ML20209B323
| ML20209B323 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/28/1987 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20209B222 | List: |
| References | |
| BAW-1988, NUDOCS 8704280412 | |
| Download: ML20209B323 (40) | |
Text
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' - BRW-1988 February 1987 CRYSTAL RIVER UlTP 3
- Cycle 7 Reload Report -
8704280412 870415 PDR ADOCK 05000302 P PDR Babcock &Wilcox a McDermott company
BAW-1988 February 1987 CRYSTAL RIVER UNIT 3
- Cycle 7 Reload Report -
BAB00CK & WIIIDX Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 Babcock &Wilcon a McDermott company
00tCENTS Page
- 1. I!ERODUCTION AND SUlHARY . . . . . . . . . . . . . . . ...... 1-1
- 2. OPEPATING HISTORY ........................ 2-1
- 3. GE;TERAL DESCRIPTICN ....................... 3-1
- 4. FUEL SYSTDi DESIGN . . . . . . . . . . . . . . . . . . ...... 4-1 4.1 Ebel Assembly Mechanical Design . . . . . . . . . . . . . . . 4-1 4.2 Ebel Rod Design . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.2.1 Claddiry Creep Collapse . . . . . . . . . ...... 4-2 4.2.2 Claddity Stress ................... 4-2 4.2.3 Cladding Strain ................... 4-2 4.3 Thermal Design ...................... 4-3 4.4 Material Design . . . . . . . . . . . . . . . . . . . . . . 4-3 4.5 Operating Experience .................... 4-3
- 5. lAlclear Design . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Physics Characteristics . . . . . . . . . . . . . . . . . . . 5-1 5.2 Charges in ?Alclear Design . . . . . . . . . . . . . . . . . . 5-2
- 6. THERMAIeHYDRAULIC DESIGN . . . . . . . . . . . . . . . ...... 6-1
- 7. ACCIDDE AND TRANSIDE ANALYSIS . . . . . . . . . . . ...... 7-1 7.1 General Safety Analysis . . . . . . . . . . . . . . . . . . . 7-1 7.2 Accident Evaluation . . . . . . . . . . . . . . . . . . . . .
7.3 Dose Consequences of Accidents 7-1
. . . . . . . . . ...... 7-2
- 8. PROEOSED MDDIFICATIONS 70 TE00iICAL SPECIFICATIONS . . ...... 8-1
- 9. STARTUP PROGRAM - PHYSICS TESTING . . . . . . . . . . ...... 9-1 9.1 Precritical Tests . . . . . . . . . . . . . . . . ...... 9-1 9.1.1 Control Rod Trip Test . . . . . . . . . . ...... 9-1 9.1.2 RC Flow ....................... 9-1 9.2 Zero Power Physics Tests . . . . . . . . . . . . ...... 9-1 9.2.1 Critical Boron Concentration . . . . . . . . . . . . . 9-1 9.2.2 Ter:perature Reactivity Coefficient . . . . ...... 9-2 9.2.3 Control Rod Group / Boron Reactivity Worth . . . . . . . 9-2 9.3 Power Escalation Tests ................... 9-3 9.3.1 Core Power Distribution Verification at Intermediate Power Invel (IPL) and 100% FP With Nominal Control Rod Position . . . . ...... 9-3 Babcock &Wilcox a McDermott company
CONTENTS (Cbnt'd)
Page 9.3.2 Incore Vs. Excore Detector Imbalance Correlation Verification at the IPL ............... 9-4 9.3.3 Tenperature Reactivity Coefficient at 100% FP . . . . 9-4 9.3.4 Power Doppler Reactivity Coefficient at 100% FP . . . 9-5 9.4 Phare for Use if Acceptance Criteria Not Met ...... 9-5
- 10. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 i List of Tables Table 3-1 Cycle 6 Distarged Ebel . . . . . . . . . . . . . . . . . . . . . 3-3 4-1 Ibel Assembly Design Parameters . . . . . . . . . . . . . . . . . 4-5 5-1 Ihysics Panuneters, Crystal River 3, Cycles 6 and 7 . . . . . . . 5-3 5-2 Shutdown ManJ i n Calculation for Crystal River 3 Cycle 7 . . . . . 5-5 6-1 M, vin n Design Conditions, Cycles 6 and 7 . . . . . . . . . . . .
6-2 7-1 Otraparison of Key Parameters for Acx:ident Analysis ....... 7-4 7-2 IOCA Limits for Crystal River 3 . . ............... 7-4 7-3 Ccxnparison of ESAR and Cycle 7 Accident Doses . . . . . . . . . . 7-5 8-1 Tbchnical Specification Changes . . . . . . . . . . . . . . . . . 8-2 List of Fic7ures Figure 3-1 Ebel Shuffle for Crystal River 3 Cycle 7 ............ 3-4 3-2 Enrichment and BOC Burnup Distribution for Crystal River 3 Cycle 7 After 465 EFFD Cycle 6 ................. 3-5 3-3 Control Rod Iccations arx1 Group Designations for Crystal River 3 Cycle 7 . ........................ 3-6 3-4 IEP Enrichment and Distribution for Crystal River 3 Cycle 7 . . . 3-7 5-1 BOC (4 EFPD), Cycle 7 TW-Dimensicnal Relative Power Distribution - HFP, Equilibrium Xe.wt, Bank 8 Inserted . . . . . 5-6 Babcock &Wilcox a McDermott comparty s
- 1. INRODUCTICN AND SIM4ARY
'Ihis report justifies the operation of Ctystal River Unit 3 (cycle 7) at a rated core power level of 2544 MWt. Included are the required analyses to support cycle 7 operation; these analyses enplay analytical techniques and design bases established in reports that have received technical approval by the U.S. Nuclear Regulatory hinaian (NRC; see references) .
'Ihe cycle 7 rated thermal power is 2544 MWt, which was the ultimate core power level identified in the Crystal River Unit 3 Final Safety Analysis Report (FSAR) .1 'Ihe cycle 7 core has been designed with a cycle lifetime of 550 effective full power days (EFPD) and utilizes burnable poison red assablies (BPRAs) to aid in reactivity control.
'Ihe 'Ibchnical Specifications have been reviewed, and the modifications for cycle 7 are justified in this report.
Based on the analyses, performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for emertJency core cooling (ECC) , it has been concluded that Crystal River 3 cycle 7 can be safely operated at a core power level of 2544 MWt.
1-1 Babcock & Wilcox a McDermott company
- 2. OPERATING HISIORY Cycle 6, the current Crystal River Unit 3 operating cycle, is the reference fuel cycle for the nuclear and thermal-hydraulic analyses performed for cycle 7 operation. Cycle 6 achieved criticality on Atxjust 18, 1985, cmpleted power elation testiry on September 16, 1985, and is scheduled for cmpletion in September 1987 after approximately 465 EFPD. No operating anmalies have emwred during previous cycle operations that would adversely affect fuel performance in cycle 7.
