci-L fied.

CRYSTAL RIVER - UNIT 3 Babcock & Wilcox 8-35

I The contained water volume limits include allowance for water not available g

because of discharge line location and other physical characteristics. The 5

limits on contained water volume and boron concentration ensure a pH value of between 7.2 and 11.0 of the solution sprayed within the containment after a design basis accident. The pH band minimizes the evolution of iodine and min-imizes the effect of chlcrides and caustic stress corrosion cracking on me-chanical systems and components.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section (1) ensure that acceptable power distribu-tion limits are maintained, (2) ensure that the minimum SHUTDOWN MARGIN is E

maintained, and (3) limit the potential effects of a rod ejection accident.

5 OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The ACTION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensure that the origi-nal criteria are met.

For example, misalignment of a safety or regulating rod requires a restriction in THERMAL POWER. The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent ex-g ceeding the assumptions used in the safety analysis.

3 The position of a rod declared inoperable due to misalignment should not be g

included in computing the average group position for determining the g

OPERABILITY of rods with lesser misalignments.

I I

I CRYSTAL RIVER - UNIT 3 8-36 Babcock & Wilcox

i t

I

!I i

]

9.

STARTUP PROGRAM - PHYSICS TESTING 3

The planned startup test program associated with core performance is outlined j

below. These tests verif y that core performance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of 1,

the unit.

9.1.

Precritical Tests 4

9.1.1.

Control Rod Trip Test il j

Precritical control rod drop times are recorded for all control rods at hot 1

full-flow conditions before zero power physics testing begins. Acceptance I5 criteria state that the rod drop time from fully withdrawn to 75% inserted l

shall be less than 1.66 seconds at the conditions above.

It should be noted that safety analysis calculations are based on a rod drop i

time of 1.40 seconds from fully withdrawn to two-thirds inserted.

Since the most accurate position indication is obtained from the zone reference switch i

at the 75%-inserted position, this position is used instead of the two-thirds j

inserted positien for data gathering. The acceptance criterion of 1.40 seconds j

corrected to a 75%-inserted position (by rod insertion versus time correlation) is 1.66 seconds.

f 9.1.2.

RC Flow RC Flow with four RC pumps running will be measured at hot zero power, steady-

)

state conditions. Acceptance criteria require that the measured flow be with-in allowable limits.

!I I

!I Babcock & Wilcox 9-1

I 9.2.

Zero Power Physics, Tests 9.2.1.

Critical Boron Concentration Criticality is obtained by deboration at a constant dilution rate. Once criti-cality is achieved, equilibrium boron is obtained and the critical boron con-centration determined. The critical boron concentration is calculated by cor-recting for any rod withdrawal required in achieving equilibrium boron. The acceptance criterion placed on critical boron concentration is that the actual g

boron concentration mt st be within 100 ppm boton of the predicted value.

9.2.2.

Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at a' proximately the all-g rods-out configuration and at the hot zero power rod insertion limit. The 5

average coolant tenperature is varied by first decreasing then increasing tem-perature by 5 F.

During the change in temperature, reactivity feedback is cc,m-pensated by discrete change in rod motion, the change in reactivity is then calculated by the summation of reactivity (obtained from reactivity calculation on a strip chart recorder) associated with the temperature change. Acceptance criteria state that the measured value shall not differ from the predicted value by more than !0.4 x 10-4 (Ak/k)/ F (predicted value obtained from Physics Test Manual curves).

The moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has i

been measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain moderator coef ficient. This value must not be in excess of the acceptance criteria limit of +0.9 x 10-" (Ak/k)/*F.

9.2.3.

Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6, and 7) are measured at hot zero power conditions using the boron / rod swap method. The boron / rod swap g

method consists of establishing a deboration rate in the reactor coolant sys-E tem and compensatiag for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 in inctemental steps. The reactivity changes that occur during these measurements are calculated based on Reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each of the I

Babcock & Wilcox 9-2

I I

controlling groups are then summed to obtain integral rod group worths. The acceptance criteria for the control bank group worths are as follows:

1.

Individual bank 5, 6, 7 worth:

predicted value - measured value 100 1 15 measured value 2.

Sum of groups 5, 6, and 7:

predicted valde - measured value x 100 s 10 measured value 9.2.4.

