ML20116F659
| ML20116F659 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/30/1985 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20116F636 | List: |
| References | |
| BAW-1860, NUDOCS 8505010164 | |
| Download: ML20116F659 (78) | |
Text
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BAW-1860 April 1985
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CRYSTAL RIVER UNIT 3 j
-- Cycle 6 Reload Report --
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8505010164 850425 PDR ADOCK 05000302.
P PDR Babcock &Wilcox a McDermott company l
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BAW-1860 April 1985 L
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CRYSTAL RIVER UNIT 3
-- Cycle 6 Reload Report --
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BABC0CK & WILC0X Nuclear Power Division P. O. Box 10935 Lynct.uarg, Virginia 24506-0935 N & M8com a McDermott company
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CONTENTS Page 1.
INTRODUCTION AND
SUMMARY
..................... 1-1 2-1 4
2.
OPERATING HISTORY 3-1 3.
GENERAL DESCRIPTION.......................
4.
FUEL SYST EM DE SI GN........................ 4-1 4-1 4.1.
Fuel Assembly Mechanical Design 4.2.
Fuel Rod and Gray APSR Designs............... 4-1 4-1 4.2.1.
Cladding Creep Collapse 4-2 4.2.2.
Cladding Stress 4-2 4.2.3.
Cladding Strain 4-3 4.3.
The rmal De s i g n.......................
4-3 4.4.
Material Design 4-3 4.5.
Operati ng Experience....................
5.
NUCLEAR DESIGN.......................... 5-1 5-1 5.1.
Physics Characteristics 5-2 5.2.
Changes in Nuclear Design 6.
THERMAL-HYDRAULIC DESIGN..................... 6-1 7-1 7.
ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.
General Safety Analysis 7-1 7.2.
Accident Evaluation 7.3.
Dose Consequences of Accidents............... 7-2 8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS........ 8-1 l
9.
STARTUP PROGRAM -- PHYSICS TESTING................ 9-1 9-1 9.1.
Precritical Tests.....................
9-1 l
9.1.1.
Control Rod Trip Test 9.1.2.
Reactor Coolant Fl ow................ 9-1 l
9.2.
Zero Power Physics Tests.................. 9-1 9.2.1.
Critical Boron Concentration............ 9-1 9-2 9.2.2.
Temperature Reactivi ty Coefficient.
9.2.3.
Control Rod Group Reactivi ty Worth......... 9-2 9.2.4 Ejected Control Rod Reactivi ty Wcrth........ 9-3
- ii -
Babcock & Wilcox a McDermott company
CONTENTS (Cont'd)
Page 9.3.
Power Escal a ti o n Te s ts................... 9-3 9.3.1.
Core Power Distribution Verification at 40, 75, and 100% FP With Nominal Control Rod Position 9-3 9.3.2.
Incore Versus Excore Detector Imbalance Correla-tion Verification at 40% FP............ 9-5 Y
9.3.3.
Temperature Reactivity Coefficient at 100% FP... 9-5 9.3.4.
Power Doppler Reactivity Coefficient at 100% FP.. 9-5 i
9.4.
Procedure for Failure to Meet Acceptance Criteria 9-6
- 10. REF E RE NC E S............................
10-1 List of Tables Table 4-5 4-1.
Fuel Assembly Design Parameters 5-1.
Physics Parameters, Crystal River 3 Cycles 5 and 6....... 5-3 5-2.
Shutdown Margin Calculation for Crystal River 3 Cycle 6 5-5 6-1.
Thermal-Hydraulic Design Conditions 6-2 7-1.
Comparison of Key Parameters for Accident Analysis....... 7-4 7-4 7-2.
LOCA Limits for Crystal River 3 7-3.
Comparison of FSAR and Cycle 6 Accident Doses 7-5 8-2 8-1.
Technical Specification Changes 8-2.
Quadrant Power Til t Limits................... 8-31 List of Figures Figure 3-1.
Fuel Shuffle for Crystal River 3 Cycle 6............
3-3 3-2.
Enrichment and B0C Burnup Distribution for Crystal River 3 Cycle 6 Af ter 490 EFPD Cycle 5.................
3-4 3-3.
Control Rod Locations and Group Designations for Crystal 3-5 River 3 Cycle 6 i
3-4.
LBP Enrichment and Distribution for Crystal River 3 Cycle 6 3-6 4-6 4-1.
Gray Axial Power Shaping Rod......
5-1.
B0C (4 EFPD), Cycle 6 Two-Dimensional Relative power Distribu-l tion -- HFP, Equilibrium Xenon, Bank 8 Inserted 5-6 8-1.
Reactor Core Safety Limits................... 8-3 8-2.
Trip Setpoint for Nuclear Overpower Based on RCS Flow and 8-4 Axial Power Imbalance f
- 1ii.
hock &WIIcom a McDermott company J
Figures (Cont'd)
Figure Page 8-3.-
Pressure / Temperature Limits at Maximum Allowable Power for Minimum DNBR........................ 8-10 8-4.
Regulating Rod Group Insertion Limits for Four-Pump Operation From 0 to 200 110 EFPD...................... 8-14 8-5.
Regulating Rod Group Insertion Limits for Four-Pump Operation From 200 i10 to 400 i10 EFPD.................. 8-15
+
8-6..
' Regulating Rod Group Insertion Limits for Four-Pump Operation After 400 i10 EFPD....................... 8-16 8-7.
Regulating Rod Group Insertion Limits for Three-Pump Operation From 0 to 200 10 EFPD..................... 8-17 8-8.
Regulating Rod Group Insertion Limits for Three-Pump Operation From 200 i10 to 400 10 EFPD.................. 8-18 8-9.
Regulating Rod Group Insertion Limits for Three-Pump Operation After 400 t10 EFPD....................... 8-19 8-10.
Control Rod Locations and Group Designations for Crystal 8-20 River 3 Cycle 6 8-11.
Axial Power Shaping Rod Group Insertion Limits From 0 to 400 !10 EFPD........................... 8-22 8-12.
Axial Power Shaping Rod Group Insertion Limits After 400 110 EFPD............................... 8-23 8-13.
Axial Power Imbalance Envelope for Four-Pump Operation From 8-25 0 to 400 110 EFPD........................
8-14.
Axial Power Imbalance Envelope for Four-Pump Operation After 400 i10 EFPD.......................... 8-26 8-15.
Axial Power Imbalance Envelope for Three-Pump Operation 1
From 0 to 400 i10 EFPD..................... 8-27 8-16.
Axial Power Imbalance Envelope for Three-Pump Operation
- Af ter 400 110 EFPD....................... 8-28 i
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Batscock &WHeos a McDermott company
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1.
INTRODUCTION AND
SUMMARY
v This report justifies the operation of Crystal River Unit 3 (cycle 6) at a rated core power level of 2544 MWt.
Included are the required analyses to support. cycle 6 operation; these analyses employ analytical techniques and design bases established in reports that have received technical approval by the U.S. Nuclear Regulatory Commission (NRC; see references).
The cycle 6 rated themal power is 2544 MWt, which was the ultimate core power level identified in the Crystal River Unit 3 Final Safety Analysis Report (FSAR).I The cycle 6 core has been designed with a cycle lifetime of 425 10 effective full power days (EFPD) and utilizes burnable poison rod assemblies (BPRAs) to aid in reactivi ty control.
Gray axial power shaping rods have also been included in the cycle 6 design.
The Technical Specifications have been reviewed, and the modifications for cycle 6 are justified in this report.
Based on the analyses perfomed, which take into account the postulated ef-fects of fuel densification and the Final Acceptance Criteria for emergency l
core cooling (ECC), it has been concluded that Crystal River 3 cycle 6 can be safely operated at a core power level of 2544 MWt.
i i
t 1-1 Babcock &Wilcox a McDermott company
2.
OPERATING HISTORY v
Cycle 5, the current Crystal River Unit 3 operating cycle, is the reference h
fuel cycle for the nuclear and thermal-hydraulic analyses performed for cycle 6 operation.
Cycle 5 achieved criticality on July 24, 1983, completed power escalation testing on August 3, 1983, and was completed on March 9, 1985 after 484.4 EFPD.
No operating anomalies have occurred during previous cycle operations that would adversely affect fuel performance in cycle 6.
Cycle' 6 is scheduled to start operation in July 1985 at a rated power level of 2544 MWt. The design cycle length is 425 t10 EFPD.
2-1 Babcock &Wilcox a McDermott company
c:
4 3.
GENERAL DESCRIPTION The Crystal River Unit 3 reactor core is' described in detail in chapter 3 of the FSAR for the unit.1 The core consists of 177-fuel assemblies _ (FAs),
each of which is a 15x15 array containing 208 fuel rods,16 control rod guide; tubes, and one incore instrument guide tube.
The FAs in batches 6, 7,.and 8 have an average nominal fuel loading of 463.6 kg of uranium, whereas _ the batch 4 assemblies maintain an average nominal fuel loading of
. 468.6-kg 'of uranium.-
The cladding is col d-worked Zircaloy-4 with an outside diameter (00) of 0.430 inch and a wall thickness of 0.0265 inch.
The fuel consists of dished-end, cylindrical pellets of uranium ' dioxide (see Table 4-1 for data).
Figure 3-1_is the core loading diagram for cycle 6 of Crystal River 3.
The-initial enrichments of batches 4A, 4E, 6A, 6B, 7A, and 7B were 2.64, 2.64, 235, _ respectively.
The design enrichment
~
2.62, 2.95, 3.29,. and 2.95 wt %
U
' of fresh batch 8 is 3.49 wt % 235 Four batch 4A assemblies that were U
discharged at the end of cycle ~ 4 were re-inserted into the core interior and one batch 4E assembly that was discharged at the end of cycle 3 was re-inserted as the center assembly.
Thirty-two batch 6 assemblies, 32 batch 5 assemblies, and one batch 4 assem-bly will be discharged at the end of _ cycle 5.
All batch 6 and 7 assemblies
_ will be shuffled to new locations with twenty batch 6 and twenty batch 7A FAs. on - the core periphery.
The fresh batch 8 assemblies - will be loaded
-into the core interior in a symmetric checkerboard pattern.
Figure 3-2 is an eighth-core map showing the burnup and initial enrichment of each assembly at the beginning of cycle (B0C) 6.
- Cycle 6 will be operated in a feed-and-bleed mode.
Core reactivity is con-trolled by 60 full-length Ag-In-Cd control rod assemblies (CRAs), 44 BPRAs, 3-1 Batacock&MIcom a McDermott company
....o.
