Text

Cycle 6-Reload Rept
	ML20154H368
Person / Time
Site:	Davis Besse
Issue date:	04/30/1988
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20154H301	List: ML20154H298; ML20154H326; ML20154H368; ML20154H398;
References
	1516, BAW-2038, NUDOCS 8805250315
	Download: ML20154H368 (77)
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I Docket Nd. 50-346 BAW-2038 l License No. NPF-3 April 1988 Serial No. 1516 l Attachment 3 l

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L DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 6 -- RELOAD REPORT

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BAW-2038 April 1988 DAVIS-BESSE NUCEAR ICWER STATION j UNIT 1, CYCE 6 - REIDAD REPORT l ~

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1. IIEROIIJC1' ION AND

SUMMARY

, . . . . . . . . . . . . . . . . . . . . 1-1

2. OPERATING HIS'IORY . , .. ........... . . . ...... 2-1

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3. GDIERAL DESCRIPI' ION . ... ......... . . . . ...... 3-1

4. FUEL SYSTD4 DESIGi . . ..... . .. .. . . . . . . ...... 4-1 1 4.1. Fuel Assenbly Mechanical Design . . .. . . . . ...... 4-1 4.2. Fuel Rod and Gray APSR Design . . . . . . . . . ...... 4-1 4.2.1. Cladding Collapse . . . . . . . . . . . ...... 4-1 f A. Fuel Rod . . . . . . . .. . . . . . ...... 4-1 B. Gray APSR ... . . . . . . . . . . ...... 4-1

, 4.2.2. Cladding Stress . . .. . . . ,. . . . . . . ... . . 4-2 A. Ebel Rod . . . . . . . . . . . . . . . . . . . . 4-2 i B. Gray APSR ... . . . . . . . . . . ...... 4-2 4.2.3. Cladding Strain . . . . . . . . . . . . ...... 4-2 A. Ebel Rod . .. . .. . . . . . . . . ...... 4-2 B. Gray APSR .. . . . . . . . . . . . ...... 4-2 4.3. 'Ihermal Design . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4. Material Ccr:patibility . . . . . . . . . . . . . . . . . . . 4-3 4.5. Operating Experience . . . . . . . . . . . . . . . . . . . . 4-3

5. NUCIEAR DESIQi . . ..... .. . . . . . . . . . . . ...... 5-1 5.1. Physics Characteristics . . . . . . . . . . . . ...... 5-1 5.2. Changes in Nuclear Design . . . . . . . . . . . ...... 5-2

6. 'IHERMAIrHYDRAUIlC DESIGN . .......... . . . . ...... 6-1

7. ACCIDDC AND 'IRANSIDE ANALYSIS . . . . . . . . . . . ...... 7-1 7.1. Ceneral Safety Analysis . . . . . . . . . . . . ...... 7-1 7.2. Accident Evaluation .... .. . . . . . . . . ...... 7-1

8. PROICSED MODIFICATIONS 'IO TEODIICAL SPECIFICATIONS . . ...... 8-1

9. STARIUP PROGRAM - FHYSICS TESTING . . . . . . . . . . ...... 9-1 9.1. Precritical 'Ibsts .... . . . . . . . . . . . ,..... 9-1 9.1.1. Control Rod Trip Test . . . . . . . . . ...... 9-1 9.1.2. RC Flow .... . . .. . . . . . . . . ...... 9-1 9.2. Zero Power Ihysics Tests . . . . . . . . . . . . . . . . . . 9-1 9.2.1. Critical Boron Concentration . . . . . . . . . . . . 9-1 9.2.2. Temperature Reactivity Coefficient . . . ...... 9-2 9.2.3. Conttrl Rod Group / Boron Reactivity Vbrth . . . . . . 9-2 Babcock &Wilcox a McDermott company

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9.3. Power Escalation Tests . ... . .. .......... . . 9-3 9.3.1 Core Symmetry Test . .. .. ... . ... . . . . . 9-3 9.3.2. Core Power Distribution Verificaticn at Intermediate Power I.evel (IPL) and 100%FP With Naninal Control Rod Position . ... . . . . . 9-3 9.3.3. Incore Vs. Excore Detector Imbalance Correlation Verification at the IPL ..... . . . 9-4 9.3.4. 'hrparature Reactivity Cbefficient at %100%FP . . . . 9-4 9.3.5. Power Doppler Reactivity Coefficient at N100%FP . . 9-5 (

9.4. Procedure for Use if Acceptance Criteria Not Met . . . . . . 9-5

10. REFERENCES . .. . . . . . . . ... .. . .... ... . . . . . 10-1 List of Tables Table f 4-1. Fuel Design Parameters . . ... .. . . ... .. ...... 4-5 5-1. Davis-Besse Unit 1, Cycle 6 Physics Parameters ..... . . . 5-3 5-2. Shutdown Margin Calculation for Davis-W, Cycle 6 . . . . . 5-4 '

6-1. 'Ihermal-Hydraulic Design Cbnditions . . . . . . ... . . . . . 6-3 7-1. Conparison of Key Parameters for Accident Analysis ...... 7-3 7-2. Bcuniing Values for Allowable IDCA Peak Linear Heat Rates . . . . . .... ..... . ... . . . . . 7-3 8-1. Reactor Protection System Instrumentation Trip Setpoints i (Tech. Spec, Table 2.2-1) . ... ... ... . ... . . . . . 8-4 8-2. Quadrant Power Tilt Limits (Tech. Spec. Table 3.2-1) . . . . . 8-31 8-3. EiiB Margin (Tech. Spec. Table 3.2-2) . ... . .. ...... 8-33 8-4. Reactor Protection System Instrumentation Surveillance Requirements (Tech. Spec. Table 4.3-1) . ... ... . . . . . 8-34 -