Cycle 7 is scheduled to start operation in Dh_" 1987 at a rated power level of 2544 NWt. 'Ihe design cycle length is 550 EFPD.
2-1 Babcock &Wilcox a McDermott company
- 3. GENERAL DESuumON h Crystal River Unit 3 reactor core is described in detail in chapter 3 of the FSAR for the Unit.1 @e core consists of 177 fuel nedlies (FAs),
each of which is a 15-by-15 array containirg 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. 'Ihe FAs in batches 2, 5, 7, 8, and 9 have an average ncminal fuel loadiry of 463.6 kg of uranium, whereas the batch 4 amombly maintains an average naminal fuel loading of 468.6 kg of uranium. 'Ihe claMing is cold-worked Zircalcy-4 with an outside diameter (OD) of 0.430 inch and a wall thickness of 0.0265 inch. 'Ihe 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 7 of Crystal River 3. 'Ihe initial enridmaults of batches 2C, 4B, 5C, 7C, and 8 were 2.54, 2.64, 2.62, 3.29, and 3.49 wt % U, respectively. 'Ihe design enrichment of fresh batch 9 is 3.84 wt % U. Of the thirteen FAs to be re-inserted for cycle 7, the center a membly is a batch 4 FA discharged at the end of cycle 3. 'Ihe eight batch 2 FAs, discharged after cycle 2, occupy interior core locations and four batch 5 FAs, discharged at the end of cycle 4, are on the core perimeter.
Table 3-1 lists the fuel asserrblies that will be discharged at the end of cycle 6. 'IW.nty-four batch 7 and all batch 8 assemblies will be shuffled to new locations with most batch 8 FAs on or near the core periphery. 'Ihe eighty fresh batch 9 assemblies will be loaded into the core 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 7.
Cycle 7 will be operated in a fecxi-and-bleed mode. Core reactivity is controlled by 60 full-length Ag-In-Ctl control rod amMlles (CRAs), 68 BPRAs, and soluble boren shim. In addition to the full-length CPAs, eight Inconel axial power shaping rods (gray APSRs) are provided for additional control of the axial power distribution. 'Ihe cycle 7 locations of the 68 3-1 Babcock & Wilcox a McDermott company
control rods with their respective designations are indicated in Figure 3-3.
The cycle 7 locations and enridments of the BPRA clusters are shown in Figure 3-4. The gray APSRs will be withdrawn at 500 i 10 EFPD of operation.
3-2 Babcock &Wilcom a McDermott company
Table 3-1. Cvele 6 Discharued Riel Batch No. of Assemblies Cycles Burned 4 5 3 6A 24 3 6B 12 3 7A 44 2 7B _a 2 Total Discharged 93 3-3 Babcock &Wilcox a McDermott company
Figure 3-1. Fuel Shuffle for Crystal River 3 Cycle 7 x
I 8 8 9 8 8 A M04 K06 F K10 M12 8 8 9 8 9 8 9 8 9 8 LO3 F NO3 F 008 F N13 F L13 SC 9 8 9 8 9 8 9 8 9 5C C HIS F K02 F K04 F K12 F K14 F R08 CY4 CYA 8 9 8 9 7C 9 2C 9 7C 9 8 9 8 0 C10 F L11 F 806 F P07 F 810 F N06 F C06 CY2 E 9 8 9 8 9 7C 9 7C 9 8 9 8 9 F 809 F M10 F A07 F A09 F LOS F 807 F 8 8 9 7C 9 2C 9 7C 9 2C 9 7C 9 8 8 F 011 C12 F F02 F P09 F POS F K02 F F14 F C04 005 CY2 CY2 8 9 8 9 7C 9 7C 9 7C 9 7C 9 8 9 8 I G F09 F 009 F G01 F B05 F E14 F GIS F 007 F F07 9 8 9 2C 9 7C 9 48 9 7C 9 2C 9 8 9
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b=H F H13 F K14 F M14 F G09 F E02 F G02 F H03 F CY2 CY3 CY2 8 9 8 9 7C 9 7C 9 7C 9 7C 9 8 9 8 K
LO9 F N09 F K01 F M02 F P11 F K15 F N07 F LO7 8 8 9 7C 9 2C 9 7C 9 2C 9 7C 9 8 8 L N11 012 F LO2 F G14 F B11 F B07 F L14 F 004 NCS CY2 CY2 9 8 9 8 9 7C 9 7C 9 8 9 8 9 M F P09 F Fil F R07 F R09 F E06 F P07 F 8 9 8 9 7C 9 2C 9 7C 9 8 9 8 N
010 F E10 F P06 F B09 F P10 F F05 F 006 CY2 SC 9 8 9 8 9 8 9 8 9 SC 0 A08 F G02 F G04 F G12 F G14 F H01 CY4 Cy4 p 8 9 8 9 8 9 8 9 8 F03 F 003 F C08 F 013 F F13 8 8 9 8 8 0 E04 G06 F G10 E12 I
I 2 3 4 5 6 7 8 9 10 11 12 13 14 15 1
Batch Previous Cycle Location Discharge Cycle of Re-inserts 3-4 Babcock & Wilcox a McDermott company I lr _
Figure 3-2. Enrichment and B0C Burnup Distribution for Crystal River 3 Cycle 7 After 465 EFPD Cycle 6 8 9 10 11 12 13 14 15 2.64 3.84 3.29 3.84 2.54 3.84 3.49 3.84 H
17,272 0 24,744 0 17,584 0 18,870 0 3.29 3.84- 3.29 3.84 3.49 3.84 3.49 K
24,732 0 25,719 0 19,822 0 19,499 2.54 3.84 3.29 3.84 3.49 3.49 17,584 0 25,287 0 15,361 19,202 3.49 3.84 3.49 3.84 M
19,991 0 16,664 0 3.49 3.84 3.49
" 20,028 0 18,659 2.62 0
11,754 P
R X.XX Enrichment, Initial XX,XXX Burnup (mwd /mtu), BOC 3-5 Babcock & Wilcox a uccemott company
Figure 3-3. Control Rod Locations and Group Designations for Crystal River 3 Cycle 7 Fuel Transfer Canal )
A B 1 6 1 C 3 5 5 3 D 7 8 7 8 7 E 3 5 4 4 5 3 F 1 8 6 2 6 8 1 r
G 5 4 2 2 4 5 H W- 6 7 2 2 7 6 -Y K 5 4 2 2 4 5 L 1 8 6 2 6 8 1 il , 3 5 4 4 5 3 N l 7 8 7 8 7
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0 l 3 5 5 3 P l l 1 6 1 R i l Z
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 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_
3-6 Babcock & Wilcom a Mcoennott company
1 Figure 3-4. LBP Enrichment and Distribution for Crystal River 3 Cycle 7 8 9 10 11 12 13 14 15 H 0.800 0.800 1.177 K 0.800 0.800 0.800 0.200 L 0.800 1.134 0.800 M 0.800 1.134 1.194 N 0.800 1.194 0.000 0 1.177 0.800 0.000 P 0.200 R
No. of Concentration BPRAs wt % BaC 8 1.194 4 1.177 8 1.134 32 0.800 X.XXX LBP Concentration 8 0.200 (wt % B4 C in Al230) -_
8 0.000 Total 68 3-7 Babcock & Wilcox a Mcoenrott company
- 4. FUEL SYSTai DESIGN 4.1 Fbel Assembly Mechanical Desian The types of fuel amomblies and pertinent fuel design parameters for Crystal River 3 Cycle 7 are listed in Table 4.1. Batch 2C fuel assemblies are of the Mark B3 design. All other batches are of the Mark B4 design.