In addition to the full-length CRAs, eight Inconel axial power shaping rods (gray APSRs) are provided for additional control of the axial power distribution.
The cycle 6 locations of the 68 control rods,'with their respective designations are indicated in Figure 3-3.
The cycle 6 locations and enrichments of the BPRA clusters are shown in Figure 3-4.
The gray APSRs will be withdrawn at 400 t10 EFPD.of operation.
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Figure 3-1.
Fuel Shuffle for Crystal River 3 Cycle 6 1
FUEL TRAriSFER CA?;AL I
6A 7A 7A 7A 6A A
P12 M04 EOS M12 PO4 6A 7A 7A 8
6A 8
7A 7A 6A B
R09 NO3 N02 F
BIO F
M14 N13 R07 68 8
7A 8
7A 8
7A 8
7A 8
68 C
RIO F
LO3 F
K04 F
K12 F
L13 F
L01 6A 8
7A 8
7A 8
68 8
7A 8
7A 8
6A D
K15 F
207 F
LOS F
C13 F
L11 F
C14 F
K01 7A 7A 8
6A 8
78 7A 78 8
6A 8
7A 7A E
C12 C10 F
B06 F
K06 R08 K10 F
F14 F
C06 C04 6A 7A 8
7A 8
68 8
4A 8
68 8
7A 8
7A 6A F
N14 Bil F
E10 F
L15 F
E08 F
R06 F
E06 F
B05 NO2 CY4 7A a
7A 8
78 8
7A 7A 7A 8
78 8
7A 8
7A G
011 F
009 F
F09 F
H13 B09 008 F
F07 F
007 F
005 7A 6A 8
68 7A 4A 7A 4E 7A 7A 68 8
6A 7A
-Y W~"
(h H05 F02 F
C03 H15 pg}
G02 Kg{
K14 hcl 013 F
L14 H11 7A 8
7A 8
78 8
7A 7A 7A 8
78 8
7A 8
7A K
N11 F
N09 F
LO9 F
C08 P07 H03 F
LO7 F
N07 F
N05 6A 7A d
/A 8
68 8
4A 8
68 8
7A 8
7A 6A L
014 P11 F
M10 F
A10 F
MOS F
F01 F
M06 F
P05 002 CY4 7A 7A 8
6A 8
78 7A 78 8
6A 8
7A 74 M
012 010 F
LO2 F
G06 A08 GIO F
P10 F
006 004 6A 8
7A 8
7A 8
68 8
7A 8
7A 8
6A M
G15 F
K02 F
F05 F
003 F
F11 F
P09 F
G01 68 8
7A 8
7A 8
7A 8
7A 8
68 0
F15 F
F03 F
G04 F
G12 F
F13 F
A06 6A 7A 7A 8
6A 8
7A 7A 6A p
A09 003 E02 F
P06 F
E14 013 A07 6A 7A 7A 7A 6A R
812 E04 N08 E12 804 I
Z 2
3 4
5 6
7 8
9 10 11 12 13 14 15 Batch
~
Previous Cycle Location Discharge Cycle of Re-inserts 3-3 Babcock & WHcom a McDermott company
l Figure 3-2.
Enrichment and B0C Burnup Distribution for Crystal River 3 Cycle 6 After 490 EFPD Cycle 5 8
9 10 11 12 13 14 15 2.64 3.29 2.64 3.29 2.95 3.49 2.62 3.29 H
17,272 17,161 16,610 11,918 21,175 0
22,097 20,270 3.29 3.49 2.95 3.49 3.29 3.49 3.29 K
19,685 0
18,936 0
20,038 0
19,012 2.95 3.49 3.29 3.49 3.29 2.62 L
17,780 0
20,098 0
13,138 19,614 2.62 3.49 3.29 3.29 M
22,093 0
18,200 16,014 3.29 3.49 2.62 N
17,168 0
19,093 2.95 0
17,753 P
R X.XX Enrichment, Initial XXXXX Burnup (mwd /mtu), BOC 3-4 Babcock &Wilcox a McDermott company I
Figure 3-3.
Control Rod Locations and Group Designations for Crystal River 3 Cycle 6 Fuel Transfer l
Canal A
B 1
6 1
C 2
5 5
2 D
7 8
7 8
7 E
2 5
4 4
5 2
F 1
8 6
3 6
8 1
G 5
4 3
3 4
5
-Y H
W-6 7
3 3
7 6
K 5
4 3
3 4
5 L
1 8
6 3
6 8
1 it 2
5 4
4 5
2 i
N l
7 8
7 8
7 l
l 2
5 5
2 0
P l
l 1
6 1
R I
i Z
2 3
4 5
6 7
8 9
10 11 12 13 14 15 X
Group Number Group No. c' Rocs Function 1
8 Safety Safety 2
o 3
8 Safety 4
8 Safety 5
12 Control 6
8 Control 7
8 Control 8
8 APSRs Total 68 3-5 Babcock & Wilcox a McDermott company
1 Figure 3-4.
LBP Enrichment and Distribution for Crystal River 3 Cycle 6 8
9 10 11 12 13 14 15 1.1 K
1.4 1.4 i
L 1.4 1.4 0.8 M
1.4 1.4 N
1.4 1.4 0
1.1 0.8 l
P R
Concentration, No. of wt. 8 C X.X LBP Concentration BPRAs 4
(wt % B C in Al 0 )
32 1.4 4
23 4
1.1 8
0.8 Total 44 3-6 Babcock &Wilcox a McDermott company
. ~., _,,. _,,
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t 4.
FUEL SYSTEM DESIGN r
4.1.
Fuel Assembly Mechanical Design i
The types of fuel assemblies and pertinent fuel design parameters for Crystal River Unit 3 cycle 6 are listed in Table 4-1.
The batch 8 fuel assembly holddown springs are made from Inconel 718 material which provides added margin over the previous spring design.
All of the fuel assemblies are mechanically interchangeable.
Retainer assemblies will be used on the two fuel assemblies that contain regenerative neutron source assemblies and on the 44 batch 8 assemblies that contain BPRAs.
The justification for the design and use of these retainers is described in references 2 and 3.
4.2.
Fuel Rod and Gray APSR Designs There has been a change in the pellet design for batch 8 fuel rods.
The fuel pellet length / diameter (L/D) ratio has been decreased from 1.63 to 1.18.
This change in L/D ratio will not adversely affect fuel performance and is expected to decrease local cladding strain at high burnups.
The results of the mechanical evaluations of the fuel rods and gray APSRs are discussed below.
4.2.1.
Cladding Creep Collapse Batch 6 fuel is more limiting than batches 4, 7, and 8 because of its longer previous incore exposure time.
The batch 6 assembly power histories were analyzed to detennine the most limiting three-cycle power history for creep collapse.
This power history was then compared to the power history used in a generic creep collapse analysis which includes increased peaking due to the effects of very low leakage (VLL) fuel cycles.
The generic analysis was performed based on reference 4 and is applicable to the batch 6 design. Pre-dicted creep collapse time based on the generic analysis was greater than 4-1 Babcock &WHcom i
a McDermott company l
33,000 effective full power hours (EFPH), which is greater than the maximum expected residence time of 30,038 EFPH.
The gray APSRs that are to be used in cycle 6 were designed to improve creep life.
Cladding thickness and rod ovali ty control, which are the primary factors controlling the creep life of a stainless steel material, have been improved to extend the' creep life of the gray APSR.
The minimum design clad-ding thickness of the Mark B APSR is 18 mils, while that of the gray APSR is 24 mils.
Additionally, the gap width between the end plug and the Inconel absorber material was reduced.
Finally, the ovality in the gap area will also be controlled to tighter tolerances.
The gray APSR is shown in Figure 4-1.
4.2.2.
Cladding Stress The Crystal River 3 cycle 6 stress parameters are enveloped by a conserva-tive fuel rod stress analysis.
The methods used for the analysis of this cycle have been used in previous cycles.
The gray APSR design was analyzed to demonstrate that it meets specified de-sign requirements.
The APSR was analyzed for cladding stress due to pres-sure, temperature, and ovality.
It was found that the gray APSR has sufff-cient cladding and weld stress margins.
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 plastic strain is less than 1% at design local pellet burnup and heat genera-tion rate.
The design burnup and heat generation rate are higher than the worst-case values cycle 6 fuel is expected to experience.
The strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density and the lower tolerance value for the cladding inside diameter (10).
The gray APSR was analyzed for cladding strain due to thermal and irradia-tion swelling.
The results of this analysis showed that no cladding strain is induced due to thermal expansion or irradiation swelling of the Inconel absorber.
4-2 Babcock & WHcom a McDermott company
4.3.
Thermal Design All fuel in the cycle 6 core is thermally similar.
The fresh batch 8 fuel inserted for cycle 6 operation introduces no significant dif ferences in fuel thermal performance relative to the fuel remaining in the core.
The thermal analyses for all fuel were performed with the TAC 025 code.
Nominal undensified input parameters used in this methodology are provided in Table 4-1.
Densification effects are accounted for in the TACO? code densifica-tion model.
Linear heat rate (LHR) to fuel melt capability for all fuel was determined with the TAC 02 fuel pin performance code.
The analysis performed for cycle 6 demonstrates that 20.5 kW/ft is a conservative limit to preclude center-line fuel melt (CFM) for all fuel batches.
The maximum fuel rod burnup at the end of cycle (E0C) 6 is predicted to be less than 40,500 mwd /mtU.
Fuel rod internal pressure has been evaluated with TAC 02 for the highest burnup fuel rod and is predicted to be less than the nominal reactor coolant system pressure of 2200 psia.
4.4.
Material Design The batch 8 fuel assemblies are not new in concept, nor do they utilize dif-ferent materials from the previous batches.
These materials have been used with success in similar reactor environments.
Therefore, the chemical com-patibility of all possible fuel-cladding-coolant-assembly interactions for the batch 8 fuel assemblies is acceptable.
4.5.
Operating Experience Babcock & Wilcox (B&W) operating experience with the Mark B 15x15 fuel as-sembly has verified the adequacy of its design.