List of Fiaures Figure 3-1. Davis-Besse Cycle 6 Core I.cading Diagram ... ... . . . . . 3-3 3-2. Enrichment and Burnup Distribution for Davis-Besse 1, Cycle 6 ..... ...... . ... . . . . . 3-4 3-3. Control Rod Incations for Davis-Besse 1, Cycle 6 .. . . . . . 3-5 3-4. Davis-Besse Cycle 6 BPRA Enrichment and Distribution . . . . . 3-6 4-1. Gray Axial Power Shaping Rod .. ... .... ... . . . . . 4-6 5-1. IOC (4 EFPD), Cycle 6 Two-Dimensional Relative Ptwer Distribution - Full Power, Equilibrium Xenon, All Rods Out, APSRs Inserted . . . ...... .... ... . . . . . 5-5 8-1. Reactor Core Safety Limit (Tech. Spec. Figure 2.1-1) . . . . . 8-2 8-2. Reactor Core Safety Limit (Tech. Spec. Figure 2.1-2) . . . . . 8-3 f 8-3. Trip Setpoint for Flux - AFlux/ Flow (Tech. Spec. Figure 2.2-1) .... ... . ...... . . . . 8-5 Babcock & Wilcox a McDermott company

List of Fiqures (Cont'd)

Page 8-4. Pressure / Temperature Limits at Maxirum Allowable Power for Minimum INER (Ibch. Spec. L cs Figure 2.1) . . . . . 8-10 8-5. Minimum Boric Acid Tank Available Volume as Function of

{ Stored Boric Acid Concentration - Davis-Besse 1 (Tech. Spec. Figure 3.1-1) . . . . . . .... ..... . . . 8-14 8-6. Regulating Group Position Limits, O to 325110 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 6 (Ibch.

Spec. Figure 3.1-2a) . . . . . . . . . ............ 8-16 8-7. Regulating Group Position Limits After 325i10 EFPD, Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-2b) . . . . . . . . . . . . . . 8-17 8-8. Regulating Group Position Limits, O to 325i10 EFPD, Three RC Pumps, - Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-3a) . . . . . . .......... . . 8-18 8-9. Regulating Group Position Limits After 325i10 EFPD, Three RC Pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-3b) . . ... ......... 8-19 8-10. Control Rod Core I.ocations and Group Assignments -

Davis-Besse 1, Cycle 6 (7bch. Spec. Figure 3.1-4) ..... . . 8-20 8-11. APSR Position LiBits, O to 325110 EFPD, Four RC Purps - Davis-12:sse 1, Cycle 6 (Ibch. Spec. Figure 3.1-Sa) . . . . . . ............ 8-22 8-12. APSR Position Limits After 325110 EFPD, Three or Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 6 (TG 5. Spec. Figure 3.1-Sb) . . . . . . . . . . . . . . 8-23 8-13. APSR Position Limits, O to 325i10 EFPD, 7hree RC Pumps - Davis-Besse 1, Cycle 6 (7bch.

Spec. Figure 3.1-5c) . . . . . . . . . ............ 8-24 8-14. AXIAL ICWER IMBMANCE Limits, Four RC Punps -

Davis-Besse 1, Cycle 6 (7bch. Spec. Figure 3.2-1) . . . . . . . 8-26 8-15. AXIAL ICWER IMBMANCE Limits, Three RC Punps -

Divis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.2-2) . . . . . . . 8-27 Babcock &WHcom a McDermott company

1. INIRODUCI' ION AND SUlt9RY

'Ihis report just Zies operation of Davjs-Besse Nuclear Power Station Unit 1 at the rated core power of 2772 }Ht for cycle 6. 'Ihe required c.nalyses are included as outlined in the Nuclear Regulatory Comnission (NRC) document, "Guidance for Proposed License Amendments Relating to Refuelirs," June 1975.

'Ihis report utilizes the analytical techniques and design bases that have been subnitted to the NRC and approved by that agemy.

Cycle 6 reactor and fuel parameters related to power capability are summarized in this report ard ccrapared to cycle 5. All accidents analyzed in the Davis-Besse Final Safety Analysis Reportl (FSAR) or the Updated Safety Analysis Report 2 (USAR), as applicable, nave been reviewed for cycle 6 operation, ard in cases where cycle 6 characteristics were conservative capared to cycle 1, no new analyses were perfomed.

'Ibe 'Ibchnical Specifications nave been reviewed and modified where required for cycle 6 operation. Based on the analyses performed, taking into account the emergency core cooling system (ECCS) Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Divis-Besse Unit t

1, cycle 6 can ba operated safely at its licensed core power level of 2772 tMt.

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2. OPERATING HISIORY me reference cycle for the nuclear and themal-hydraulic analyses of Davis-Besse Unit 1 is the recently complettA cycle 5, which achieved criticality on January 15, 1985. Power escalation began on Januarf 19, 1985 aM reached 93% of full power (2772 MWt) on January 29, 1985. Nomal operation continued until shutdown on June 9, 1985 after a transient. After 18 nonths, in which numerous operations and plant design enhancements were implemented, cycle 5 again achieved criticality on December 22, 1986. On March 20, 1987 full power was reached.

During cycle 5 operation, no operating ancanalies occurred that would ad-versely affect fuel perfomance during cycle 6. We cycle 5 ncaninal design and cycle 6 desired licensed lengths are 400 aM 415 effective full power days (EFPD), respectively. Cycle 6 was analyzed out to 405 EFPD and the validity of the Technical Specifications has been verified out to 415 EFPD.

Berefore, the resulting Technical Specification Limiting CoMitions for Operation ac) r -date cycle 6 operation through 415 EFPD. Se APSRs were pilled at 325 EFPD to increase the lifetime of cycle 5. We APSR pull coupled with a power coastdown resulted in a cycle 5 length of approximately 395 EFPD. Se cycle 6 design also includes an APSR pull and power coastdown.

Se cycle 6 design minimizes the number of fuel assemblies that are cross core shuffled to reduce the potential for quadrant power tilt amplification.

Se cycle 6 shuffle pattern is discussed in section 3.

1 2-1 Bat > cock & Wilcom a MCoffmott Company

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3. GDfERAL DESCRIPTICH ,

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he . Davis-Besse Unit i reactor core is described in detail in chapter 4 of the'USAR2 for.the unit. We cycle 6 core consists of 177 fuel assemblies (FAs), each of khich is a 15x15 array containing 208 fuel rods,16 cer] trol rod guide tubes, and one incere instrument guide tube. All FAs in batches

,6, 7, and 8 have a constant ncminal fuel loading of 468.25 kg of uranium.