The Batch 8 and 9 fuel assemblies contain a redesigned holddown spring made frm Inconel 718 material. The new holddown spring design provides added margin over the previous sprirxy design. All of the fuel asserblies are mechanically interchangeable.
The batch 9 fuel uses Zircaloy rather than Inconel as the material for the intermediate spacer grids as reported in reference 2. The NRC safety evaluation 3 of that report requires that a licensee who is incorporating that design suhait a plant-specific analysis of cmbined seismic and IOCA loads according to Appendix A to Standard Review Plan 4.2. The analysis that was presented in reference 2 envelopes the Crystal River 3 plant design requirements. Therefore, the margin of safety reported for the Mark BZ fuel n " mbly is applicable to Crystal River 3.
Batch 2C, Batch 4B and Batch SC are reinserted fuel a = M lles. These fuel neombles and other proposed reinserts were examined using a video camera with the results being taped. Ebel assemblies were examined for evidence of any mechanical characteristics which would prevent their operation in a safe manner.
Examination of the fuel assemblies showed that the fuel assemblies were in good condition with no characteristics that would prevent their reinsertion. Based on the results of the poolside examination these fuel assemblies are capable of good mechanical performance in-reactor.
Retainec assemblies will be used on the two fuel assemblies that contain regenerative neutron source assemblies and on the 68 batch 9 fuel assemblies that contain BPRAs. The justification for the design arxi use of these retainers is described in references 4 and 5.
4-1 Babcock &Wilcox a McDermott company
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l 4.2 Fbel Rod Desian h prepmssure of the batch 9 fuel ammnblies has been lowered 50 psi.
'Ihis was done to provide a higher burnup limit for pin pressure while still providing an adequate creep collapse life. He batch 2C Mark B3 fuel nemnblies have the lowest prepressure, and they are the most limiting fuel annomblies in teIms of Creep collapse. h results of the mechanical evaluations of the fuel rods are dWeal below.
4.2.1 Claddim Creen Collapse Batch 2C fuel is more limiting than batches 4B, SC, 7C, 8 and 9 because of its previous incore exposure and its lower pres.ussure. h batch 2C an=nhly power histories were analyzed to determine a limiting power history for their three cycles of operation. 'Ihis power history was used to model the specific operation of these fuel an=nblies for creep collapse using the methods from reference 6.
All other fuel asenhly power histories were analyzed to determine the most limiting power history for each fuel batch. R ese power histories were then ccxqnrect to the power histories used in the generic creep collapse analyses.
h generic creep collapse analyses were based on the methods frun reference 6.
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All predicted creep collapse times exceed the maximum expected residence times. 'Ihe most limiting fuel batch 2C has a conservatively predicted creep collapse time of 28036 Effective Full Power Hours (EFRI) which exceeds the maximum expected residence time of 27614 EFFH.
i 4.2.2 Claddim Stress
'Ihe Crystal River 3 Cycle 7 stress parameters are enveloped by conservative fuel red stress analyses. 'Ihis includes the lower prepressure batch 9, and the older Mark B3 batch 2C. 'Ihe methods used for the analysis of this cycle have been used in previous cycles.
4.2.3 Claddim Strain h fuel design criteria specify a limit of 1% on cladding plastic tensile circumferential strain. h pellet is designed to assure that cladding plastic strain is less than 1% at design local pellet burnup and heat generation rate. 'Ihe design burnup and heat generation rate are higher than 4-2 Babcock & Wilcox a McDermott company L
the worst case values Cycle 7 fuel is expected to experience. % e strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density and the lower tolerance value for the claddirq inside diameter (ID).
4.3 Wermal Desian All fuel in the cycle 7 core is thermally similar. Se design of the batch 9 Mark BZ nemblies is such that the thermal performance of this fuel is equivalent to the fuel design used in the remainder of the core. he analysis for all fuel was performed with the n002 code as described in reference 7. Ncminal undensified input parameters used in the analysis are presented in Table 4-1. Densification effects were accounted for in n 002, he results of the thermal design evaluation of the cycle 7 core are summarized in Table 4-1. Cycle 7 core protection limits were based on linear heat rate (UIR) to centerline fuel melt limits determined by the TACD2 code.
Batch 2C has the lowest DIR to centerline fuel melt (19.2 kW/ft), but does not limit core operation heatm of its lower enrichment and relatively high prior exposure, which result in a rwhmi power production capability relative to other fuel in the core.
The mvb= fuel n=mbly burnup at EOC 7 is predicted to be less than 43,000 mwd /mtU (batch 7C) . %e fuel red internal pressures have been evaluated with BCD2 for the highest burnup of each fuel rod type and are predicted to be less than the naninal reactor coolant pressure of 2200 psia.
4.4 Material Desian The batch 9 fuel awmblies are not new in concept, nor do they utilize different cm ponent materials. Reinserted fuel a w mblies of similar spent fuel pool residence time have been used in similar reactor environments.
Werefore, the chemical empatibility of all possible fuel-claddirg-coolant-assably interactions for the batch 9 fuel nMMlles, and the batch 2C,4B and SC reinserts is acceptable.
4.5 Operatina Experience n'hwk & Wilcox operating experience with the Mark B 15x15 fuel assembly has verified the adequacy of its design. As of Novmber 30, 1986, the following experience has been accumulated for eight B&W 177 fuel assably plants using the Mark B fuel assably:
4-3 Babcock & Wilcox a McGermott company
CLmulative net Current Max FA burrn20,Mt1/mtU(a) electric Reactor Cycle Incore Discharced cutout Mhh(b)
Oconee 1 10 37,660 50,598 63,104,143 Oconee 2 9 30,620 41,592 58,756,845 Oconee 3 10 39,310 39,229 58,383,983
'Ihree Mile Island 6 20,220 33,640 27,175,718 Arkansas Nuclear 8 47,640 47,560 50,684,806 One, Unit 1 Rancho Seco 7 26,100 38,268 39,066,480 Crystal River 3 6 27,310 31,420 36,058,868 Davis-Besse 5 31,020 32,790 25,233,177 (a) As of November 30, 1986 (b) As of June 30, 1986 4-4 Babcock &Wilcom a McDermott company
Table 4-1. Ebel Acc:=hly Desian Parameters Batch 2C 4B SC 7C 8 9 Ebel a==hly type Mk-B3 Mk-B4 Mk-B4 Mk-B4 Mk-B4 Mk-BZ No of acc:nnblies 8 1 4 24 60 80 Ebel rod OD, in. .430 .430 .430 .430 .430 .430 Ebel rod ID, in. .377 .377 .377 .377 .377 .377 Undensified active 144.0 143.6 141.8 141.8 141.8 141.8 fuel length, in.