As of October 31, 1984, the following experience has been accumulated for eight B&W 177-fuel as-sembly plants using the Mark B fuel assembly:
4-3 Babcock & WHcom a McDermott company
Max FA burnup, mwd /mtU Cumulative net Current electric Reactor cycle Incore Discharged output, MWh Oconee 1 8
39,645 50,598 53,858,413 Oconee 2 7
33,440 36,800 48,891,084 Oconee 3 8
31,090 35,463 48,970,284 f
Three Mile Island 5
25,200 32,400 23,840,053 Unit 1 Arkansas Nuclear 7
33,890 36,820 42,862,522 One, Unit 1 Raacho Seco 6
33,350 38,268 36,303,116 Crystal River 3 5
28,550 29,900 31,985,818 Davis-Besse 5
28,165 32,790 23,290,256 e
t 4-4 Babcock &Wilcox a McDermott company
Table 4-1.
Fuel Assembly Design Parameters Batch No.
4A 4E 6A 6B 7A 7B 8
Fuel assembly type Mk-B4 Mk-B4 Mk-B4 Mk-84 Mk-84 Mk-B4 Mk-B4 Number of assemblies 4
1 24 12 68 8
60 Fuel rod 00, in.
0.430 0.430 0.430 0.430 0.430 0.430 0.430 Fuel rod ID, in.
0.377 0.377 0.377 0.377 0.377 0.377 0.377 Flexible spacer type Spring Spring Spring Spring Spring Spring Spring Rigid spacer type Zirc-4 Zirc-4 Zirc-4 Zi rc-4 Zirc-4 Zirc-4 Zirc-4 Undensified active fuel 143.6 143.6 141.8 141.8 141.8 141.8 141.8 length, in.
Fuel pellet diameter, in.
0.3697 0.3697 0.3686 0.3686 0.3686 0.3686 0.3686
[
Fuel pellet initial density, 94 94 95 95 95 95 95 1
% TD Initial fuel enrichment 2.64 2.64 2.62 2.95 3.29 2.95 3.49 wt % 235U Estimated residence time, EFPH 22,130 21,523 30,038 30,038 21,960 21,960 10,200 i
Cladding collapse time, EFPH
>33,000
>33,000
>33,000
>33,000
>33,000
>33,000
>33,000 Nominal LHR, kW/ft at 2544 MWt 5.62 5.62 5.69 5.69 5.69 5.69 5.69 LHR capability. '<W/f t to CFM 20.5 20.5 20.5 20.5 20.5 20.5 20.5
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5.
NUCLEAR DESIGN i
- 1 5.1.
Physics Characteristics f
Table 5-1 compares the core physics parameters for the cycle 5 and 6 de-6 and the values signs.
The values for cycle 5 were generated using PDQ07 7
for cycle 6 were calculated with the N0ODLE code.
The cycle burnup (B0C to E0C) will be smaller for cycle 6 than for cycle 5 because of the shorter cycle 6 length. Figure 5-1 illustrates a representative relative power dis-tribution for the BOC 6 at full power with equilibrium xenon and nominal rod positions.
Differences in cycle length, feed batch size and enrichment, BPRA loading, shuffle pattern, and rod group designations for cycle 6 account for the differences in the physics parameters from those of cycle 5.
The critical boron concentrations for cycles 5 and 6 are given in Table 5-1.
The control rod worths differ between cycles due to the gray APSRs and changes in radial flux and burnup distributions.
Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits pre-sented in section 8.
The maximum stuck rod worths are less at BOC 6 and greater at E0C 6 than those for cycle 5.
The adequacy of the shutdown mar-gin with cycle 6 stuck rod worths is demonstrated in Table 5-2.
The follow-ing conservatisms are applied for the shutdown calculations:
1.
Poison material depletion allowance.
2.
107, uncertainty on net rod worth.
3.
Flux redistribution penalty.
Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model.
The shutdown calculation at the end of cycle 6 was analyzed at 400 EFPD and EOC.
The latest time (110 EFPD) in core life at which the APSRs are inserted will be 400 EFPD.
5-1 Babcock &WHeos a McDermott company
5.2.
Changes in Nuclear Design Core design changes for cycle 6 include the removal of the center CRA and the use of gray APSRs.
The center CRA will be replaced with a stand pipe and blind flange.
Removal of the center CRA will have a negligible effect on the nuclear parameters for cycle 6 (excess shutdown margin was decreased by less than 0.04% ak/k at B0C, and 0.03% ak/k at E0C).
Gray APSRs, which
{
are longer and use a weaker absorber (Inconel), replace the silver-indium-cadmium APSRs used in all previous cycles.
Calculations with the standard three-dimensional model veri fied that these APSRs provide adequate axial power distribution control.
As stated in section 5.1, the NOODLE code was used to calculate the physics parameters for cycle 6.
The N0ODLE modeling of the two-group homogenized fuel assembly is the same as that used in PDQ07.
However, the analytical expression N0ODLE uses for the spatial flux solution provides more accurate results than the finite difference expression used in PDQ07 when there are few flux solution points per assembly.
Reference 7 illustrates the calcula-tional accuracy attainable with N000LE in comparison to measured results for various physics parameters.
PDQ07 results are compared to measured data in references 8 and 9.
These comparisons show N0ODLE to be as accu-rate as PDQ07.
The gray APSRs will be withdrawn from the core at 400 110 EFPD in cycle 6 where the stability and control of the core in the feed-and-bleed mode with APSRs removed have been analyzed.
The calculated stability index without APSRs is -0.0381 h-1, which demonstrates the axial stability of the core.
The operational limits and reactor protection system (RPS) limits (Techni-cal Specification changes) for cycle 6 are presented in section 8.
5-2 Babcock &Wilcom a McDermott company
l Table 5-1.
Physics Parameters, Crystal River 3 Cycles 5 and 6(a)
Cycle 5 Cycle 6 Design cycle length, EFPD 460 425 i
Design cycle burnup, mwd /mtU 14,260 13,172
)
Design average core burnup -- EOC, mwd /mtU 21,571 25,291 Design initial core loading, mtU 82.1 82.1 7
Critical boron (b) -- BOC, ppmb, no Xe HZP,. group 8 inserted 1,522 1,665 HFP, group 8 inserted 1,334 1,433 Critical boron -- E0C, ppmb, eq Xe HZP 307 409 HFP (c)
(c)
Control rod worths -- BOC, HFP, %Ak/k Group 7 1.47 1.05 Group 8 0.43 0.21 Control rod worths -- E0C, HFP, %Ak/k Group 7(d) 1.45 1.10 Group 8 0.51 0.23 Max ejected rod worth (8) -- HZP, %Ak/k BOC (N12) 0.62 0.38 E0C (N12, Cy 5; L10, Cy 6) 0.57 0.40 Max stuck rod worth -- HZP, %ak/k BOC (N12)If) 1.72 1.56 E0C (N12) 1.59 1.66 Power deficit -- HZP to HFP, %Ak/k BOC
-1.50
-1.52 E0C
-2.33
-2.29 Doppler coeff, 10-5(ak/k/*F)
BOC, 100% power, no Xe
-1.47
-1.59 E0C, 100% power, eq Xe
-1.73
-1.82 Moderator coeff -- HFP,10-4(ak/k/*F)
BOC,1433 ppspb),no Xe, group 8 inserted
-0.45
-0.44 E0C, 10 ppmbic eq Xe
-2.71
-2.68 Boron worth -- HFP, ppmb/%Ak/k BOC, 1433 ppmb, group 8 inserted 121 129 E0C, 10 ppmb 104 109 5-3 Babcock &Wilcom a McDermott company
Table 5-1. (Cont'd)
Cycle 5 Cycle 6 Xenon worth -- HFP, %Ak/k BOC (4 EFPD) 2.57 2.57 E0C (equilibrium) 2.69 2.70 Effective delayed neutron fraction -- HFP 4
B0C 0.00638 0.00620 E0C 0.00520 0.00521 5
(a) Cycle 6 data are for the conditions stated in this report; the cycle 5 values given are at the core conditions identified in reference 10.
(b) HZP denotes hot zero power (532 T vg); HFP denotes hot full power (579F a
I T vg -
a (c) E0C = 425 EFPD at 100% FP (10 ppmb) for cycle 6; E0C = 450 EFPD at 100%
FP (17 ppmb) with power coastdown to 460 EFPD for cycle 5.
(d) Bank 8 worth at 400 EFPD for cycle 6 and 399 EFPD for cycle 5, the latest time in core life in which it is inserted.
(e) Ejected rod worth for groups 5 through 8 inserted at BOC and groups 5 through 7 inserted at E0C.
(f) Stuck rod worth at 400 EFPD for cycle 6 and 399 EFPD for cycle 5.
J
.f
)
5-4 Babcock &Wilcox a McDermott company
Table 5-2.
Shutdown Margin Calculation for Crystal River 3 Cycle 6
- E0C,
% ak/k
% ak/k
% Ak/k Available Rod Worth 4
Total rod worth, HZP(b) 8.42 9.16 9.03 Worth reduction due to burnup
)
of poison material
-0.42
-0.42
-0.42 Maximum stuck rod worth, HZP
-1.56
-1.66
-1.48 Net worth 6.44 7.08 7.13 Less 10% uncertainty
-0.64
-0.71
-0.71 Total available worth 5.80 6.37 6.42 Required Rod Worth Power deficit, HFP to HZP 1.52 2.29 2.29 Max allowable inserted rod worth 0.30 0.60 0.60 Flux redistribution 0.82 1.20 1.20 Total required worth 2.64 4.09 4.09 Shutdown Margin Total available minus total 3.16 2.28 2.33 required Note: The required shutdown margin is 1.00% tA/k.
(a)The APSR bank is in.
(b)HZP denotes hot zero power and HFP denotes hot full power.
5-5 MM&Mcom a McDermott company
_.________________________]
Figure 5-1.
BOC (4 EFPD), Cycle 6 Two-Dimensional Relative Power Distribution - HFP, Equilibrium Xenon Bank 8 Inserted 8
9 10 11 12 13 14 15 H
0.91 1.08 1.07 1.25 1.08 1.30 0.78 0.45 K
1.12 1.31 1.15 1.34 1.19 1.23 0.47 8
1.13 1.32 1.19 1.30 0.87 0.29 L
M 0.98 1.29 1.05 0.60 N
1.16 1.14 0.35 0
0.50 P
R X
Inserted rod group No.
X.XX Relative power density 5-6 gabcock &WIlcon a McDermott company f
1 6.
THERMAL-HYDRAULIC DESIGN The thermal-hydraulic design evaluation supporting cycle 6 operation used the methods and models described in references 1, 10, 11, and 12.
The cycle 6 reload analysis is the first application of crossflow methodology for Crystal River 3.
Crossflow analyses have been successfully utilized in the licensing of other B&W reload cores.