We one batch 1A assembly has a fuel Acading of 472.24 kg of uranium. 'v fuel consists of dished-end cylindrical pellets of uranium dioxide cla. a

, cold-worked Zircaloy 4. We undensified nominal active fuel lengths, I percent theoretical densities, fuel ard fuel red dimensjons, and other re-j lated fuel parameters ray be found in Table 4-1 of this report.

> Figure 3-1 is the cnre loading diagram for Davis-Besse Unit 1, cycle 6. One

batch 1E assembly, 16 batch 4B assemblies, 8 batch SA assemblies, ard 40 f batch 5B assemblies will be discharged at the end of cycle 5. Se fuel I asse661f es in batches 6 and 7 will be shuffled to their cycle 6 locations, with batch 7 on the, core periphery. One batch 1A assembly, presently in the sper;t fuel pcol, with an initial uranium-235 enrichment of 1.98 wt % will be reinserted in cycle 6 as the center FA. Batches 6 and 7 have initial enrichments hf 2.99 and 3.19 wt %, respectively. We feed batch, ccrisisting of 64 batc: t / 8 assemblies with uranium enrichment of 3.13 wt %, will be t

i.6 in the core interior in a tynnetric checkerboard pattern with the liatch 6 FAs. Figure 3-2 is a quarter-core map showing each assembly's i burnup at the reginning of cycle (BOC) 6 and its initial enrichment.

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Cycle 6 is cgnated in a feed-and-bleed mode. %e core reactivity is con-trolled by 53 full-length Ag-In-Cd control rod assemblies (CRAs), 64 burnable poison red assemblies (BPRAs), and soluble boron. In addition to the full-length control rods, eight Inconel-600 axial power shaping rods (gray APSRs) are provided for additional control of the axial power distribution. We cycle 6 locations of the control Itxis and the group designations are indicated in Figure 3-3. %e core locations of the 61 control rods in cycles 5 end 6 are the same, however, the rod group 3-1 Bat > cock &WHcom a McDermott company a

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designations differ between cycles. 'Ihe changes in the red group designations are to (1) decrease the worth of group 7 to be campatible with the gray APSR irbalance control, and (2) increase the wrth of group 4 to facilitate control during physics testing. 'Ihe cycle 6 locations and en-richnents of the BPPAs are shchin in Figure 3-4. ,

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Figure 3-1. Davis-Besse Cycle 6 Core Loading Qiagram 1 N N I I

7 7 7 7 7 b A K4 K2 M6 K14 K12 7 7 7 8 7 8 7 7 7 8 M4 L3 N3 F 08 F N13 L13 M12 7 8 7 8 6 8 6 8 7 8 7 C M10 F K6 F A7 F A9 F K10 F LS

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7 8 6 8 6 8 6 8 6 8 6 8 7 D C11 F R10 F 84 F B11 F B12 F L1 F C5 7 7 8 6 8 6 8 6 8 6 8 7 7 E

C10 F9 F B5 F B6 F 810 F E14 F F7 C6 7 7 8 6 8 6 8 6 8 6 8 6 8 7 7 02 F H1 F A10 F A8 F C14 i C4 07 F 09 C12 F 7 8 6 8 6 8 6 7 6 8 6 8 6 8 7 G 89 F G1 F F2 F C3 Fa C13 F F14 F G15 F B7 7 7 8 6 8 6 7 l^ 7 6 8 6 8 7 7 ~

w.= a til H13 F (2 F F1 N11 F10 MS L15 F M14 F H3 FS cy l 7 8 6 8 6 8 6 7 6 8 6 8 6 8 7 K P9 F K1 F L2 F 03 E6 013 F L14 F K15 F P7 7 7 8 6 8 6 8 6 8 6 8 6 8 7 7

1. N9 012 F N2 F R8 F R6 F HIS F N14 F 04 N7 7 7 8 6 8 6 8 6 8 6 8 7 7 M 010 L9 F M2 F P6 F P10 F P11 F L7 06 y 7 8 6 8 6 8 6 8 6 8 6 8 7 till F F15 F P4 F P5 F P12 F A6 F N5 7 8 7 8 6 8 6 8 7 8 7 0 F11 F G6 F R7 F R9 F G10 F [6 7 7 7 8 7 8 7 7 7 p

E4 F3 03 F C8 F C13 F13 (12 7 7 7 7 7 R G4 G2 E10 G14 G12 t

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8 9 10 11 12 13 14 15 1 2 3 4 5 6 7 i Batch

- Cycle 5 Location (F = Fresh Assembly)

Cy 1 = reinserted from cycle 1 i

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Figure 3-2. Enrichment and Burnup Distribution for Davis-Besse 1, Cycle 6 8 9 10 11 12 13 14 15 1.98 3.19 2.99 3.13 2.99 3.13 3.19 3.19

" 17213 17475 12644 17676 16035 0 17336 0 1

3.19 2.99 3.13 2.99 3.13 2.99 3,13 3.19 K

17673 15952 0 20075 0 17166 0 15293 2.99 3.13 2.99 3.13 2.99 3.13 3.19 3.19 L

16085 0 17563 0 17841 0 13622 17148 3.13 2.99 3.13 2.99 3.13 3.19 3.19 M

0 20079 0 17337 0 17230 16697 2.99 3.13 2.99 3.13 2.99 3,13 3.19 N

17336 0 17845 0 16090 0 16972 3.13 2.99 3.13 3.19 3.13 3.19 0

0 17165 0 17280 0 17425 3.19 3.13 3.19 3.19 3.19 P

17206 0 13632 16691 16951 3.19 3.19 3.19 R

17475 15288 17135 x.xx Initial Enrichment xxxxx BOC Burnup, mwd /mtU 3-4 Babcock & Wilcox a McDermott company

Figure 3-3. Control Rod Locations for Davis-Besse 1, Cycle 6

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8 6 3 6 8 4 L 4 2 5 5 2 M t 7 8 7 8 7 N l 2 5 5 2 0 l 4 6 4 P l l R I i I'