Ebel pellet OD, in. .3700 .3697 .3686 .3686 .3686 .3686
. Ebel pellet initial 92.5 94 95 95 95 95 si density, % 'ID Initial fuel enriduiend. 2.54 2.64 2.62 3.29 3.49 3.84 Wt. % U-235 Estimated residence 27614 24523 28750 35986 24360 13200 time, EFHI ClaMing collapse 28036 >37000 >37000 >37000 >37000 35990 time, EFIM Nominal IFR, kW/ft 5.60 5.62 5.69 5.69 5.69 5.69 g at 2544 Mit 2
g IHR capability, kW/ft 19.2 20.5 20.5 20.5 20.5 20.5
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- 5. NUCIEAR IESIGN '
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5.1. Ihysics Characteristics Table 5-1 ocupares the core physics parameters for the cycle 6 and 7 designs. 'Iha values for both cycles were generated using the N00 DIE code.8
'Ihe cycle burnup (BOC to EOC) will be larger for cycle 7 than for cycle 6 because of the longer cycle 7 design length. Figure 5-1 illustrates a representative relative power distrihition for the beginning of cycle 7 at full power with equilibrium xenon and nominal rod positions.
Differences in cycle length, feed batch size and enrichment, BPRA loading, and shuffle pattern for cycle 7 account for the differences in the physics parameters fra those of cycle 6. 'Ihe critical bortm menskations for cycle 6 and 7 are given in Table 5-1. )
'Ihe vad.wl rod wrths differ between cycles due to the changes in ramal flux and burnup distributions. Calculated ejected rod worths and their adherence to criteria are considered at all timaa in life and at all power ,
levels in the devalz- d. of the rod position limits in the Technical Specifications. 'Ihe mavi=im stuck red worths are less in cycle 7 than those for cycle 6. 'Ihe Wiary of the shutdown margin with cycle 7 stuck rod ,
worths is dags4= Lated in Table 5-2. 'Ihe following conservatisms are applied for the shutdown calculations:
- 1. Poison material depletion allowance.
- 2. 10% uncertainty on net rod worth.
- 3. Flux redistribution penalty.
Flux redistribution was accounted for since the shutdown analysis was calculated using a two-dimensional model. 'Ihe shutdown calculation at the end of cycle 7 was analyzed at 500 EFPD and EOC. 'Ihe latest time (i 10 l EFPD) in core life at which the APSRs are inserted will be 500 EFPD.
5-1 M EtM8com a McDermott comparty 1
5.2. Garnes in Nuclear Desian
'Ibe only core design change for cycle 7 is the increase in cycle lifetime to 550 EFPD.
'Ihe gray APSRs will be withdrawn frun the core at 500 i 10 EFPD in cycle 7 where the stability and whvl of tha core in the feed-and-bleed mode with APSRs removed have been analyzed. 'Ihe calculated stability index without APSRs is -0.026 h-1, which chuokates the axial stability of the core.
'Ibe operational limits and Reactor Protection System (RPS) limits (Technical Specification changes) for cycle 7 are disc 1ssed in section 8.
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5-2 Batacock&Wilcon a McDermott company
l Table 5-1. Miysics Parameters, Crystal River 3, Cycles 6 and 7(a) {
1 Cvele 6 Cvele 7 1
Design cycle length, EFPD -
425 550 i
1 Design cycle burnup, Mid/mtU 13,172 17,050 Design average core burnup - EOC, Mid/mtU 25,291 27,898 Design initial core loading, mtU 82.1 82.1 Critical boron - BOC, ppnb, no Xe(b)
HZP, group 8 inserted 1,665 1,976 HFP, group 8 inserted 1,433 1,828 Critical boron - EOC, ppnb, eq Xe HZP 409 224 HFP (c) (c)
Centrol red worths - BOC, HFP, %Ak/k Group 7 1.05 0.95 Group 8 0.21 0.17 Control rod worths - EOC, HFP, %Ak/k Group 7 1.10 1.04 Group 8(d) 0.23 0.19 Max ejected rod worth (e) - HZP, %Ak/k l
BOC (N12) 0.38 0.33 l EOC (L10, Cy 6; N12, Cy 7) 0.40 0.30 Max stuck rod worth - HZP, %Ak/k BOC (M13) 1.56 1.23 EOC (M13)(f) 1.66 1.53 Power deficit - HZP to HFP, %Ak/k i
BOC -1.52 -1.30 EOC -2.29 -2.32 Doppler coeff,10-5 (ak/k 0F)
BOC, 100% power, no Xe -1.59 -1.51 EOC, 100% power, eq Xe -1.82 -1.83 Moderator coeff - HFP,10-4 (ak/k OF)
BOC,1828 ppenb, no Xe, group 8 inserted -0.44 -0.03 1 EOC, 0 ppmb, eq Xe -2.68 I
-2.89 i
5-3 Batacock&WHcom ;
a McDermott company l 1
Table 5-1. Physics , Crystal River 3, Cveles 6 and 7(a (Continued)
Cvele 6 Cvele -7 Boron worth - HFP, ppnb/hk/k BOC, group 8 inserted 129 140 EOC 109 114 Xenon worth - HFP, hk/k BOC (4 EFPD) 2.57 2.47 EDC (equilibrium) 2.70 2.67 Effective delayed neutron fraction - HFP BOC 0.00620 0.00639 EOC 0.00521 0.00522 (a) Cycle 7 data are for the conditions stated in this report; the cycle 6 values given are at the core conditions identified in reference 9.
l (b) HZP denotes hat zero power (532F Tavg); HFP denotes hot full power (579F Tavg)+ l l
(c) EOC = 528 EFPD at 100% FP (2 ppab) for cycle 7; EOC = 425 EFPD at 100%
FP (10 ppnb) with power coastdown to 465 EFPD for cycle 6.
(d) Group 8 worth at 500 EFPD for cycle 7 and 400 EFPD for cycle 6, the l latest time in core life in which it is inserted.
i (e) Ejected rod worth for groups 5 through 8 inserted at BOC and groups 5 through 7 inserted at EOC.