The use of crossflow models, which can predict flow redistribution effects in an open lattice reactor core, provides significant improvement in departure from nucleate boiling ratio (DNBR) margins relative to the traditional closed-channel modeling.
Thermal-hydraulic crossflow applications for reload cores are fully de-scribed in reference 12.
Cycle 5 and 6 thennal-hydraulic design conditions are listed in Table 6-1.
The design axial peak has been increased from 1.50 to 1.65 to provide addi-tional margin for maneuvering analyses and resulting control rod insertion limits.
The cold nominal stack height has been used in the cycle 6 crossflow analyses.
As discussed in reference 12, the nominal cold stack height is less than the minimum hot stack height.
The minimum hot stack height in-cludes both thermal expansion and densification effects.
The use of the smaller cold stack height is conservative for DNBR analyses.
The minimum DNBR at the design overpower condition (112% FP) is equal to 2.07.
Although the design axial peak has been increased by 10%, the bene-fits of crossflow analyses has resulted in additional DNBR margin relative to cycle 5.
A rod bow penalty has not been considered in the cycle 6 analysis as justi-fied by reference 13.
6-1 Babcock & Wilcox a McDermott company
Table 6-1.
Thermal-Hydraulic Design Conditions Cycle 5 Cycle 6 2544 MWt 2544 MWt Design power level, MWt(a) 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 h
Reference design radial local power peaking factor, FaH 1.71 1.71 g
Reference design axial flux shape 1.5 cosine 1.65 cosine Hot channel factors:
Enthalpy rise, Fq 1.011 1.011 Heat flux, Fq" 1.014 1.014 Flow area 0.98 0.98 Active fuel length, in.
140.2(b) 10.8(c) 176x103 174x103 Average hegt flux at 100% power, Btu /h-ft 452x103 492x103 Maximum hegt flux at 100% power, Btu /h-ft CHF correlation BAW-2 BAW-2 Minimum DNBR (% power) 2.05(112) 2.07(112)
(a)Used in analysis.
(b) Based on densified length.
(c) Cold nominal stack height.
l i
6-2 mg g gg, a McDermott company l
"z 1
r I,
i i
7.
ACCIDENT AND TRANSIENT ANALYSIS 1
7.1.
General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in
[
cycle 6 parameters to determine the effect of the cycle 6 reload and to ensure that thermal performance during hypothetical transients is not degraded.
L y
The effects of fuel densification on the FSAR accident results have been
[
evaluated and are reported in reference 11.
Since batch 8 reload FAs con-l.
tain fuel rods that have a theoretical density higher than those considered in the reference 11 report, the conclusions in that reference are still g
g valid with the exception of the four-pump coastdown and the locked-rotor ac-h cident.
The locked-rotor accident was re-evaluated at 102% of 2568 MWt for
[
cycle 3 operation and remains valid for cycle 6.
The cycle 4 four-pump L
coastdown analysis, performed with an initial power level of 102% of 2544
[
MWt and a pump monitor delay time of 1.5 seconds, bounds cycle 6 and re-f mains valid.
In addition, the single RC pump coastdown analysis accounted for equipment errors and delay times associated wi th the Rosemount fl ow i
transmitters, which replace the BY transmitters used in previous cycles.
7.2.
Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: core ther-j mal parameters, thermal-hydraulic pa rameters, and kinetics pa rameters, including the reactivity feedback coefficients and control rod worths.
?
Core themal properties used in the FSAR accident analysis were design op-erating values based on calculational values plus uncertainties.
The cycle 6 thermal-hydraulic maximum design conditions are compared to the previous 1
cycle 5 values in Table 6-1.
These parameters are common to all the acci-dents considered in this report.
A comparison of the key kinetics param-
[
eters from the FSAR and cycle 6 is provided in Table 7-1.
7-1 Babcock &Wilcox f
a McDermott company E-
A generic loss-of-coolant accident (LOCA) analysis for a B&W 177-FA lowered loop nuclear steam supply (NSS) system has been performed using the Final Acceptance Criteria Emergency Core Cooling System (ECCS) Evaluation Model (reported in BAW-1010314), al ong wi th an upgraded fuel performance model (reported in BAW-177515).
These analyses are generic since the limiting values of key parameters for all plants in this category were used.
Fur-thermore, the combination of average fuel temperatures as a function of LHR and lifetime pin pressure data used in the generic LOCA limit analyses is conservative compared to those calculated for this reload.
Thus, the analy-sis and the LOCA limits reported in BAW-10103 and BAW-1775 provide conserva-tive results for the operation of the reload cycle.
Table 7-2 shows the bounding values for allowable LOCA peak LHRs for Crystal River 3 cycle 6 fuel as a function of burnup.
The LOCA kW/ft limits have been reduced for low burnup to ensure conservative limits based on an interim bounding analyt-ical assessment of NUREG 0630 on LOCA and operating kW/f t limits.
It is concluded from the examination of cycle 6 core thermal and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ability of the Crystal River 3 plant to ope ate safely during cycle 6.
Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 6 is bounded by previously accepted analyses.
The initial conditions for the transients in cycle 6 are bounded by the FSAR, with the exception of the four-pump coastdown and locked-rotor accidents, which were re-evaluated for the conditions discussed in section 7.1.
7.3.
Dose Consequences of Accidents The radiological dose consequences of the accidents presented in chapter 14 of the FSAR were reevaluated for this reload report.
The reason for the reevaluation is that, even though the FSAR dose analyses used a conservative basis for the amount of plutonium fissioning in the core, improvements in fuel management techniques have increased the amount of energy produced by Since 239 u has different fission yields than 235g, P
fissioning plutonium.
the mixture of fission product nuclides in the core changes slightly as the i
239 u to 2350 fission ratio changes, i.e., plutonium fissions produce more P
7-2 Babcock &Wilcou a McDermott company u
of some nuclides and less of other nuclides.
Since the radiological doses associated with each accident are impacted to a different extent by each nuclide and by various mitigating factors and plant design features, the E
radiological consequences of the FSAR accidents were recalculated using the specific parameters applicable to cycle 6.
The bases used in the dose calculations are identical to those presented in the FSAR except as follows:
1.
The fission yields, the half-lives, and the Xe/Kr dose conversion fact--
p ors used in the new calculations are based on more current data.
2.
The steam generator tube rupture accident evaluation considers the in-creased amount of steam released to the environment via the main steam relief and atmospheric dump values because of a slower depressurization due to reduced heat transfer rate af ter tripping the reactor coolant pumps (a post-TMI-2 modification).
3.
The accident evaluations which consider time release doses (CRE, LOCA, and MHA) utilize more conservative 1) iodine species composition, 2) spray removal cycles, and 3) five percentile atmospheric dispersion fac-tors from FSAR chapter 2.
The calculated MHA doses are compared to NRC Safety Evaluation Report 16 (SER) values rather than the FSAR chapter 14 and 14A values.
A comparison of the radiological doses presented in the FSAR and SER to those calculated specifically for cycle 6 (Table 7-3) shows that some doses are higher and some are lower than the FSAR and SER values.
However, all doses are either bounded by the values presented in the FSAR and SER or are a small fraction (10%) of the 10 CFR 100 limits.
The 10 CFR 100 limits are 300 rem to the thyroid or 25 rem to the whole body.
The small increases in some doses are essentially of fset by reductions in other doses.
Thus, the radiological impact of accidents during cycle 6 are not significantly dif-ferent than those described in chapter 14 of the FSAR.
7-3 Sabcock &WHcom J McDermott company
l Table 7-1.
Comparison of Key Parameters for Accident Analysis FSARI Predicted densificationII cycle 6 1
Parameter value value BOL Doppler coeff,10-5 ak/k/*F
-1.17
-1.58 E0L Doppler coeff,10-5 Ak/k/*F
-1.30
-1.82 4
BOL moderator coeff,10-4 ok/k/*F
.0(a)
-0.44 E0L moderator coeff,10-4 ak/k/*F
-4.0(b)
-2.68 All-rod bank worth at BOL, HZP, % ak/k 12.9 8.42 Boron reactivity worth, HFP, ppm /1% ak/k 100 129 Max ejected rod worth, HFP, % ak/k 0.65 0.249 Dropped rod worth, HFP, % ak/k 0.40
<0.20 Initial baron conc., HFP, ppm 1150 1433 (a)+0.50x10-4 ak/k/*F was used for the moderator dilution accident.
(b)-3.0x10-4 Ak/k/*F was used for the steam line failure and dropped rod accident analyses.
Table 7-2.
LOCA Limits for Crystal River 3 LHR, kW/ft 77 days Elevation, ft 0-30 days 30-77 days to E0C 2
13.5 15.0 15.5 4
16.1 16.6 16.6 6
17.5 18.0 18.0 8
17.0 17.0 17.0 10 16.0 16.0 16.0 7-4 Batcock &WHcom A M(Dermott Comparty
Table 7-3.
Comparison of FSAR and Cycle 6 Accident Doses FSAR dose, Cycle 6 rem dose, rem Fuel Handling Accident 2 hr EAB thyroid 10.6 14.0 2 hr EAB whole-body 0.16 0.21 Control Rod Ejection Accident 2 hr EAB thyroid 1.67 2.26 2 hr EAB whole-body 0.878 0.002 30 day LPZ thyroid 0.003 1.13 30 day LPZ whole-body 0.002 0.001 Steam Line Break 2 hr EAB thyroid 0.488 0.45 2 hr EAB whole-body 0.0044 0.002 Steam Generator Tube Failure 2 hr EAB thyroid 0.0043 1.47 2 hr EAB whole-body 0.004 0.13 LOCA 2 hr EAB thyroid 0.55 2.37 2 hr EAB whole-body 0.017 0.008 30 day LPZ thyroid 0.073 0.59 30 day LPZ whole-body 0.011 0.004 Maximum Hypothetical Accident 2 hr EAB thyroid 133.0(a) 93.1 2 hr EAB whole-body 3.0(a) 1.61 30 day LPZ thyroid 25.0(a) 20.7 30 day LPZ whole-body
<1.0(a) 0.30 Waste Gas Decay Tank Rupture 2 hr EAB thyroid 1.44 1.29 2 hr EAB whole-body 1.08 0.37 (a) Dose from NRC Safety Evaluation Report (reference 16).
7-5 Babcock & WHeem a McDermott company
l r
l 8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS All Technical Specifications have been reviewed by Florida Power Corporation and B&W and revisions have been made to accommodate cycle 6 operation.
Table 8-1 lists the Technical Specification changes and cross-references this reload report number with the Technical Specification numbers.