8 9 10 11 12 13 14 15 1 2 3 4 5 6 7 X Group Number Grouc No. of Rods Function 1 4 Safety 2 8 Safety 3 4 Safety 4 9 Safety 5 12 Control 6 8 Control 7 8 Control 8 8 APSRs Total 61 3-5 Babcock & Wilcox J McDermott company f

Figure 3-4. Davis-Besse Cycle 6 BPRA Enrichment and Distribution 9 10 11 12 13 14 15 8

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N 1.4 1.1 0.2 0 1.4 0.5 0.2 P 0.2 R

x.x BPRA Concentration, wt% 4B C in Al23 0' 3-6 Babcock & Wilcox a McDermott company

4. FUEL SYSTD4 DESIGN

4.1. Fuel Assenbly Mechanical Desictn he fuel assembly types (FAs) and pertinent fuel parameters for Davis-Besse 1, cycle 6 are listed in Table 4-1. Se Batch 7 and 8 FAs are the Mk-B5 design, while the other batches are the Mk-B4 design. We Mk-B5 FAs are l

identical in concept to the Mk-B4 with only a change to the upper end fitting design which eliminates retainers for burnable poison rod assembly (BPRA) holddown.

( Sixty-four BPPAs are used with the Batch 8 FAs. Also, eight gray APSRs and 53 CPAs are used in cycle 6.

I Fuel Rod and Gray APSR Desiczn

(. 4.2.

he fuel rod and gray APSR design and mechanical evaluation are dia'tM

\ below.

4.2.1. Claddina Collapse A. Fuel Rod he power histories were reviewed for each fuel assembly in cycle 6. Se mosi: limiting power history for each of the four batches of fuel was found.

me most limiting assembly is the one with the highest burrr.Ip. Rese fcur power histories were ccupared to a generic analysis or to a previous creep collapse analysis. Se generic creep ovalization analysis is based on reference 3 and is applicable to the batch 1A, 6, 7, and 8 designs.

me creep collapse analyses predict collapse times longer than 35000 EFEH.

Se longest incore exposure time for cycle 6 is 25800 EFPH for batch 6 (Table 4-1) .

B. Gray APSR me gray APSRs used in cycle 6 are designed for improved creep 13fe.

Cladding thickness arxl rod ovality, which are the prinary factors controlling the time until creep collapse, are improved to extend the life 4-1 Babcock &Wilcox a McDermott company

of the gray APSR. Se minimm design clackling thickness of the Mark B black APSR is 18 mils, while that of the Mk-B gray APSR is 24 mils. Additionally,-

the gap width between the end plug and the Inconel-600 absorber material is reduced. Finally, the gap area ovality is controlled to tighter tolerances.

S e gray APSR is shown in Figure 4-1. ,

4.2.2. Claddincs Stress A. Fuel Rod l

A conservative fuel rod stress analysis envelopes the Davis-Besse Unit 1 cycle 6 stress values. Se methods used for the analysis of cycle 6 have been used in the previous cycles.

B. Gray APSR Re gray APSR design has demonstrated the ability to meet specified mechanical design requirements. Se APSR cladding stress analysis includes pressure, temperature ard ovality effects. Se gray APSR has sufficient cladding and weld stress margins.

4.2.3. Claddincs Strain A. Riel Rod m e fuel design criteria specify a limit of 1.0% on cladding plastic tensile circumferential strain. Se pellet design ensures that plastic cladding strain is less than 1% at design local pellet burnup and heat generation rates. Se design values are higher than the worst-case values the Davis-Besse Unit 1, cycle 6 fuel is expected to see. Se strain analysis is based on the upper tolerance values for the fuel pellet diameter and density, and the lower tolerance for the clading inside diameter (ID) .

B. Gray APSR l

he gray APSR strain analysis incitxles thermal and irradiation swellit.]

effects. Se results of this analysis show that no cladding strain is irriuced due to thermal expansion or irradiation swelling of the Inconel absorber.

4.3. termal Desicin All fuel in the cycle 6 core is themally similar. Se fresh batch 8 fuel inserted for cycle 6 operation introduces no significant differences in fuel 4-2 Babcock &Wilcon a McDermott company

eap a thermal performance relative to the fuel remaining in the core. 'Ihe thennal analyses for all fuel were performed with the TACO 2 code 4. Ncaninal urdensified input paremeters used in this methodology are provided in Table 4-1. Densification effects are accounted for in the TACD2 code densification model.

Linear heat rate (IER) to fuel melt capability for all fuel was determined with the TACD2 fuel pin perfonunce code. 'Ihe analyses perfonned for cycle 6 demonstrate that 20.5 kW/ft is a conservative limit to preclude centerline fuel melt (CFM) for all fuel batches.

'Ihe maximum fuel rod burnup at EDC 6 is predicted to be less than 38300 t

FMd/mtU. Fuel rod internal pressure has been evaluated with TACO 2 for the highest burnup fuel rod and is predicted to be less than the reactor coolant system pressure of 2200 psia at the core outlet.

I 4.4. Material Corpatibility The compatibility of all possible fuel-cladding-coolant-assembly interactions for batch 8 FAs is identical to that of present fuel.

4.5. Ooeratirn Experience I

Babcock & Wilcox operating experience with the Mark B 15x15 fuel assembly I has verified the adequacy of its design. As of October 31, 1987, the following experience has been accumulated for eight B&W 177 fuel asserbly plants using the Mark B fuel asse:rbly:

Cumulative Current Max FA burnuo, FMd/mtU(a) net electric Eeactor Cycle incore Discharued output.Mhh(b)

Oconee 1 10 45,908 50,598 66,183,044 Oconee 2 9 40,580 41,592 60.968,626 Oconee 3 10 33,290 39,701 60,843,663

'Ihree Mile Islard 6 26,090 33,444 29,469,976 Arlartsas Nuclear 8 51,540 47,560 51,626,035 One, Unit 1 Rarrho Seco 7 26,242 38,268 39,045,954 4-3 Babcock &Wilcox a McDermott company

Cumulative Current Max FA burnup, Mtl/mtU(a) not electric Reactor Cycle Incore Discharned output Msh(b)

Crystal River 3 6 35,350 31,420 38,512,798 Davis-Besse 5 36,960 32,790 25,236,663 (a)As of October 31, 1987.

(b) As of Decerber 31, 1986.