(f) Stuck rod worth at 550 EFPD for cycle 7 and 400 EFPD for cycle 6.
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5-4 Babcock &WHcom a McDermott company
Table 5-2. Shutdown Margin mlonation for Crystal Plver 3 Cycle 7 BOC, 500 EFPD,(a) EOC,
% Ak/k % Ak/k % Ak/k Available Rod Worth Ibtal rod worth, HZP(D) 7.97 8.85 8.84 Worth reduction due to burnup of poisen material -0.42 -0.42 -0.42 Mavi== stuck rod wrth, HZP -1.23 -1.51 -1.53 Net Worth 6.32 6.92 6.89 p less 10% traxu.i.ainty -0.63 -0.69 -Q. M Total available worth 5.69 6.23 6.20 Recuired Rod Worth Power deficit, HFP to HZP 1.30 2.26 2.32 Max allowable inserted red worth 0.30 0.60 0.60 Flux redistribution 0.28 0.65 0.62 Total required worth 1.88 3.51 3.54 Shutdown Mamin Total available minus total required 3.81 2.72 2.66 EZ3: Required shutdown margin is 1.00% ak/k.
(a) The APSR group is in.
(b) HZP denotes hot zero power and HFP denotes hot full power.
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- 5-5 MM EMICOE a McDermott company
Figure 5-1. B0C (4 EFPD), Cycle 7 Two-Dimensional Relative Power Distribution - HFP, Equilibrium Xenon, Bank 8 Inserted 8 9 10 11 12 13 14 15 0.91 1.21 0.93 1.25 0.95 1.30 1.05 0.80 H
l K 0.96 1.30 1.16 1.19 0.54 {
0.93 1.23 L 0.91 1.25 0.97 N 1.28 0.90 0.38 N 1.15 1.30 1.08 0.93 N 1.11 1.09 0.46 0 0.47 P
R Inserted rod group No.
I X.XX Relative power density I :
5-6 nh & WIIcox a uccemett company l
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- 6. 'Ihermal-Hydraulic Design i i
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, 'Ihe thamal-hydraulic design evaluaticn mT!= ting cycle 7 operation util imi the methods and Fvialm described in References 1, 9,10 and 11 as i
supplemented by Reference 2, which inplements the 500 (Reference 12) CHF correlation for analysis of the Mark BZ fuel amas=nbly. Section 5 of Reference 2 demonstrates that a full Mark BZ core and a full Mark B core provide practically the same departure frun nucleate boiling (INB) margin for both steady-state and transient canditions and thus, the current pr==we-tauperature and flux /f1w trip setpoints as r almlated for Mark B are applicable for a similarly-configured Mark BZ core.
'Ibe cycle 7 transition core includes 80 fresh Mark BZ fuel ammamblies, of which 68 contain BPRAs and the Ivannining 12 have open, or unplugged, u. shul i red guide tubes. 'Ihe balance of the core cculsists of Mark B fuel
- amaamblies, 29 of which have open guide tuhaa. 'Ihe Mark BZ anaamblies, with J
their Zircaloy ir*amaritate aparw grids, exhibit a higher pr==ane drop than the Mark B amaamblies. 'Ihis tends to divert scune of the coolant flow fran the Mark BZ to the lower-powered Mark B fuel and creates the need to
, consider a " transition core penalty." 'Ihe core bypass flow is 7.6% for cycle 7. 'Ihe thermal-hydraulic design evaluation considered a full Mark BZ core with a bypass flow of 8.8%. 'Ihis increaaarl bypass flow ===== tion provides sufficient ocxiservatism to offset the transition core effect.
, Table 6-1 provides a sunmary ocuparison of the IEB analysis parameters for cycles 6 and 7.
No rod bow penalty has been considered in the cycle 7 analysis as justified !
l by Reference 13. l 4
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6-1 Balscock&WI8com a McDermott company
-. . _ _ _ . , _ . . . _ _ _ _ _ _ _ _ , _ _ . . _ . _ , ~ , . _
Table 6-1. Maximum Desian Conditions. Cycles 6 and 7 Cycle 6 Cycle 7 2544 MWt 2544 MWt Design power level, MWt (a) 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, gpn 374880 374880 Core bypass flow, % (a) 9.1 8.8
[NBR modeling Crossflow Crossflow Reference design radial-local power peaking factor 1.71 1.71 Reference design axial flux shape 1.65 cosine 1.65 cosine ibt ca nnel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.97 Active fuel length, in. (b) 141.8 141.8 Avg heat flux at 100% power, 103 Btu /h-ft2 174 174 Max heat flux at 100% power, 103 Btu /h-ft2 492 492 OIF correlation B&W-2 BWC OIF correlation DNB limit 1.3 1.18 Minimum NBR at 112% power 2.07 1.77 (c)
(a)Used in the analysis.
(b) Cold nominal stack height.
(c) Calculated for limiting Mk-BZ assembly.
6-2 MItM8com a McDermott company
- 7. ACCIIENT AND 'IRANSIENI' ANALYSIS 7.1. General Safety Analysis Each FSAR accident analysis has been examined with r==Lt to cianges in cycle 7 parameters to determine the effect of the cycle 7 reload and to ensure that thermal performance during hypothetical transients is not degraded.
'Ihe effects of fuel densification on the FSAR accident results have been j evaluated and are zwted in reference 14. Since batd1 9 reload FAs contain fuel rods that have a t h u.etical density higher than those considered in the reference 14 report, the conclusions in that reference are still valid with the exception of the four-puup coastdown and the locked-rotor accident. 'Ihe locked-rotor wirL=1t was re-evaluated at 102% of 2568 :
Mit for cycle 3 operation and zwunains valid for cycle 7. 'Ihe cycle 4 four-punp coastdown analysis, performed with an initial power level of 102% of 2544 Mit and a punp monitor delay time of 1.5 seconds, bounds cycle 7 and ranains valid.
I 7.2. Accident Evaluation l l
'Ihe key parameters that have the greatest effect cm determining the autocme i of a transient can typically be classified in three major areas: core Nmal parameters, hma1-hydraulic parameters, and kinetics parameters, including the reactivity faarihack coefficients and Mul rod worths.
Core thermal properties used in the FSAR accident analysis were design operating values haw on calculational values plus uncertainties. 'Ihe cycle 7 thermal-hydraulic maximum design conditions are compared to the I previous cycle 6 values in Table 6-1. 'Ibese parameters are ccanon to all the accidents considered in this report. A~ ramrison of the key kinetics parameters frcan the FSAR and cycle 7 is provided in Table 7-1.