The review of the Technical Specifications based on the analysis presented in this report and the proposed modifications contained in this section en-sure that the Final Acceptance Criteria ECCS limits will not be exceeded nor will the thermal design criteria be violated.
The operating limits present-ed in the proposed Technical Specifications were determined with the method-ology described in reference 17.
8-1 h ock M 8com a McDermott company
Table 8-1.
Technical Specification Changes Tech Spec No.
Report Page Nos.
(Figure, Table Nos.)
(Figure, Table Nos.)
Reason for change (Figure 2.1-2) 8-3(Figure 8-1)
Revised for cycle 6 operation.
(Figure 2.2-1) 8-4(Figure 8-2)
Revised for cycle 6 operation.
2.1 BASES 8-5,8-6,8-7 Revised for cycle 6 operation.
2.2 BASES 8-8,8-9 Revised for cycle 6 operation.
(BASES Figure 2.1) 8-10(Figure 8-3)
Revised for cycle 6 operation 3.1.2.8 8-11 Borated water volume require-3.1.2.9 8-12 ments were revised for cycle 6 x
operation.
3.1.3.6 8-13 Figure Nos. were deleted (Figure 3.1-1) 8-14(Figure 8-4)
(Figure 3.1-la) 8-15(Figure 8-5)
Revised regulating rod group (Figure 3.1-2) 8-16(Figure 8-6) limits for three-and four-(Figure 3.1-3) 8-17(Figure 8-7) pump operation for cycle 6.
(Figure 3.1-3a) 8-18(Figure 8-8)
(Figure 3.1-4) 8-19(Figure 8-9)
(Figure 3.1-7) 8-20(Figure 8-10)
Revised for cycle 6 operation 3.1.3.9 8-21 Figure Nos. were deleted (Figure 3.1-9) 8-22(Figure 8-11)
Revised APSR position limits (Figure 3.1-10) 8-23(Figure 8-12) for cycle 6 operation 3.2.1 8-24 New figure No. was added (Figure 3.2-1) 8-25(Figure 8-13)
Revised axial power imbalance (Figure 3.2-la) 8-26(Figure 8-14) envelope for cycle 6 opera-(Figure 3.2-2) 8-27(Figure 8-15) tion.
(Figure 3.2-2a) 8-28(Figure 8-16) 3.2.2 8-29 Fn was changed. N 3.2.3 8-30 Multiplier in FAH expression was changed.
(Table 3.2.2) 8-31(Table 8-2)
Quadrant power tilt limits were revised for cycle 6 opera-tion.
3/4.1.2 BASES 8-32,8-33 Borated water volume require-ments were revised for cycle 6 operation.
3/4.2 BASES 8-34,8-35,8-36 Revised for cycle 6 operation.
{
(
hock &Micos 8-2 a McDermott company
L Figure 8-1.
Reactor Core Safety Limits (Tech Spec Figure 2.1-2)
-- 120
(-33.8.112)
(31.112)
- 110 Acceptable 4 Pump L
(-48.5,99.6)
Operation
- 100
(-33.8,89.6)
~~ 90 (31,89.6) b (48.2,85.2)
- 80
(-48.5,77.2) 3 & 4 Pump
- 70 Operation g-2 (48.2,62.8)
- 60
$-- 50 3
E-40 44 f- - 30 O
$-- 20 u
E
- 10 a
a a
i i
t i
i i
t i
1
-60
-50
-40
-30
-20
-10 0
10 20 30 40 50 60 Axial Power Imbalance, "
8-3 Babcock & WIIcom a McDermott company 1
Figure 8-2.
Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance (Tech Spec Figure 2.2-1)
(-17,108)
-- 110 (17,108)
"I = 1.0 Acceptable M2 = -1.815 4 Pump
(-34.7,90.3 Operation
- 90 j
(-17,80.67)
(17,80.67)
" " UU (34.7,75.86)
- 70 Acceptable 3 & 4 Pump 5
(-34.7,62.97)
Operation j
j-- 50 (34.7,48.53)
?
,% -- 40 44
- g-- 30 3
o.
g-- 20 t
2a: -- 10 a
i 1
R I
I I
I I
I I
-60
-50
-40
-30
-20
-10 0
10 20 30 40 50 60 Axial Power Imbalance, %
l i
1 8-4 Babcock & Wilcox a McDermott company
L 2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the l
nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation
(
temperature.
Operation above the upper boundary of the nucleate boiling regime would re-sult in excessive cladding temperature because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat trans-fer coefficient.
DNB is not a directly measurable parameter during opera-tion and therefore THERMAL POWER and Reactor Coolant Temperature and Pres-sure have been related to DNB through the BAW-2 DNB correlation.
The DNB correlation has been developed to predict the DNB flux and the location of DN3 for axially uniform and non-unifom heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indica-tive of the margin to DNB.
The minimum value of the DNBR during steady state operation, nomal opera-tional transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curve presented in Figure 2.1-1 represents the coaditions at which a DNBR of 1.30 or greater is predicted for the maximum possible thermal power l 112% when the reactor coolant flow is 139.7 x 106 lbs/hr, which is 106.5%
of the design flow rate for four operating reactor coolant pumps.
This curve is based on the following nuclear power peaking factors wi th potential fuel densification effects:
Fh=2.82; FAH = 1.71; F$=1.65 The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and fo nn the core DNBR design basis.
CRYSTAL RIVER - UNIT 3 B 2-1 8-5 Babcock & Wilcox a McDermott company o.
l 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 outlet pressure, providing a more conservative margin to the safety limit.
The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and poten-tial fuel rod bow:
i J
1.
T e 1.30 DNBR limit produced by a nuclear power peaking factor of 2.82 or the combination of the radial peak, axial peak and l F
=
position of the axial peak that yields no less than a 1.30 DNBR.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.5 kW/f t.
I Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps and three pumps, respective-ly.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1.
The curves of BASES Figure 2.1 represent the conditions at which a DNBR of 1.30 or greater is predicted at the maximum possible thermal power for the l number of reactor coola.1t pumps in operation.
These curves include the potential effects of fuel rod bow and fuel densifi-cation.
1 The DNBR as calculated by the BAW-2 DNB correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher.
Extrapolation of the correlation beyond its published quality range of 22%
is justified on the basis of experimental data.
CRYSTAL RIVER - UNIT 3 B 2-2 8-6 Babcock &WWIIcom a McDermott company
t l
SAFETY LIMITS BASES I
For each curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22% for that particu-situation l lar reactor coolant pump situation.
The DNBR curve for three-pump opera-tion is more restrictive than any other reactor coolant pump c
because any pressure / temperature point above and te the left of the three-pump curve will be above and to the left of the other curves.
2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the-integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the contain-ment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure.
The Reactor Coolant Sys-tem piping, valves and fittings, are designed to USAS B 31.7, February, 1968 Draft Edition, which pennits a maximum transient pressure of 110%,
2750 psig, of component design pressure.
The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code require-ments.
The entire Reactor Coolant System is hydrotested at 3125 psig,125% of de-sign pressure, to demonstrate integrity prior to initial operation.
CRYSTAL RIVER - UNIT 3 B 2-3 8-7 Babcock & Wilcox a McDermott company
LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temperature - High The RCS Outlet Temperature High trip less than or equal to 618*F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.
J Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE i
The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accommodate flow decreasing transients from high power.
The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level setpoint produced by the power-to-flow ratio provides overpower DNB protec-tion for all modes of pump operation.
For every flow rate there is a maxi-mum permissible power level, and for every power level there is a minium per-missible low flow rate.
Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is greater than or equal to 108% and reactor flow rate is 100%, or flow rate is less than or equal to 92.59% and power level is 100%.
2.
Trip would occur when three reactor coolant pumps are operating if power is greater than or equal to 80.67% and reactor flow rate is 74.7%, or flow rate is less than or equal to 69.44% and power is 75%.
For safety calculations the maximum calibration and instrumentation errors for the power level were used.
B 2-5 8-8 Babcock &WHcom a McDermott company
i LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent re-actor thennal limits from being exceeded.
These thermal limits are either power peaking kW/ft limits or DNBR limits.
The AXIAL POWER IMBALANCE re-duces the power level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.
The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbalance bound-l artes by 1.08% for a 1% flow reduction.
V h-RCS Pressure - Low, High, and Variable Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted.
During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high powe r, the RCS Pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint.
The trip setpoint for RCS Pressure-High, 2300 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient.
The RCS Pressure-High trip is backed up by the pressurizer code safety valve for RCS over pressure protection and is therefore set lower than the set pressure for these valves, 2500 psig.
The RCS Pressure-High trip also backs up the Nuclear Overpower trip.
The RCS Pressure-Low,1800 psig, and RCS Pressure-Variable low, (11.59 Tout
- F-5037.8) psig, Trip Setpoints have been established to maintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction.
It also prevents reactor operation. at pressures below the valid. range of DNB correlation limits, protecting against DNB.
Due to the calibration and instrumentation errors, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (11.59 Tout F-5077.8) psig.
B 2-6 8-9 M &MCOE a McDermott company
I Figure 8-3.
Pressure /Teecerature Limits at Maximum Allowable Power for Minimum DNBR (Tech Spec Bases Figure 2.1) 1 CURVE 2 t
2200 3 PUMP 3
E.
2o N
2 c.
4J 000 CURVE 1 5
4 PUMP E8
)
1800 i
1 580 600 620 640 Reactor Outlet Temperature, F Reactor Coolant Flow Pumps operating Curve Flow, lbs/hr (% design)
Power (RTP)
(type of limit) 1 139.7 x 106 (106.5%)
112%
4 pumps (DNBR) 2 104.4 x 106 (79.6%)
89.6%
3 pumps (DNBR) 8-10 Babcock &Wilcon a McDermott company
f REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCES - SHUTDOWN
(-
LIMITING CONDITION FOR OPERATION 3.1.2.8 As a minimum, one of the following borated water sources shall be 4
OPERABLE:
a.
A con entrated boric acid storage system and associated heat trac-
[
ing with:
I I
1.
A ninimum contained borated water volume of 600 gallons, l
2.
Be ueen 11,600 and 14,000 ppm of boron, and 3.
A minimum solution temperature of 105'F.
b.
The borated water storage tank (BWST) with:
1.
A minimum contained borated water volume of 13,500 gallons.
2.
A minimum boron concentration of 2,270 ppm, and 3.
A minimum solution temperature of 40*F.
APPLICABILITY: MODES 5 and 6.