Ejected Control Rod Reactivity Worth Af ter CRA groups 7, 6, and 5 have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth ob-tained by adding the incremental changes in reactivity by boration.

Af ter the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is then swapped in versus the controlling rod group and the worth determined by the change in the previcusly calibrated controlling rod group position.

The boron swap and rod swap values are aver-I aged and error-adjusted to determine ejected rod worth. Acceptance criteria for the ejected rod worth test are as follows:

I predicted value - measured value x 100 s 20 measured value 2.

Measured value (error-adjusted) c 1.0% t.k/k The predicted ejected rod worth is given in the Physics Test Manual.

9.3.

Power Escalation Tests I

I 9.3.1.

Core Power Distribution Verification at 440, 75, and 100% FP With Nominal Control Rod Pasition Core power distribution tests are performed at 40, 75, and 100% full power (FP). The test at 40% FP is essentially a check on power distribution in the core to identify any abnormalities before escalating to the 75% FP platecu.

Rod index is established at a nominal full power rod configuration at which the core power distribution was calculated. APSR position is established to provide a core power imbalance corresponding to the imbalance at which the core power distribution calculations were performed.

9-3 Babcock & Wilcox

I The following acceptance criteria are placed on the 40% FP test:

1.

The worst-case maximum linear heat rate must be less than the LOCA limit.

2.

The minimum DNBR must be greater than 1.30.

3.

The value obtained from the extrapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30 or the extrapolated value of imbalance must fall outside the RPS power / imbalance /

flow trip envelope.

4.

The value obtained from the extrapolation of the worst-case maximum linear heat rate to the next power plateau overpower trip setpoint must be less than the fuel melt limit or the extrapolated valve of imbalance must fall outside the RPS power / imbalance / flow trip envelope.

5.

The quadrant power tilt shall not exceed the limits specified in the Tech-nical Specifications.

6.

The highest measured and predicted radial peaks shall be within the follow-ing limits:

predicted value - measured value x 100 s8 measured value 7.

The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value x 100 measured value c 12 Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and ther-g mal calculational models, thereby verifying the acceptability of data from W

these models for input to safety evaluations.

Items 3 and 4 establish the criteria whereby escalation to the next power pla-teau may be accomplished without exceeding the safety limits specified by the g

safety analysis with regard to DNBR and linear heat rate.

5 The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test except that core equilibrium xenon is established prior to the 75

=

and 100% FP tests.

Accordingly, the 75 and 100% FP measured peak acceptance criteria are as follows:

I 9-4 Babcock s. Wilcox

[

1.

The highest measured and. predicted radial peaks shall be within the follow-ing limits:

predicted value - measured value

[

x 100 s5 measured value 2.

The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value

[

100

$ 7.5 measured value 9.3.2.

Incore Vs Excore Detector Imbalance

{

Correlation Verification at N40% FP Imbalances are set up in the core by control rod positioning.

Imbalances are

{

read simultaneously on the incore detectors and axcore power range detectors for various imbalances. The excore detector offset versus incore detector off-set slope must be at least 1.15.

If the excore detector offset versus incore detector offset slope criterion is not met, gain amplifiers on the excore de-tector signal processing equipment are adjusted to provide the required gain.

[

9.3.3.

Temperature Reactivity Coefficient at 4100% FP

{

The average reactor coolant temperature is decreased and then increased by about 5 F at constant reactor power. The reactivity associated with each tem-perature change is obtained from the change in the controlling rod group posi-

[

tion.

Controlling rod group worth-is measured by the fast insert / withdraw method. The temperature reactivity coefficient is calculated from the mea-sured changes in reactivity and temperature.

Acceptance criteria state that the moderator temperature coefficient shall be negative.

9.3.4.

Power Doppler Reactivity Coef ficient at N100% FP

[.

Reactor power is decreased and then increased by about 5% FP.

The reactivity change is obtained from the change in controlling rod group position. Control rod group worth is measured using the fast insert / withdraw method. Reactivity corrections are made for changes in xenon and reactor coolant temperature that occur during the measurement. The power doppler reactivity coefficient is calculated from the measured reactivity change, adjusted as stated above, and f

the measured power change.