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A generic loss-of-coolant accident (IOCA) analysis for a B&W 177-FA lowered loop nuclear steam supply (NSS) system has been performed using the Final ,
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7-1 l Babcock &WHcom 1
a McDermott company
_ -- _ .-_.,,._m, , _ _ , , , . m., ._m
! Auc nAai =s Criteria Energency Core Coolirq Systaan (ECCS) Evaluation model (reported in BAW-1010315), alorg with an u a:gc n Gimi fuel performance model .
i (reported in BAN-177516). 'Ihese analyses are generic since the limitirq values of key parameters for all plants in this category were used.
Furthermore, the cambination of average fuel tenparatures as a function of UR and lifetime pin pressure data used in the generic IOCA limit analyses is conservative e=n= red to those calculated for this reload. 'Ihus, the
- analysis . and the IOCA limits lunau.ted in BAW-10103 and BAW-1775 provide ,
conservative results for the operaticm of the reload cycle. Table 7-2 shows the boundirq values for allowable IOCA peak IHRs for Crystal River 3 cycle 7 1 fuel as a functicn of burnup. 'Ihe IOCA kW/ft limits have been r=dinad for
[ low burnup to ensure conservative limits haaad on an interim boundirg l l analytical a=aa==narit17 of NUREG-0630 on IDCA and operatirq kW/ft limits.
It is concluded frun the examination of cycle 7 core thermal and kinetics p%=i. ties, with raanart to acceptable previous cycle values, that this core reload will not adversely affect the ability of the Crystal River 3 plant to operate safely durirg cycle 7. Considerirg the previously accepted design basis used in the FSAR and amhaa%2 ant cycles, the transient evaluation of
! cycle 7 is bounded by previously accepted analyses. 'Ihe initial conditions i for the transients in cycle 7 are bounded by the ESAR, and the four-punp coastdown and locked-rotor accidents, which were previously re-evaluated for the conditions di-maad in section 7.1, and the boron dilution event which was reevaluated for this cycle.
7.3. Dose Consecuen s of Accidents i 'Ihe radiological dose ocmsequences of the accidents s===ded in Chapter 14
) of the ESAR were reevaluated for this reload report. 'Ibe reason for the reevaluation is that, even though the FSAR dose analyses used a conservative I l basis for the amount of plutonium fissionirg in the core, inprovenents in fuel maragu=i techniques have irse the fraction of fissions prrainad I by plutonium. Since 239 Pu has different fission yields than 23b, the
! mixture of fission product nuclides in the core changes slightly as the 239 Pu to b fission ratio changes, i.e., plutonium fissions produce more of sane rnir1idaa and less of other nuclides. 'Ihe radiological doses associated with each accident are inpacted to a different extent by each nuclide and by various mitigating factors and plant design features. 'Ihe l radiological c =;equences of the FSAR accidents were recalculated using the
- 7-2
- a h &WHeens a ucoermore connpeny
l l
vific parameters applicable to cycle 7 as given in the updated ESAR except in the case of the steam generator tube rupture. 'Ihat accident evaluation considers the increased amount of steam released to the envimment via the main steam relief and ati@mric dunp valves because of a slower depressurization due to rdM heat transfer rate after tripping the reactor coolant punps.
A ernmrison of the radiological h presented in the FSAR to those calculated specifically for cycle 7 is given in Table 7-3. '1hese el- are either bounded by the values presented in the FSAR or are a small fraction of the 10 CFR 100 limits, i.e., below 30 Rem to the thyroid of 2.5 Rem to the whole body. 'Ihe small increases in some h are aas:aritially offset by reductions in other riew:aa. 'Ihus, the radiological inpact of accidents during cycle 7 is not significantly different than that described in Chapter 14 of the FSAR.
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l l 7-3 l Babcock &Wilcom I a McDermott company 1
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Table 7-1. -----rison of Kay Parameters for Accident Analysis FSAR1 Predicted ,
densification14 cycle 7 Para ==ter value
{
value BOL Doppler coeff, 10-5 gjyjoF -1.17 -1.51
, EOL Doppler coeff, 10-5 afgjoF -1.30 -1.83 BOL mndarator coeff, 10 4 A/k/OF O(a) -0.03 i
EOL mndarator coeff, 10-4 A/k/OF -4. 0 (b) - -2.89
! All-rod bank worth at BOL, HZP, % A/k 12.9 7.97 l
- Boron reactivity worth, HFP, pps/1% A/k 100 140 l
! Max ejected rod worth, HFP. % A/k O.65 0.23 i
Ds@ red worth, HFP, % A/k O.40 $0.23 i Initial boron conc., HFP, ppa 1150 1828 i
i O
(a)+0.50x10-4 A/k/ F was used for the htor dilution accident.
(b)-3.0x104 A/k/0 F was used for the steam line failure and &# rod accidant analyses.
1 Table 7-2. IDCA Timits for Crystal River 3 i
- IHR kW/ft i.
75 EFPD
! Elevation, ft 0-30 EFPD 30-75 EFPD to EOC
, 2 13.5 15.0 15.5 4 16.1 16.6 16.6 6 16.5 18.0 18.0 8 17.0 17.0 17.0 10 16.0 16.0' 16.0 1
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! 7-4 Balseock&WI8cENE ;
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l Table 7-3. Otznarison of FEAR ard Cycle 7 Accident Doses FSAR dose, Cycle 7 rem dose, rem Ebel Handlim Accident 2 hr EAB thyroid 14.0 12.4 2 hr EAB whole-body 0.19 0.82 Control Rod EMection Accident 2 hr EAB thyroid 0.65 2.25 2 hr EAB whole-body 0.0008 0.002 30 day IBZ thyroid 0.35 1.13 30 day IPZ whole-body 0.0002 0.001 Steam Line Break 2 hr EAB thyroid 0.503 0.51 2 hr EAB whole-body 0.0033 0.002 Steam Generator Tube Failure 2 hr EAB thyroid 0.0043 1.65 2 hr EAB whole-body 0.004 0.074 I.Q:a 2 hr EAB thyroid 2.19 3.01 2 hr EAB whole-body 0.016 0.008 30 day IPZ thyroid 0.517 0.25 30 day IPZ whole-body 0.0081 0.004 Maximum Hvoothetical Accident 2 hr EAB thyroid 86.8 63.1 2 hr EAB whole-body 2.28 1.55 30 day IPZ thyroid 18.40 9.11 30 day IPZ whole-body 0.440 0.29 Waste Gas Decav Tank Ruoture 2 hr EAB thyroid 1.43 1.43 2 hr EAB whole-body 0.92 0.42 l 7-5 l BalBCock &WilCOE a McDermott company
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- 8. PROPOSED PODIFICATIONS TO TEGNICAL SPECIFICATICES All Technical Specifications have been reviewed by Florida Power Corporation and B&W and revisions have been made to aO --- -ute cycle 7 operation.
Table 8-1 lists the Technical Specification changes. The Technical Specification changes are being subnitted to the NRC under separate cover.