ACTION:
With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATION-or positive reactivity changes until at least one borated water source is restored to OPERABLE status.
SURVEILLANCE REQUIREMENTS 4.1.2.8 The above required borated water source shall be demonstrated OPERABLE:
a.
At least once per 7 days by:
1.
Verifying the boron concentration of the water, 2.
Verifying the contained borated water volume of the tank, and 3/4 1-14 8-11 Batacock &WHcom a McDermott company
l REACTIVITY CONTROL SYSTEMS l
B0 RATED WATER SOURCES -- OPERA' TING LIMITING CONDITION FOR OPERATION 3.1.2.9 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 6000 gallons, l
]
2.
Between 11,600 and 14,000 ppm of boron, and W
3.
A minimum solution temperature of 105*F.
b.
The borated water storage tank (BWST) with:
1.
A minimum contained borated water volume of 415,200 gallons, 2.
Between 2,270 and 2,450 ppm of boron, and 3.
A minimum solution temperature of 40*F.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With the concentrated boric acid storage system inoperable, re-store the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equiva-lent to 1% Ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the con-
)
centrated boric acid storage system to OPERABLE status within the l
next 7 days or be in HOT SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
1 3/4 1-16 8-12 Babcock &WHcom a McDermott company
REACTIVITY CONTROL SYSTEMS REGULATING ROD 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 k
and 7.
APPLICABILITY: MODES 1* and 2*#.
l 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:
a.
Restore the regulating groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.
Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or c.
Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- See Special Test Exceptions 3.10.1 and 3.10.2.
- With Kpff > 1.0.
3/4 1-25 8-13 Batscock &WWilcon a McDermott company
Figure 8-4.
Regulating Rod Group Insertion Limits for Four-Pump Operation From 0 to 200 +10 EFPD (Tech Spec Figure 3.1-1) 110
(
100 (300,102)
{
90 (267,92 h 80
(
J 2
(250,80) 70 E
.E UNACCEPTABLE 60 OPERATION e
{ 50 (175,50) cr
" 40 bg 30 ACCEPTABLE OPERATION 20
(
,15) 10 I
I I
I 0
0 50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25' 50 75 100 t
i I
I
-1 I
I I
i I
Group 5 Group 7 0
25 50 75 100
)
t I
i 1
1 Group 6 8-14 Babcock &Wilcox a McDermott company
Figure 8-5.
Regulating Rod Group Insertion Limits for Four-Pump Operation From 200 10 to 400 10 EFPD (Tech Spec Figure 3.1-la) 110
(
100 (300,102) 90 (267,92) h 80 2
(250,80)
- 70 UNACCEPTABLE y
OPERATION
[ 60 E
50 a
E 175,50) 40 j 30 ACCEPTABLE 20 OPERATION 58,15) 10 0
0 50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25 50 75 100 1
I I
I I
i i
i Group 5 Group 7 0
25 50 75 100 l
i i
1 1
Group 6 1
8-15 Babcock & Wilcox a McDermott company L
l Figure 8-6.
Rei,ulating Rod Group Insertion Limits for Fou -Pump Operation After 400 +10 EFPD (Tech Spec Figure 3.1-2) 110 (265,102)
(300,102) 100 -
(260,92) l 90 b 80 -
250,80) 22 UNACCEPTABLE 70 OPERATION EE 60 r
E 50 3
(175,50) w 40 -
2 y 30 0
ACCEPTABLE 20 OPERATION 58,15) 10 -
0,5) 0 I
I 0
50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25 50 75 100 t
i I
i I
L I
i I
i Group 5 Group 7 0
25 50 75 100 t
i I
I i
Group 6 8-16 Babcock & Wilcox a McDermott company
I L
Figure 8-7.
Regulating Rod Group Insertion Limits for Three-(
Pump Operation From 0 to 200 10 EFPD (Tech Spec Figure 3.1-3) 110
~
100< -
k 90 E
j 80 - -
(275,77)
(300,77)
I 70 2
UNACCEPTABLE (267,69)
I b 60 ATION 7
(250,60) y 50 n
- 40 h
175,37.5) c2 30 20 ACCEPTABLE 10
/ (32,11.75)
OPERATION 0
I i
i i
1 0
50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25 50 75 100 1
I I
I l
l l
l l
g Group 5 Group 7 l
?
25 90 p 1p Group 6 8-17 Babcock & Wilcox a M :Dermott company
Figure 8-8.
Regulating Rod Group Insertion Limits for Three-Pump Operation From 200 10 to 400 10 EFPD (Tech Spec Figure 3.1-3a) 110 100 90 80 (275,77)
(300,77)
' 70 UNACCEPTABLE (267,69)
OPERATION e
$ 60- -
(250,60) 7 50
%" 40 175,37.5) b 30
" 20 ACCEPTABLE OPERATION 10 58,11.75)
(0,4.25)'
0 0
50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25 50 75 100 t
I i
I I
l l
l l
l Group 5 Groi ) 7 0
25 50 7,5 100 Group 6 8-18 Babcock &Wilcox a McDermott company
Figure 8-9.
Regulating Rod Group Insertion Limits for Three-Pump Operation After 400 10 EFPD (Tech Spec Figure 3.1-4) 100 1
90 g 80 (267,77) g (300,77) 70 -
(260,69) i E
(250,60)
~
j 50 UNACCEPTABLE OPERATION a:
}
a 40 i
(175,37.5) 3g.
g a.
20 ACCEPTABLE OPERATION i
10 (58,11.75)
(0,4.25) 0 e
I i
0 50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25 50 75 100 1
I I
I I
I l
t i
i Group 5 Group 7 0
25 50 75 100 t
i i
e i
Group 6 l
l 8-19 Babcock & Wilcox a McDermott company
l r
Figure 8-10 Control Rod Locations and Group Designations for Crystal River 3 Cycle 6 l
(Tech Spec Figure 3.1-7)
Fuel Transfer l
Canal L
A B
1 6
1 L
C 2
5 5
2 D
7 8
7 8
7 s
E 2
5 4
4 5
2 F
1 8
6 3
6 8
1 i-i G
S 4
3 3
4 5
-Y
[
H W-6 7
3 3
7 6
K 5
4 3
3 4
5 L
1 8
6 3
6 8
1 M
2 5
4 4
5 2
i i
s I
7 8
7 8
7 l
l 2
5 5
2 0
P l
l 1
6 1
R i
l E
~
Z s.
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 t
X Group Number Group No. of Rods Function 1
8 Safety 2
8 Safety 3
8 Safety 4
8 Safety 5
12 Control 6
8 Control 7
8 Control 8
_8 APSRs Total 68 I
8-20 Babcock &Wilcox a McDermott company l
REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.9 The axial power shaping rod group sha',' he limited in physical in-sertion as shown on Figures 3.1-9 and 3.1-10.
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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.
Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figure within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or c.
Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- With Keff 1 1.0.
1 3/4 1-37 8-21 Babcock & Wilcox a McDermott company
Figure 8-11.
Axial Power Shaping Rod Group Insertion Limits From 0 to 400 10 EFPD (Tech Spec Figure 3.1-9) 110 UNACCEPTABLE OPERATION
- (0,102)
(100,102) 90 b
80 B
o.
70 ACCEPTABLE OPERATION E
E
[
60 a
e 50 a
E
$e 40 30 20 10 0
I I
I I
I I
I I
i 0
10 20 30 40 50 60 70 80 90 100 Rod Position, % Withdrawn 8-22 Babcock & Wilcox c
a McDermott company
Figure 8-12.
Axial Power Shaping Rod Group Insertion Limits After 400 10 EFPD (Tech Spec Figure 3.1-10) 110 100 90 APSR INSERTION NOT ALLOWED IN THTS TIME INTERVAL 80 s.
E E
70
$w N
60 E
[
50 i
3 40 c.
30 20 10 0
t i
i i
i i
1 0
10 20 30 40 50 60 70 80 90 100 Rod Position, % Withdrawn 8-23 Babcock & Wilcox a McDermott company
5
-=
N 3/4.2 POWER DISTRIBUTION LIMITS
=
AXIAL POWER IMBALANCE 2:
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, 3.2-2, and 3.2-2a.
l C
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 3
minutes, or 5ai b.
Be in at least HOT STANDBY within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
asq SURVEILLANCE REQUIREMENTS
_s
=
4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits E
in each core quadrant at least once every 12 nours when above 40% of DATED THERMAL POWER except when an AXIAL POWER IMBALANCE monitor is inoperable, then calculate the AXIAL POWER IMBALANCE in each core quadrant with an inop-erable monitor at least once per hour.
A
'd ca
- See Special Test Exception 3.7 0.1.
]
b
=
m3
=94 3/4 2-1
-d 7
8-24 Babcock & Wilcox 3
a McDermott company 4
-7
Figure 8-13.
Axial Power Imbalance Envelope for Four-Pump Operation From 0 to 400 +10 EFPD (Tech Spec Figure 3.2-1)
-- 110
(-23.102)
, igg (15,102)
(-25,92)
(15,92)
. 90
(-26,80)
-- 80 (20,80)
I'
- 70 g-- 60 2
(-28,50)
}- - 50 (20,50) ile-
. 40 x,.
h ACCEPTABLE E
UNACCEPTABLE OPERATION
"-- 30 OPERATION s'
e.. 20
-- 10 l
I I
I I
I I
-40
-30
-20
-10 0
10 20 30 40 Axial Power Imbalance, %
l 8-25 Babcock & Wilcox a McDermott company
}
l Figure 8-14.
Axial Power Imbalance Envelope for Four-Pump Operation After 400 +10 EFPD (Tech Spec Figure 3.2-la) l
-- 110
(-21,102)
(15,102)
- Auu
(-25,92)
(15,92)
- 80 (20,80) l
(-28,80)
- 70 h-- 60 8
c.
50 (20,50)
(
(-28,50) j-7-- 40 ACCEPTABLE h
UNACCEPTABLE OPERATION OPERATION 0
- u E-- 20
- 10 I
i I
I l
i f
l l
-40
-30
-20
-10 0
10 20 30 40 Axial Power Imbalance, %
8-2 Babcock & Wilcox a McDermott company
Figure 8-15.
Axial Power Imbalance Envelope for Three-Pump Operation From 0 to 400 10 EFPD (Tech Spec Figure 3.2-2)
.,100
.90
..80
(-23,77)
(15,77)
(-25,69)
(15,69)
(-26,60) b 3
50
(-28,37.5,
- y. 40 (20,37.5) i 3
.30 a
-20 ACCEPTABLE 2
UNACCEPTABLE OPERATION OPERATION I
- -10 i
i t
l l
g
-40
-30
-20
-10 0
10 20 30 40 Axial Power Imbalance, %
8-27 Babcock & Wilcox a McDermott company
Figure 8-16.