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m Table 4-1. Fuel DesicIn Parameters Batch 1A 6 7 8 FA type Mk-B4 Mk-B4 Mk-B5 Mk-B5 Number of assemblies 1 48 64 64 Fuel rod OD, in. 0.430 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 0.377 Flexible spacer type Spring Spring Spring Spring Rigid spacer type Zirc-4 Zirc-4 Zirc-4 Zirc-4 Undensified active fuel length, in. 143.5 143.20 143.20 143.20 l Fuel pellet (mean)

( dia., in. 0.3675 0.3686 0.3686 0.3686 1

Fuel pellet initial l density, % TD mean 96 95 95 95 l

i Initial fuel enrich-ment, wt % 235U 1.98 2.99 3.19 3.13 Average Burnup 12600 17500 16500 0 BOC, mwd /mtU Exposure Time 18700 25800 19300 9700 EOC, (EFIH)

Cladding Collapse >35000 >35000 >35000 >35000 Time, (EFRI)

Ncxninal Linear 6.13 6.14 6.14 6.14 Heat Rate at 2772 MWt, }S/ft Mininun Linear 22.0 20.5 20.5 20.5 Heat Rate to Melt,

}S/ft 4-5 Babcock &Wilcox a McDermott company

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 ccrupares the core physics parameters for the CJcle 5 and 6 de-signs. We values for cycle 5 were generated using PD2075 ard the values for cycle 6 were calculated with the N00DE code 6. Se calculational-j differences resultirq from the upgrading of the modeling are negligible.

Both PD207 and NOODE have been verified to produce results within the l bounds of the same measurement uncertainties (see section 5.2 for a further discussion). Differences in core physics parameters are to be expected between the cycles due to the charges noted below in section 5.2, such as f the longer cycle 6 length. Figure 5-1 illustrates a representative relative f power distribution for the Boc 6 at full power with equilibrium xenon, all rods out and gray APSRs inserted.

l Se ejected rod worths in Table 5-1 are the maxinum calculated values.

Calculated ejected rod worths and their adherence to criteria are 'wisidered at all times in life and at all power levels in the developnent of the rod position limits presented in section 8. ne adequacy of the shutdown margin t with cycle 6 rod worths is shown in Table 5-2. W e following conservatisms were applied for the shutdown margin calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth.

f 3. A maxinum flux redistribution penalty.

4. A maximum power deficit with mininum boron.

Se maxinum flux redistribution was taken into account to ensure that the effects of operational maneuvering transients were included in the shutdown nargin analysis.

l 5-1 Babcock &Wilcox 4 MCDermott Company

LL chanaes in Nuclear Desicin Core design changes for cycle 6 include, (1) the increase in cycle lifetime to 405 EFPD, (2) an increase in the BPRA concentrations, (3) the variation in the loading pattern between cycles 5 aM 6, (4) the second transition cycle to the EP, loa leakage fuel cycle design, (5) the revised control red groupings, (6) the removal of the regenerative neutron sources which will result in a sourceless startup, and (7) the replacement of the Silver-l Indium-Cadmium APSPs used in all previous cycles with gray APSRs, khich have ]

l a longer absorter section but use a weaker absorber (Inconel-600) . nese I design differenccs can explain the differences in the physics parameters between cycles 5 and 6 as shown in Table 5-1. Calculations with the l standard three-dimensional model verified that changing to the gray APSR I 1

design provides adequate axial power distribution control. As stated in section 5.1, the NOODE code was used to calculate the physics parameters for cycle 6. We NCODE redeling of the twm hamcgenized fuel asserbly is the same as that used in PD207. Mcuever, the analytical expression ;

l NOODE 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 6 illustrates the calculational accuracy attainable with NOODE in ccxuparison to measured results for varicus physics parameters. PD207 results are ccmpared to measured data in reference 7. Rese cognrisons shcra NOODE to be as accurate as PDQ07.

Other calculational models and the methods used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle. No significant operational or procedural changes exist with regard to axial or radial pcuer shape, xenon, or tilt control. We stability and control of the core with APSRs withdrawTi has been analyzed.

We calculated stability index without APSRs is -0.023 h-1, which demon-strates the axial stability of the core.

5-2 Babcock &Wilcox a Mconmott Company

Table 5-1. Davis-Besse Unit 1. Cycle 6 mvsics Parameters Cvele $ Cvele 6 Design cycle length, EFPD ~390 405

. Cycle burnup, mwd /mtU 13,043 13,545 Average core burnup - EOC, mwd /mtU 22,797 24,335 Initial core loading, mtU 82.9 82.9 Critical boron (a) - BOC, No Xe, [pa HZP 1,485 1,451 HFP 1,255 1,285 Critical boron (a) - EDC, Eq. Xe, pga HZP 324 204 HFP 10(b) lo(b)

Control rod worths - HFP, IOC, % ak/k Group 6 1.13 1.18 Group 7 1.42 1.05 l Group 8 0.38 0.21 Control rod worths - HFP, EOC, % ak/k 1.46 1.14 Group 7

! Group 8 la la Max ejected rod worth - HZP, % Ak/k (location) i IOC, Groups 5-8 inserted 0.60 0.36 l

(N-12) (L-10) f EOC, Groups 5-7 inserted 0.55 0.40 I

(N-12) (Ir10) f Max stuck rod worth - HZP, % Ak/k (location)

BOC 0.80 0.66 (N-12) (M-11)

EDC O.76 0.79 (M-11) (M-11)

Power deficit - HZP to HFP, Eq. Xe, % ak/k BOC (4 EFPD) -1.76 -1.71 EOC -2.48 -2.51 Doppler coeff - HFP, 10-3 % Ak/k/0F BOC, No Xe,1285 ppn -1.50 -1.55 EOC, Eq. Xe, O ppn,(b)(c) Group 8 inserted Group 8 withdrawn -1.76 -1.84 Moderator coeff - HFP, 10-2 % Ak/k/OF L')C, No Xe, 1285 . (c) -0.81 -0.59 IOC, Eq. Xe, 0 ppn -2.86 -2.84 Boron worth - IEP . % ak/k BOC (1285 p I 123 124 EDC (O Ipn) W 106 107

Xenon worth - HFP, % ak/k BOC (4 EFPD) 2.63 2.63 IDC (equilibrium) 2.73 2.78 Effective delayed neutron fraction - HFP IOC 0.0063'. 0.00626 IDC 0.00524 0.00518 5-3 Babcock & Wilcom a McDermott (c,inpany

1 Table 5-1 (Continued) I l

(a) Control rod group 8 is inserted at BOC and withdrawn at EOC.