This review of the Technical Specifications haw on the analysis presented in this report ensure that the Final Acceptance Criteria ECCS limits will not be e nor will the thermal design criteria be violated. The operating limits presented in the proposed Technical Specifications were determined with the methodology h ibed in reference 18.
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8-1 Balm:ock &WHcom a McDermott company
Table 8-l'. Technical Soecification Chanaes Tech Spec No.
(Fiaure. Table Nos.) Reason for Chanae (Figure 2.1-2) Revised for cycle 7 operation.
(Figure 2.2-1) Revised for cycle 7 operation. ,
2.1 BASES Revised for cycle 7 operation.
3.1.3.6 Figure Nos. were deleted and action statement was changed to incorporate shutdown margin curves. _
3.1.3.9 Time of APSR withdrawal was changed and the IID was reworded.
(Figure 3.1-1) Revised regulating rod group limits for cycle 7.
(Figure 3.1-2) Revised regulating rod group limits for cycle 7.
(Figure 3.1-3) Revised regulating rod group limits for cycle 7.
(Figure 3.1-4) Revised regulating rod group limits for cycle 7. l (Figure 3.1-7) Revised for cycle 7 operation.
3.2.1 Figure Nos. were deleted.
(Figure 3.2-1) Revised axial power imbalance (Figure 3.2-2) envelope for cycle 7 operation.
(Table 3.2-2) Quadrant power tilt limits were revised for cycle 7 operation.
3/4.1.2 BASES Borated water volume requirements were revised for cycle 7 operation, i 3/4.2 BASES Revised for cycle 7 operation. '
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4 8-2 Eh &WI8ces a McDermott company
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- 9. SMRIUP PROGRAM - IEYSICS TESTI'DG ;
'Iha planned startup test program anamiated with core performance is outlined belw. 'Ihese tests verify that core performance is within the
, assumpticms of the safety aalysis and provide information for continued safe operation of the unit.
9.1. Precritical Tests 9.1.1. Control Rod Trio Test i
Precritical control rod drop timaa are recorded for all ocotrol rods at hot full-flw canditions before zero power physics testing begins. %;.ar..w.
criteria state that the rod drop time frm fully withdrawn to 75% inserted i
shall be less than 1.66 seconds at the canditions above.
It should be noted that safety analysis calculations are haamri on a rod drop fra fully withdrawn to two-thirds inserted. Since the most accurate q position indication is obtained frtaa the zona reference switch at the 75%-inserted positicn, this position is used instead of the two-thirds inserted position for data gathering.
) 9.1.2. RC Fl w i
- Reactor coolant f1w with four RC pungs running will be maamwed at hot shutdown ocnditions. Acceptance criteria require that the maamwed flw be
. within allowable limits.
t 9.2. Zero Power Physics Tests 9.2.1. Critical Boron Cor.xnL ation
.I once initial criticality is achieved, equilibrium boron is obtained and the i
1 critical boren cu.xnk ation determined. 'Ihe critical boren concentration j is calculated by correcting for any rod withdrawal required to achieve equilibrium boron. 'Ihe acceptance criterion placed on critical boren
- m a a. cation is that the actual boren m.xukation must be within 100 ppn boron of the predicted value.
! 9-1 maken,arayWHcom a McDermott company
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- 9.2.2. 'D=-ature Dame +1vity Nficiarit
'Iha isotbarmal HZP 'wature coefficient is naamwei at apprtacimately the
)
all-ztxis-cut configuration. During changes in temperature, reactivity ;
faadhack may be ocupensated by control rtxt acmenent. 'Iha change in I reactivity is then calculated by the amenation of reactivity (obtained frta l a reactivity calm 21ator strip chart -
recorder) aannciated with the temperature dange. A - -f a is critaria state that the naamned value shall not differ frta the predicted value by maate than i 0.4x10-4 Ak/k/CF.
'Iha andarator coefficient of reactivity is calculated in conjunction with the temperature coefficient naamwnnent. After the temperature coefficient 1 i
, has been naamwed, a predicted value of fuel Doppler coefficient of l
\
reactivity is addad to obtain the andarator coefficient. 'Ihis value must not be in *=mana of the 7-- Maas critaria limit of +0.9x10-4 Ak/k/CF. -
9.2.3. Ocmtrol Rod Group / Boron Reactivity Worth I a
Ocmtrol rod group reactivity worths (groups 5, 6, and 7) are maa=wed at hot j
t zero power canditions using the baron / red swap method. 'Ihis technique ;
consists of establishing a deboration rate in the reactor coolant system and j ocupensating for the reactivity danges frtat this deboration by inserting ocmtrol red groups 7, 6, and 5 in ira ,_.Lal steps. 'Ihn reactivity changes ;
that occur during these ma==nunents are caloilated based on r==<+i=ater f
l data, and differential rod worths are obtained frca the naaawed reactivity worth versus the change in rod group position. 'Ihe differential red worths I of each of the wibulling groups are then ===ad to obtain integral. rod ,
i i group worths. 'Ihe m.=pLumis criteria for the control bank grtmp worths
! are as follows:
i
- 1. Indivichm1 bank 5, 6, 7 worth:
credicted value - mannured value x 100 < 15
,e measured value -
f 2. Sums of groups 5, 6, and 7:
I credicted value - measured value value x 100 5 10 l
'Ihe boron reactivity worth (differential boron worth) is maamned by i
{
dividing the total inserted rod worth by the boron change mada for the rod f
l 9-2 gg
- j. a McDermott company
,,.--w+-.,%- --,.3- +--.---.-ww_ .,,--%.,g,, -.mg-r- .,7 p-, ,--ey,p., ..,%p--wn..-m,yq- n,.3,.,y. . ,y
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worth test. The acceptance- criterion for Insasured differential boron worth ,
is as follows: l
- 1. oredicted value - measured value ;
g x 100 $ 15 ,
The predicted rod worths and differential boron worth are taken frm the PIM.
9.3. Power Escalation Tests 9.3.1. Core Ibwer Distribution Verification at Irtamaiiate Power Ievel (IPL) and 100% FP With Nminal Control Rod Position Core power distribution tests are perfonned at the IPL and 100% full power (FP). Equilibrium xenon is established prior to both the IPL and 100% FP tests. The test at the IPL is esim d.ially a check on power distribution in the core to identify any abnonnalities before escalatiIg to the 100% FP plateau. Peaking factor criteria are applied to the IPL core power distribution results to determine if additional tests or analyses are required prior to 100% FP operation.
The following acceptance criteria are placed on the IPL and 100% FP tests:
- 1. The worst-case maxinum IHR nust be less than the IOCA limit.
- 2. The mininn ENBR nust be greater than 1.30.