Axial Power Imbalance Envelope for Three-Pump Operation After 400 10 EFPD (Tech Spec Figure 3.2-2a)
-- 100 90
-- 80
(-21,77)
(15,77) 70 (15,69)
(-25,69)
(-28,60)
- 60 (20,60)
Ex E-- 50
- 40
(-28,37.5) g-(20,37.5) m
-- 30 a
5-
- 20 ACCEPTABLEOPERATIONj-UNACCEPTABLE OPERATION o.
- 10 a
i l
i t
l g
g
-40
-30
-20
-10 0
10 20 30 40 k
Axial Power Imbalance, %
8-28 Babcock &Wilcox a McDermott company
POWER DISTRIBUTION LIMITS NUCLEAR HEAT FLUX H0T CHANNEL FACTOR - F0 LIMITING CONDITION FOR OPERATION 3.2.2 Fg shall be limited by the following relationships:
3.13
)
Fg <_ 7 l
THERMAL POWER d ere P = RATED THERMAL POWER and P < 1.0.
APPLICABILITY: MODE 1.
ACTION:
With Fg exceeding its limit:
a.
Reduce THERMAL POWER at least 1% for each 1% F0 exceeds the limit within 15 minutes and similarly reduce the NucTear Overpower Trip Setpoint and Nuclear Overpower based on RCS Flow and AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
Demonstrate through in-core mapping that Fg is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
c.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that i
Fn s demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to ex-ceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95%
or greater RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.2.1 Fg shall be determined to be within its limit by using the incore detectors to obtain a power distribution map:
I CRYSTAL RIVER - UNIT 3 3/4 2-4 8-29 Babcock &Wilcox a McDermott company
POWER DISTRIBUTION LIMITS NUCLEARENTHALPYRISEHOTCHANNELFACTOR-F[H LIMITING CONDITION FOR OPERATION 3.2.3 FyH shall be limited by the following relationship:
FN $ 1.h D + 0.3( W l
69 THERMAL POWER where P = RATED THERMAL POWER and P 1 1.0 APPLICABILITY: MODE 1.
ACTION:
With FN exceeding its limit:
H Reduce THERMAL POWER at least 1% for each 1% that F!H exceeds a.
the limit within 15 minutes and similarly reduce the Nuclear Overpower Trip Setpoint and Nuclear Overpower based on RCS Flow and AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
Demonstrate through in-core mapping that F!H is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
c.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by a or b, above; subsequent POWER OPERATION may pro-ceed provided that F H is demonstrated through in-core mapping to be within its lim t at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL PCWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter attaining 95% or greater RATED THERMAL POWER.
3/4 2-6 8-30 Babcock & Wilcox a McDermott company
Table 8-2.
Quadrant Power Tilt Limits (Tech Spec Table 3.2-2)
Steady state Transient Maximum limit limit limit QUADRANT POWER TILT as Measured by:
?
Symmetrical Incore 3.20 9.08 20.0 Detector System Power Range Channels 1.61 6.96 20.0
\\
Minimum Incore 1.73 4.40 20.0 Detector System l
r 3/4 2-11 8-31 Babcock & Wilcox a McDermott company
REACTIVITY CONTROL SYSTEMS BASES l
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY r
E This specification ensures that the reactor will not be made critical with
(
the Reactor Coolant system average temperature less than 525'F.
This limi-E tation is required to ensure that (1) the moderator temperature coefficient t
is within its analyzed temperature range, (2) the protective instrumenta-tion is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor
(
pressure vessel is above its minimum RTNDT temperature, m
a j
3/4.1.2 B0 RATION SYSTEMS L
The boron injection system ensures that negative reactivi ty control is available during each mode of facility operation.
The components required I-to perform this function include (1) borated water sources, (2) makeup or i
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.
r_-
e With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional F
capability in the event an assumed failure renders one of the systems inop-erable.
Allowable out-of-service periods ensure that minor component re-k pair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.
i The boration capability of either system is sufficient to provide a SHUT-DOWN MARGIN from all operating conditions of 1.0%
k/k after xenon decay and cooldown to 200*F.
The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires either 4980 gal-e f
lons of 11,600 ppm boric acid solution from the boric acid storage tanks or 35,681 gallons of 2,270 ppm borated water from the borated water storage i tank.
The requi rements for a minimum contained volume of 415,200 gallons of c
- s borated water in the borated water storage tank ensures the capability for borating the RCS to the desired level.
The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.
_a Therefore, the larger volume of borated water is specified.
Al so, the 6,000 gallons minimum BAST requirement per Specification 3.1.2.9 is conser-vative for this cycle.
--~
?
With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injec-
[
tion system becomes inoperable.
[
B 3/4 1-2
-[
8-32 Babcock &Wilcox a McDermott company g
The boron capability required below 200*F is sufficient to provide a SHUT-00WN MARGIN of 1.0% ak/k after xenon decay and cooldown from 200*F to 140*F.
r This condition requires either 390 gallons of 11,600 ppm boron from the l
boric acid storage system or 1,990 gallons of 2,270 ppm boron from the borated water storage tank.
To envelope future cycle BWST and BAST con-tained borated water volume requirements, minimum volumes of 13,500 gallons L
and 600 gallons, respectively, are specified.
I
)
t-t f
B 3/4 1-2.a 8-33 Babcock & Wilcox a McDermott company
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity dur-ing Condition I (Nonnal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DNBR in the core > 1.30 during nor-mal operation and duri ng short tenn transients, (b) maintaining the peak linear power density < 18.0 kW/ft during nonnal operation, and (c) maintain-ing the peak power deiisity < 20.5 kW/ft during short term transients.
In ad-l dition, the above criteria Tnust be met in order to meet the assumptions used for the loss-of-coolant accidents.
The power-imbalance envelope defined in Figures 3.2-1, 3.2-la, 3.2-2, and 3.2-2a and the insertion limit curves, Figures 3.1-1, 3.1-la, 3.1-2, 3.1-3, 3.1-3a, 3.1-4, 3.1-9, and 3.1-10 are based on LOCA analyses which have de-fined the maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200*F following a LOCA.
Operation outside of the power-imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur.
The power-imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined by Figures 3.1-1, 3.1-la, 3.1-2, 3.1-3, 3.1-3a, 3.1-4, 3.1-9, and 3.1-10, and if the steady state limit QUADRANT POWER TILT exists.
Additional conservatism is introduced by application of:
a.
Nuclear uncertainty factors, b.
Thermal calibration uncertainty.
(
c.
Fuel densification effects.
I d.
Hot rod manufacturing tolerance factors.
The conservative application of the above peaking augmentation factors com-pensates for the potential peaking penalty due to fuel rod bow.
/
The ACTION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensure that the orig-inal criteria are met.
The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:
Fq Nuclear Heat Flux Hot Channel Factor, is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimen-sions.
CRYSTAL RIVER - UNIT 3 B 3/4 2-1 8-34 Babcock & Wilcox a McDermott company
POWER DISTRIBUTION LIMITS BASES N
FAH Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.
It has been detemined by extensive analysis of' possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are me;., provided:
Fg i 3.13; FfH11.71 Power Peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking.
It has been detennined that the above hot channel fac-tor limits will be met provided the following conditions are maintained.
~1.
Control rods in a single group move together with no individual rod insertion differing by more than 6.5% (indicated position) from the group average height.
2.
Regulating rod groups are sequenced with overlapping groups as re-quired in Specification 3.1.3.6.
3.
The regulating rod insertion limits of Specification 3.1.3.6 and the axial power shaping rod insertion limits of Specification 3.1.3.9 are maintained.
4.
AXIAL POWER IMBALANCE limits are maintained.
The AXIAL POWER IM-BALANCE is a measure of the difference in power between the. top and bottom halves of the core.
Calculations of core average axial peaking factors for many plants and measurements from operating plants under a variety of operating conditions have been corre-lated with AXIAL POWER IMBALANCE.
The correlation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained within the limits of Figures 3.2-1, 3.2-la, 3.2-2, and 3.2-2a.
The design limit power peaking factors are the most restrictive calculated at full power for the range from all ~ control rods fully withdrawn to mini-mum allowable control rod insertion and are the core DNBR design basis.
Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are met.
When ugng incore detectors to make power distribution 4_
maps to determine Fg and FAH :
The measurement of total peaking factor, Fheas, shall be increased a.
by 1.4 percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.
CRYSTAL RIVER - UNIT 3 B 3/4 2-2 8-35 Batscoc8: AWWilcon a McDermott company
i POWER DISTRIBUTION LIMITS BASES b.
The measurement of enthalny rise hot channel factor, F$H, shall be increased by 5 percent to account for measurement error.
For Conditinn II events, the core is protected from exceeding 20.5 kW/ft l locally, and from going below a minimum DNBR of 1.30 by automatic protec-tion on power, AXIAL POWER IMBALANCE, pressure and temperature.
Only condi-tions.1 through 3, above, are mandatory since the AXIAL POWER IMBALANCE is an explicit input to the Reactor Protection System.
The QUADRANT POWER TILT limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodi-cally during power operation. For QUADRANT POWER TILT, the safety (measure-ment independent) limit for Steady State is 4.49, for Transient State is l 11.07, and for the Maximum Limit is 20.0.
The QUADRANT POWER TILT limit at which corrective action is required pro-vides DNB and linear heat generation rate protection with x-y plane power tilts.
The limit was selected to provide an allowance for the uncertainty associated with the power tilt.
In the event the tilt is not corrected, the margin for uncertainty on Fo is reinstated by reducing the power by 2 percent for each percent of tilt in excess of the limit.
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the FSAR initial assumptions and have been analytically demonstrated adequate to maintain a DNBR of 1.30 or greater throughout each analyzed transient.
The 12-hour periodic surveillance of these parameters through instrument j
readout is sufficient to ensure that the parameters are restored wi thin
/
their limits following load changes and other expected transient operation.
The 18-month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication chan-nels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12-hour basis.
i CRYSTAL RIVER - UNIT 3 B 3/4 2-3 8-36 Batscock &WHcom a McDermott company
e k
l 9.