[

9-5 Babcock & Wilcox

I The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual. Acceptance criteria state that the measured value shall be more negative than -0.55 x 10-" (Ak/k)/% FP.

=

9.4.

Procedure for Failure to Meet g

Acceptance Criteria E

Florida Power Corporation reviews the results of all startup tests to ensure that all acceptance criteria are met.

If the review of the test indicates that the results are well within the acceptance criteria, no further evalua-tion is conducted.

If the review indicates that the results are approaching or close to the acceptance criteria limits, further evaluation of that par-ticular test or other supporting tests is performed to look for trends. This evaluation will determine whether additional support data are required to dis-cover any G normal conditions.

If acceptance criteria for any test are not met, an evaluation is performed before the test program is continued. This evaluation is performed by site test pere,onnel with participation by Babcock g

& Wilcox technical personnel as required. Further specific actions depend W

on evaluation results. These actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for anom-alle:', or design personnel perf orming detailed analyses of potential safety p:ch '. bece ise of parameter devi ation. Power is not escalated until evalua-tion shova i

u p? :.

sc.i'et y will not be compromised by such escalation.

I I

I 9-6 Babcock & Wilcox

E o

REFERENCES F

1 Crystal River Unit 3. Final Safety Analysis Report, Docket 50-302, Florida Power Corporation.

2 BAW-1607, Rev. 1. Babcock &

Crystal River Unit 3, Cycle 3 Reload Report, __

Wilcox, Lynchburg, Virginia, April 1980.

3 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Vir-ginia, May 1978.

J. H. Tayler (B&W) to S. A. Varga (NRC), Letter, "BPRA Retainer Reinser-tion," January 14, 1980.

5 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep-I Collapse, BAW-10084A, Rev. 2, Babcock 5 Wilcox, Lynchburg, Virginia, October 1978.

6 Crystal River Unit 3, Fuel Densification Report, BAW-1397, Babcock & Wilcox, Lynchburg, Virginia, August 1973.

7 Babcock & Wilcox Version of PDQ User's Manual, BAW-10117P-A, Babcock &

Wilcox, Lynchburg, Virginia, January 1977.

8 L. S. Dubenstein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of In-terin Procedure for Calculating DNBR Reductions Due to Rod Bow," October 18, 1979.

R. C. Jones, J. R. Biller, and B. M. Dunn, ECCS Analysis of B&W's 177-FA Lovered-Loop NSS, BAW-10103A, Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, July 1977.

10 J. H. Taylor (B&W) to R. L. Baer (NRC), Letter, "LOCA Analysis for B&W's 177-FA Plants With Lowered-Loop Arrangement (Category 1 Plants) Utilizing a Revised System Pressure Distribution," July 8, 1977.

11 J. H. Taylor (B&W) to S. A. Varga (NRC), Letter, "ECCS Small Break Analy-sis," July 18, 1978.

A-1 Babcock & Wilcox

+

12 W. P. Stewart (FPC) to R. W. Reid (NRC). Letter, " Crystal River Unit 3, Docket No. 50-302, Operating License No. DPR-72, ECCS Small Break Analy-sis," January 12, 1979.

l' C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1972.

e l'

J. H. Taylor (B&W) to L. S. Rubenstein (NRC) Letter, September 5,1980.

15 Crystal River Unit 3. Licensing Considerations for Continued Cycle 1 Op-eration Without Burnable Poison Rod Assemblies and Orifice Rod Assemblies, BAW-1490, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, July 1978.

I I

A-2 Babcock 8.Wilcox

-

v t e Letter Sequence Other
TAC:46962, (Approved, Closed)
Initiation Request, Request, Request, Request, Request Acceptance... Administration Meeting Questions Answers Results Approval, Approval, Approval, Approval, Approval Other: ML19320D593, ML19331C271, ML19331C277, ML19340B529, ML19350C772, ML20003C510, ML20009H240, ML20009H245, ML20030C723, ML20030C729, ML20032E918, ML20032E922, ML20032E929, ML20033B435, ML20038C770, ML20038C776, ML20038C783, ML20039A978, ML20039A984, ML20039F150, ML20053D278
MONTHYEAR1981-12-08 [Table View]

ML20032E929

Text

SUMMARY

SUMMARY

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