- 3. The value obtained frm extrapolation of the mininum INBR to the next power plateau overpower trip setpoint nust be greater than 1.30, or the extrapolated value of inbalance nust fall outside the RPS power / imbalance / flow trip envelope. ,,
- 4. The value obtained frun extrapolation of the worst-case maxinn IRR to the next power plateau overpower trip setpoint nust be less than the fuel melt limit, or the extrapolated value of imbalance nust fall outside the RPS power / imbalance / flow trip envelope.
- 5. The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.
9-3 Batscock &WilcoE a McDermott company
4
- 6. 'Ihe hicpest maamwed and predicted radial peaks shall be within the l following limits:
" - """ " 1 " x 100 more positive than -5
- maamired value
- 7. 'Ihe highest maamtred and predicted total paaks shall be within the following limits:
valm - measured u lm maamwed value x 100 more positive than -7.5 ;
Items 1, 2, 5, 6, and 7 are established to verify core nuclear and tharmal j calculational andala, thereby verifying the acceptability of data frun these i mndala for input to safety evaluations.
1 Items 3 and 4 establish the criteria whereby aarmlation to full power may be mlished without aw=ading the safety limits specified by the safety analysis with regard to DiBR and linear heat rate.
9.3.2. Incore Vs. Excore Detector Tuhalance J
correlation verification at the IPL J
Tmhalances, set up in the core by w Lul rod positioning, are read
] siimiltaneously on the incore detectors and excore power range detectors.
'Ihe excore detector offset versus incore detector offset slope nost be
{
j greater than 0.96. If this criterion is not met, gain anplifiers on the exoore detector signal processing amir= ant are adjusted to provide the j required gain.
j 9.3.3. humi: retire Paartivity Coefficient at 100% FP
'Ihe average reactor cx)olant t=umature is decreaamd and then inctuaamd by about 5F at w#uiunt reactor power. 'Ibe reactivity associated with eacti touwentre change is obtained from the change in the wibulling rod group position. Controlling rod group worth is maaanwd by the fast insert / withdraw method. 'Ihe tenperature reactivity coefficient is
! calculated frm the measured changes in reactivity and tenperature.
Acceptance criteria state that the mndarator t=u m eture coefficient shall l be negative.
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! 9-4 Babcock &WHeem a McDermott comparty l . - _ -. . - - - . - - - . - . . - - - - . - -
9.3.4. Power Doooler Reactivity Coefficient at 100% FP Reactor power is decr=waaM arxi then incrsu*aM by about 5% FP. 'Ihe reactivity change is obtained frun the change in c.whulling rtxi 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 tenperature that occur during the measurement.
'Ihe power Doppler reactivity coefficient is calculated frun the measured reactivity change, adjusted as stated above, and the measured power change.
'Ihe fuel Doppler reactivity coefficient is calculated in conjunction with the power Doppler coefficient measurement. 'Ihe power Doppler coefficient as measured above is nultiplied by a predicted conversion factor to obtain the fuel Doppler coefficient. 'Ihis measured fuel Doppler coefficient must be more negative than the acceptance criteria limit of -0.90 x 10-5 AM.
9.4. Procedure for Use if Acceptance Criteria Not Met If acceptan criteria for any test are not met, an evaluation is performed before the test s.wtam is continued. 'Ihis evaluation is performed by site test personnel with participation by Ra h k & Wilcox technical pel w ue_1 as required. Ebrther specific actions depend on evaluation results. 'Ihese actions can include repeating the tests with more detailed attention to test prerequisites, ~ir1M tests to seardi for ancmalies, or design personnel performing dr. tailed analyses of potential safety problems because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be c.uvi inM by such escalation.
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9-5 Babcock &WHcom a McDermott company l l
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- 10. REFERENCES
- 1. Crystal River Unit 3, Final Safety Analysis Report, Docket 50-302, Florida Power Corporation.
- 2. BAW-1781P, Rancho Seco Cycle 7 Reload Report - Volume 1 - Mark BZ Ebel Assembly Design Report, RaWk and Wilcox, Lynchburg, Virginia, April 1983.
- 3. Rancho Seco Nuclear Generating Station - Evaluation of ' Mark BZ ' Fuel Assembly Design, U.S. Nuclear Regulatory NMion, Washington, D.C.,
November 16, 1984.
- 4. BPRA Retainer Design, BAW-1496, RaWk & Wilcox, Lynchburg, Virginia, May 1978. -
- 5. J.H. Taylor to S.A. Varga, Istter, BPRA Retainer Reinsertion, January 1980.
- 6. Pregs.cuu to Determine In-Reactor Performance of B&W Fuels--Cladding Creep Collapse, BAW-10084A, Rev 2, Rahk & Wilcox , . Lynchburg, Virginia, -
October 1978.
- 7. BAW-10141P-A Rev. 1, TACO 2: Fuel Performance Analysis, Rm_ W k and Wilcox, Lynchburg, Virginia, June 1983.
- 8. NOODIE - A ML11ti-Dimensional 'IWo-Group Reactor Simulator, BAW-10152-A, Ra h k & Wilcox, Lynchburg, Virginia, June 1985.
- 9. Crystal River Unit 3, Cycle 6 Reload Report, BAW-1860, Ra h k & Wilcox, Lynchburg, Virginia, April 1985.
- 10. BAW-10156-A, LYNXT: Core Transient 'Ihermal Hydraulic Analysis Code, Rahk & Wilecx, Lynchburg, Virginia, February 1986.
- 11. BAW-1829, 'Ihermal-Hydraulic Crossflow Applications, RaWk and Wilcox, Lynchburg, Virginia, May 1984.
- 12. BAW-10143P-A, BWC Correlation of Critical Heat Flux, RaWk & Wilcox, Lynchburg, Virginia, April 1985.
10-1 m w &WUHeen A MCDer#nott compa,ty
- 13. BAW-10147P-A. Rev.1, Ebel Rod Bowirq in hWk & Wilocx Fuel Design, h W k and Wilcox, Iynchburg, Virginia, May 1983.
- 14. BAW-1397, Crystal River Unit 3, Ebel Densification Report, h W k & Wilcox, Lynchburg, Virginia, August 1973.
- 15. R. C. Jones, et al., ECCS Analysis of B&W's 177-FA IcWered loop NSS, BAW-10103, Rev. 3, hWk & Wilcox, Lynchburg, Virginia, July 1977.
- 16. M. A. Haghi, et al., TACD2 Icss-of-Coolant Accident Tdmit Analysis for 177-FA Iowered Icop Plants, BAW-1775, Rev. O, hWk & Wilcox, Lynchburg, Virginia, February 1983.
- 17. Bounding Analytical #t==: ment of NUREG-0630 Models on ICX1 kW/ft Limits with Use of FIECSET, BAW-1915P, hWk & Wilcox, Lynchburg, Virginia, May 1986.
- 18. Normal Operating Controls, BAW-10122A, Rev. 1, *Wk & Wilcox, Lynchburg, Virginia, May 1984.
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