STARTUP PROGRAM -- PHYSICS TESTING The planned startup test program associated with core performance is out-
)
lined below. These tests verify that core perfonnance is within the assump-tions of the Final Safety Analysis Report and provide confirmation for con-tinued safe operation of the unit.
9.1.
Precritical Tests 9.1.1.
Control Rod Trip Test Precritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins.
Acceptance criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.66 seconds at these conditions.
It should be noted that Final Safety Analysis Report calculations are based on a rod drop time of 1.40 seconds from fully withdrawn to two-thirds in-serted.
Since the most accurate position indication is obtained from the I
zone reference switch at the 75% inserted position, this position is used for data gathering instead of the two-thirds inserted position.
The Accep-tance Criterion of 1.40 seconds corrected to a 75% inserted position (by rod insertion versus time correlation) is 1.66 seconds.
9.1.2.
Reactor Coolant Flow Reactor coolant (RC) flow with four RC pumps running will be measured at hot zero power, steady-state conditions.
The Acceptance Criteria require that the measured flow be within allowable limits.
9.2.
Zero Power Physics Tests 9.2.1.
Critical Boron Concentration Criticali ty is obtained by deboration at a constant dilution rate.
Once criticality is achieved, equilibrium boron is obtained and the critical 9-1 Babcock & Wilcox a McDermott company
f boron concentration determined.
The critical boron concentration is calcu-lated by correcting for any rod withdrawal required in achieving equilib-rium boron.
The Acceptance Criterion placed on critical boron concentra-tion is that the actual boron concentration must be within 1100 ppm boron of the predicted value.
9.2.2.
Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limit.
f The average coolant temperature is varied by first decreasi ng and then increasing temperature by 5F.
During the changes in temperature, reactiv-ity feedback is compensated by discrete changes in rod motion; the change in reactivity is then calculated by the summation of the reactivity (ob-tained from a reactivity calculation on a strip chart recorder) associated with the temperature change.
The Acceptance Criteria state that the mea-I sured 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 measurements.
After the temperature coef-I ficient has been measured, a predicted value of isothermal fuel Doppler coefficient of reactivity (-2.0 x 10-5 ak/k/*F) is subtracted to obtain the moderator coefficient.
This value must not be in excess of the Acceptance Criteria limit of +0.9 x 10-4 ok/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. This method con-sists of establishing a deboration rate in the RC system and compensating for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 in incremental steps.
The reactivity changes that occur during these measurements are calculated based on Reactimeter data, and dif-ferential rod worths are obtained from the measured reactivity worth versus the change in rod group position.
The differential rod worths of each of the controlling groups are then summed to obtain integral rod group worths.
The Acceptance Criteria for the control bank group worths are as follows:
s 9-2 Babcock &WHcom a McDermott company n
l 1.
Individual bank 5, 6, 7 worth:
predicted value - measured value measured value
-< 15 x 100 l
2.
Sum of groups 5, 6, and 7:
predicted value - measured value x 100
< 10 measured value 9.2.4.
Ejected Control Rod Reactivity Worth After CRA groups 7 and 6 have been positioned near the minimum rod insertion limit and CRA group 5 is between 0 and 10% withdrawn, the ejected rod is borated to 100% withdrawn and the worth obtained by adding the incremental changes in reactivity by boration.
After 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 previously calibrat-ed controlling rod group position.
The boron swap and rod swap values are averaged and error-adjusted to determine ejected rod worth.
Accep tance criteria for the ejected rod worth test are as follows:
y 1.
predicted value - measured value measured value
-< 20 x 100 l
2.
Measured value (error-adjusted) < 1.0% ak/k The predicted ejected rod worth is given in the Physics Test Manual.
9.3.
Power Escalation Tests 9.3.1.
Core Power Distribution Verification at s40, 75, and 100% FP With Nominal Control Rod Position Core power distribution tests are performed at s40, 75, and 100% full power (FP).
The test at 40% FP is essentially a check on power distribution in the core to identi fy any abnormalities before escalat.ing to the 75% FP
- plateau, Rod index is established at a nominal full pover rod configuration at which the core power distribution was calculated.
APSR position is established to provide a core power imbalance corresponiing to the imbalance at which the core power distribution calculations were performed.
9-3 Babcock & Wilcox a McDermott company
The following Acceptance Criteria are placed on the 40% FP test:
1.
The worst-case maximum LHR must be less than the LOCA limit.
2.
The minimum DNBR must be greater than 1.30.
3.
The value obtained fran 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 reactor protec-tion system (RPS) power / imbalance / flow trip envelope.
4.
The value obtained from the extrapolation of the worst-case maximum LHR to the next power plateau overpower trip setpoint must be less than the fuel melt limit, or the extrapolated value of imbalance must fall out-side the RPS power / imbalance / flow trip envelope.
5.
The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.
6.
The highest measured and predicted radial peaks shall be within the fol-lowing limits:
predicted value - measured value measured value
-<8 x 100 7.
The highest measured and predicted total peaks shall be within the fol-lowing limits:
predicted value - measured value measured value
-< 12 x 100 Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and thermal calculational models, thereby verifying the acceptability of da ta from these models for input to safety evaluations.
Items 3 and 4 establish the criteria whereby escalation to the next power plateau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR and LHR.
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:
9-4 Babcock & Wilcou a McDermott company
\\
1.
The highest measured and predicted radial peaks shall be within the fol-lowing limits:
predicted value - measured value x 100
<5 measured value 2.
The highest measured and predicted total peaks shall be within the fol-p lowing limits:
predicted value - measured value x 100
< 7.5 measured value 9.3.2.
Incore Versus Excore Detector Imbalance Correlation Verification at $40% FP Imbalances are set up in the core by control rod positioning.
Imbalances are read simultaneously on the incore detectors and excore power range de-tectors for various imbalances.
The excore detector offset versus incore detector offset slope must be at least 1.15.
If this slope criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.
i 9.3.3.
Temperature Reactivity Coefficient at s100% FP r
The average RC temperature is decreased and then increased by about SF at constant reactor power.
The reactivi ty associated wi th each temperature change is obtained from the change in the controlling rod group position.
Controlling rod group worth is measured by the fast insert / withdraw method.
The temperature reactivi ty coefficient is calculated from the measured changes in reactivity and temperature.
Acceptance Criteria state that the moderator temperature coefficient shall be negative.
9.3.4.
Power Doppler Reactivity Coefficient at s100% FP Reactor power is decreased and then increased by about 5% FP.
The reactiv-ity change is obtained from the change in controlling rod group position.
Control rod group worth is measured using the fast insert / withdraw method.
Reactivity nrrections are made for changes in xenon and RC temperature that occur during the measurement.
The power Doppler reactivity coefficient is calculated from the measured reactivity changes, adjusted as stated above, and the measured power change.
9-5 Babcock & Wilcox a McDermott company
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 10-4 ak/k/% FP.
9.4.
Procedure for Failure to Meet Acceptance Criteria Florida Power Corporation reviews the results of all startup tests to en-sure that all Acceptance Criteria are met.
If the review of the test indi-cates that the results are well within the Acceptance Criteria, no further evaluation is conducted.
If the review indicates that the resul ts are approaching or close to the Acceptance Criteria limits, further evaluation of that particular test or other supporting tests is performed to look for trends. This evaluation will determine whether additional support data are requi red to discover any abnormal conditions.
If Acceptance Criteria for any test are not met, an evaluation is perfonned before the test program is
)
continued.
This evaluation is performed by site test personnel with partic-ipation by B&W technical personnel as required.
Further specific actions depend on evaluation results.
These actions can include repeati ng the tests with more detailed attention to test prerequisites, adding tests to search for anomalies, or design personnel performing detailed analyses of potential safety problems because of parameter deviation.
Power is not escalated until the evaluations show that plant safety will not be com-promised by such escalation.
9-6 Babcock &Wilcox a McDermott company
f 10.
REFERENCES k
1.
Crystal River Unit 3,
Final Safety Analysis Report, Docket 50-302, Florida Power Corporation.
2.
BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg,
Virginia, May 1978.
3.
J.
H.
Taylor to S.
A.
Varga, Le tter, "BPRA Retainer Reinsertion,"
January 1980.
4.
Program to Determine In-Reactor Performance of B&W Fuels -- Cladding Creep Collapse, BAW-10084A, Rev.
2, Babcock
& Wilcox, Lynchburg, Virginia, October 1978.
5.
Y.
H.
- Hsii, et al.,
TACO 2 Fuel Pin Performance
- Analysis, BAW-10141P-A, Rev.1, Babcock & Wilcox, Lynchburg, Virginia, June 1983.
6.
Babcock & Wilcox Version of PDQ User's Manual, BAW-10117P-A, Babcock &
Wilcox, Lynchburg, Virginia, January 1977.
7.
N0ODLE -- A Multi-Dimensional Two-Group Reactor Simulator, BAW-10152, j
Babcock & Wilcox, Lynchburg, Virginia, September 1984.
8.
Comparison of Core Physics Calculations with Measurements, BAW-10120, Babcock & Wilcox, Lynchburg, Virginia, June 1978.
9.
Power Peaking Nuclear Reliability Factors, BAW-10119, Babcock & Wilcox, Lynchburg, Virginia, November 1977.
- 10. Crystal River Unit 3,
Cycle 5 Reload Report, BAW-1767, Babcock &
Wilcox, Lynchburg, Virginia, March 1983.
I
- 11. Crystal River Unit 3, Fuel Densification Report, BAW-1397, Babcock &
Wilcox, Lynchburg, Virginia, August 1973.
- 12. Thermal-Hydraulic Crossflow Applications, BAW-1829, Babcock & Wilcox, Lynchburg, Virginia, April 1984.
10-1 Babcock &Wilcox a McDermott company
y,,
13.
J.
C.
- Moxley, et al.,
Fuel Rod Bowing in B&W Fuel
- Designs, BAW-10147P-A, Rev.1, Babcock & Wilcox, Lynchburg, Virginia.
l 14.
R.
C. Jones, et al., ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103, Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, July 1977.
15.
M. A. Haghi, et al., TAC 02 Loss-of-Coolant Accident Limit Analysis for 177-FA Lowered Loop Plants, BAW-1775, Rev.
0, Babcock & Wil cox, Lynchburg, Virginia, February 1983.
- 16. NRC Safety Evaluation Report, Docket 50302-326, Section 15.0, December 30, 1976.
- 17. Normal Operating
- Controls, BAW-10122A, Rev.
1, Babccck
& Wil cox, Lynchburg, Virginia, May 1984.
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10-2 Babcock &Wilcon a McDermott company
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