(b) Power coastdown to DOC at 10 ppu.

(c) Cycle 6 values were calculated at 1285 ppm; cycle 5 values were calculated at 1255 ppn. i (d) Cycle 6 values were calculated at 0 ppn; cycle 5 values were l calculated at 10 ppm.

Table 5-2. Shutdown Matnin Calculation for Davis-Besse. Cycle 6 EDC. % ak/k ECC, 325 EFPD 405 EFPD

% ak/k Group 8 in Grouc 8 out Available Rod Worth ibtal rod worth, HZP 7.38 7.55 7.59 Worth reduction due to burnup -0.42 -0.42 -0.42 of poison material Maximum stuck rod, HZP -Q,55 -0.74 -0,79 Net worth 6.30 6.39 6.38 Less 10% uncertainty -0,6) -0.64 -0,64 Total available worth 5.67 5.75 5.74 l Reauired Rod kbrth l

Power deficit, HFP to HZP 1.71 2.21 2.51 l Max allowable inserted rod Worth 0.26 0.38 0.40 Flux redistribution 0.27 0.60 0.63 Tctal required worth 2.24 3.19 3.54 Shutdown MaralD Total available minus total 3.43 2.56 2.20 required rod worth Note: Required shutdcun rargin is 1.00% Ak/k.

5-4 ggg a McDermott company

Figure 5-1. B0C (4 EFPD), Cycle 6 Two-Dimensional Relative Power Distribution - Full P9wer, Equilibrium Xenon, All Rods Out, APSRs Inserted (a) 8 9 10 11 12 13 14 15 H .887 1.111 1.140 1.263 1.158 1.247 1.035 .547 x 1.111 1.101 1.258 1.124 1.274 1.153 1.141 .545 8

L 1.138 1.256 1.160 1.282 1.141 1.286 .913 .394 l

l M 1.260 1.121 1.275 1.172 1.272 1.046 .628

(

8 t

( N 1.155 1.270 1.136 1.269 1.054 .933 .405 0 1.244 1.150 1.282 1.044 .93~ .497 P 1.033 1.139 .910 .627 .405 R .546 .544 .393 x Inserted Rod Group Number x.xx Relative Power Density (al liaiculated results from two-dimensional pin-3y-pin PDQ07, 5-5 Babcock & Wilcox a utoermoit company

6. 'IHERMAIeHYERAULIC DESIGN 2e themal-hydraulic design evaluation supporting cycle 6 operation used the methods and models described in references 2 ard 8. S e cycle 6 design analysis is the first application of crossflow methodology for the Davis-Besse station.

We use of crossf?w codes which can predict the flow redistribution effects in l.

an open lattice reactor core, provide significant departure frm nucleate boiling f ratio (ERE) margin improvements relative to the traditional closed-channel codes.

l 2e LYNX 1,9 LYNX 2 10 and LYNXTll crossflow codes were used in the cycle 6 design analyses. he LYNX 1 and LYNX 2 codes were used primarily to benchmark the LYNXT rodels.

k Cycle 5 ard 6 thermal-hydraulic design condit'ons are listed in Table 6-1. he reactor coolant flow, bypass flow and design axial flux shape were revised for the cycle 6 analysis. S e cycle 6 reactor coolant flow requirement was reduced approximately 2 percent relative to the cycle 5 original design value. 21s reduction provides additional operating flexibility. S e original design cycle 5 bypass flow of 10.7 percent was determined with no BPRAs. We cycle 6 bypass flow of 8.6 percent is based on 52 BPRAs and prtduces a conservative bypass flow with respect to the actual care configuration because the number of BPRAs actually in the core is 64. We design axial peak has been increased from 1.50 to 1.65 to provide additional margin for maneuvering analyses and resulting control rod insertion limits. mis change was ace--- Mted by the introduction of the crossflow methodology.

Crossflow methodology was used to reanalyze the pressure-temperature limits, the one punp ooastdown and corresponding flux / flow limit and the locPed rotor transient. All analyses showed a significant increase in DM margin relative to the cycle 5 closed-channel analysis. Se RPS pressure-tenperature trip setpoint has been recalculated based on pressure-tenperature limits cocputed with LYNXT.

me flux / flow limit detemined with LYNXT irdicates that the flux / flow setpoint may be raised frm 1.07 to 1.08. Analysis cf the locked rotor transient with 6-1 g,g,,,,g g ggg,,,

I a Mcoermott compar1y

LYNXP shows that the mininnn INBR during the transient is greater than the design limit of 1.3012, l

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Table 6-1. 'Ihemal-Hydraulic Desian Conditions Cycle 5 - Cycle 6

,% actor Coolant System Ratad 'Ihermal Power I.evel, MRt 2772 2772 N minal Core Exit Pressure, psia 2200 2200 Mininum Core Exit Pressure, psia 2135 2135 Reactor Coolant Flow, gpn 387,200 380,000 Nminal Vessel Inlet Coolant Tenperature,(a) oF 557.7 557.4 Nminal Vessel Outlet Coolant '1hperature,(a) oF 606.3 606.6 Bypass Flow, % design 10.7 8.6 Power Distribution Design Radial x In::al Peaking Factor 1.71 1.71 Design Axial Flux Shape 1.50 cosine 1.65 with tails chopped cosine Hot Channel Factors Enthalpy Rise 1.011 1.011 Heat Flux 1.014 1.014 Flow Area 0.98 0.98 Cold Fuel Stack Height,(b) inches 143.2 143.2 Average Heat Flux, 10 5Btu /hr-ft 2 1.89(c) 1.86 Maxinum Heat Flux,105 @ -ft2 4.85(c) 5.25 g DiBR Parameters Critical Heat Flux Correlation BAW-2 BAW-2 Minhum DiBR, (at 102% FP) -

2.07 Mininum DiBR, (at 112 %FP) 1.79 1.79 (a)1oo gyp (b) Smallest value of all assemblies (c) Based on densified stack height W

n:n m.m wmpany

,. .. _ . . , _ . , _ _ _ _ . _ ____m. _

7. ACCIDDR AND 'IRAIGIDE ANALYSIS 7.1. General Safety Analysig Each USAR accident analysis has been examined with respect to changes in the cycle 6 parameters to detemine the effects of the cycle 6 reload and to ensure that themal perform nce durirg hypothetical transients is not degraded. We effects of fuel densification on the USAR accident results have been evaluated and are reported in reference 13.

We radiological dose consequences of the FSAR Chapter 15 accidents have been evaluatM using conservative radionuclide source tems that bound the cycle specific source tem for DB-1 cycle 6. %e dose calculations were performed consistent with the assumptions described in the DB-1 FSAR but

! used the more conservative source terms (which bound future reload cycles) .

1 l

%e results of the dose evaluations showxl that offsite radiological doses for each accident were below the respective acceptance criteria values in I

the current NRC Standard Review Plan (!URD3-0800) .

7.2. Accident Evalua_t;LQD 2e key parameters that have the greatest effect on detemining the outcome of a transient can typically be classified in three major areas: (1) core themal, (2) themal-hydraulic, aM (3) kinetics parameters includirq the reactivity feedback coefficients and control rod worths.

Fuel thermal analysis parameters frcxn each . batch in cycle 6 are given in Table 4-1. We cycle 5 and cycle 6 themal-hydraulic maximum design conditions are presented in Table 6-1. A ccruparison of the key kinetics parameters from the USAR and cycle 6 is provided in Table 7-1.

A generic loss-of-coolant accident (II)CA) analysis for B&W 177-FA raised-loop nuclear steam systems (NSSs) has been performed usirg ,the Final Acceptance Criteria ECCS Evaluation Model.14 2e ccrnbination of average t fuel temperature as a function of linear heat rate (IER) aM the lifetime pin pressure data used in the IOCA limits analysis is conservative ccr: pared J

7-1 Babcock &Wilcox f a McDermott company

+ - - - - _ - - - -

to those calculated for this reload. R us, the analysis and the IOCA limits reported in reference 14 provide conservative results for the operation of Davis-Besse Unit 1, cycle 6 fuel. A tabulation showing the bounding values for allowable IDCA peak LHRs for Divis-Besse Unit 1, cycle 6 fuel is provided in Table 7-2.

. It is concluded by the examination of cycle 6 core themal, thermal-hydraulic, and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ability to safely operate the Davis-Besse Unit 1 plant during cycle 6. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 6 is considered to be bounded by previously accepted analyses. Se initial corviitions of the transients in cycle 6 are bounded by the USAR and/or the fuel densification report.

l l

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a McDermott company l

Dble 7-1. Coroarison of Kev Parameters for Accident Analysis

- s FSAR and densif'n report Cycle 6 Parameter value value BOL(a) Doppler coeff,10-3, % ak/k/0F -1.28 -1.55 EDL(b) Doppler coeff,10~3, % Ak/k/0F -1.45 (c) -1.84 BOL moderator coeff,10-2, % ak/k/0F +0.13 -0.59 EOL moderator coeff,10-2, g 3pjpfoF -3.0 -2.84 All rod bank worth (HZP), % ak/k 10.0 7.38 Boron reactivity worth (HFP), ppnt 1% / Ak/k 100 124 Max ejected rod worth (HFP), % ak/k O.65 0.29 Max dropped rod worth (HFP), % ak/k O.65 0.20 Initial boron conc (HFP), ppm 1407 1285 l

(a)BOL denotes beginning of life.

(b)EOL denotes end of life 3 (c) _1,77 x 10-3 % Ak/k/0 F was used for steam line failure analysis.

Table 7-2. Boundina Values for Allowable IDCA Peak Linear Heat Rates Allowable Allowable Core peak u m, peak Um, elevation, first 25 EFFD, balance of cycle, ft kW/ft kWft 2 15.5 16.5 4 16.8 17.2 6 17.0 18.4 8 17.5 17.5 10 17.0 17.0 7-3 gg a McDermott company

8. PROIOSED }ODIFICATIONS TO TECHTICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 6 operation to account for changes in power peaking and control rod worths. The effects of IUREG-0630 have been ' incorporated into the esperating limits. Figures 8-1 through 8-13 are revisions to the previous cycle Technical Specifications.

The setpoints shown in Table 8-1 may be changed by pending license amendments, and such. changes shall have no effect on the information shown in this report. Based on these Technical Specifications the final acceptance criteria ECCS limits will not be exceeded and the thermal design criteria will not be violated.

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Figure 8-1. Reactor Core Safety Limit (Tech. Spec. Figure 2.1-1) 2500 2400 -

RC High Pressure Trip (618,2300) 2300 RC High

___ Temperature Trip 2200 OPERATION (618,2124.6) >

(633.4, a 2100 -

37 2129.8)

m. RC Pressure 2000 (606.79,1983.4) Temp Trip RC Low Pressure Trip Safety Limit (621.4,1929.8) 1900 1800 - l (608.2.1729.8) ,

1700 t a a i a e 0 590 600 610 620 630 640 650 Reactor Outlet Temperature, F l

2-2 Babcock & Wilcom 8-2 a McDermoit company

Figure 8-2. ReGetor Core Safety Limit (Tech. Spec. Figure 2.1-2)

% RATED THERMAL POWER ,

. .120

(-44.0,112.0) 4 PUMP LIMIT (33.0,112.0)

(-49.0,100.0) - .100

(-44.0,90.0) (33.0,90.0) 3 PUMP LIMIT (47.1,87.2)

" "O

(-49.0,78.0 ) (

)(47.1,65.2)

- 60 UNACCEPTABLE UNACCEPTABLE OPERATION ACCEPTABLE OPERATION OPERATION FOR SPECIFIED RC PUMP j COMBINATION

- -40 l

l

. 20 l

l l i e i i i 1

-60 -40 -20 0 20 40 60 AXIAL POWER IMBALANCE, %

Required Measured Flow to Ensure

Pumps Operating Reactor Ccolant Flow, gom Compliance, gpm 4 380,000 389,500 3 283,860 290,957 2 8-3 ' * " "' * "Y l-

Table 8-1 Reactor Protection System Instrumentation Trio Setpoints (Tedi. Spec. Table 2.2-1)

Functional unit Trio setpoint Allovable values

1. Manual reactor trip Not applicable. Not applicable.

2. liigh flux $104.94% of RATED 111ERMAL $104.94% of RATED TIERMAL POWER with four punps operating 10WER with four punps operatiry#

I

$80.6% of RATED 111ERMAL 10WER $80.6% of RATED 111ERMAL l with three purrps operating ICWER with three punps operatingi

3. RC high temperaturn $6180F $6180F#

1

4. Flux - Aflux/ flow (1) Four pturp trip setpoint not to Four punp allowable value not exceed the limit line of to exceed the limit line of TY

^* Figure 2.2-1. For three punp Figure 2.2-1f. For three punp i

operation, see Figure 2.2-1. operation, see Figure 2.2-1.

5. RC low pressure (l) 21983.4 psig 21983.4 psig* 21983.4 psig** I

6. RC high pressure $2300 psig $2300.0 psig* $2300.0 psig** 1 1

7. RC pressure-tenperature(l) 2(12.60 Tout F - 5662.2) psig 2(12.60 Tout F - 5662.2) psigt j l

8. Illgh flux /nurnber of RC punps on(1) $55.1% of RATED TIIERMAL POWER $55.1% of RATED TIIERMAL 10WER with one pturp operating in eadi loop with one punp operating in eadt W loopf c> m 1 r7 1 6a 2

50.0% of RATED 711ERMAL POWER with $0.0% of RATED 711ERMAL IMER I t w w g erating in one loop and with tw w operating in one

!*S N no pumps operating in the other loop loop and no pungs operating in i . the other loopf 3b

$ R" $0.0 of RATTD TilERMAL POWER with no 50.0% of RATED 711ERMAL 10WER

'$ plugs operating or only one punp with no punps operating or only 0 Prating one gap operatingi

9. Containment pressure high $4 psig $4 psig#

Figure 8-3. Trip Setpoint for Flux --a Flux / Flow (Tech. Spec. Figure 2.2-1)

. RATED THERMAL POWER UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION

- -120 Curve shows trip

(-17.0,108.0) (17.0,108.0) setpoint for an approximately M1 =+1.00 -

M =-2.27 25% flow reduc-

~

l00 2 tion for three 4 PUMP l pump operation

(-30.6,94.4) LIMIT (283,860 gpm).

l 8

1 I The actual set-80 (17.0,80.6) point will be

(-17.0,80.6) calculated by the i (30.6,77.1) Reactor Protection System and will be l p P

(- 30.6,67.0) l l directly propor-

, EXAMPLE tional to the l l

- 60 l l actual flow with

" P"*E

IACCEPTABLE OPERATION FOR .

(30.6,49.7) lSPECIFIED RC PUMP '

l l COMBINATION - -40 I I

I l l l l l l 1 I

l - -20 ;

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, , , i ,i I, i , , _ .

-80 -60 -4L -20 0 20 40 60 80 AXIAL POWER IMBALANCE, '.

2-7 8-5 Babcock & Wilcox a McDermott company

/-

/

2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACIOR CORE 2e restrictions of this safety limit prevent overheating of the fuel cladding and,possible cladding perforatien which wculd 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 nucleate boiling regime where the heat trarsfer coefficient is large and the cladding surface terperature is slightly above the coolant saturation terperature.

Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding te::peratures because of the onset of departure from nucleate boiling (DB) and the resultant sharp reduction in heat transfer coefficient. DG is not a directly measurable parameter during operation and therefore 'IHERMAL F0WER and Reactor Coolant Te:nperature and Pressure have been related to DG through the B&W-2 EtB correlation. ne EtB correlation has been developed to predict the DG flux and the location

of us for axially unifom and non-unifom heat flux distributions. ne local DG heat flux ratio, EtGR, defined as the ratio of the heat flux that would cause EtG at a particular ccre location to the local heat flux, is irdicative of the margin to EtG.

Se minimum value of the EtGR during steady state operation, normal operational transients, and anticipated trans.4snts is limited to,l.'30. 'Ihis value corresponds to a 95 percent probability at a 95 percent confidence level that EtB will not occur and is chosen as an appropriate margin to ENB for all operating conditions.

Re curve presented in Figure 2.1-1 represents the conditions at khich a minimum EtB of 1.30 is predicted for the maximum possible themal pccer of' 112% khen the reactor coolant flow is 380,000 GRi, khich is approxinately coolant pumps. (?he 108% ofrequired minimum designmeasured ficw rateflow forisfour operating 389,500 GR4.) reactycurve Ris is based on the following hot channel-factors with potential fuel densification and fuel red ,

bowing effects:

Fg = 2.83; /AH = 1.71; [g=1.65 2 e design limit power peaking factors are the most restrictive calculated at full power for the range frcm all control rods fully withdrawn to minimum allowable control rcd withdrawal, and fom the core EtGR design . basis.

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SAFETI LIMITS BASES he curves of Figure 2.1-2 are based on the more restrictive of two thermal limits ard account for the effects of potential fuel densification and potential fuel rod bcu.

1. The 1.30 DiBR limit produ d by a nuclear power peaking factor of Fg =

2.83 or the ccobination of the radial peak, axial peak, and position of l the axial peak that yields no less than a 1.30 DiBR.

2. The cxrbination of radial and axial peak that causes central fuel melting at the hot spot. The limits are 22.0 W/ft for batch 1F ard 20.5 W/ft for batches 6, 7, and 8.

Power peaking is not a directly observable quantity ard therefore limits have been established on the basis of the ruector power imbalance produced by the power peakirq.

The specified flow rates for the two curves of Figure 2.1-2 correspond to l the analyzed minimum flcw rates with four punps and three purps, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pucp-maximnn thermal power ccabinations shown in BASES Figure 2.1.

The curves of BASES Figures 2.1 represent the conditions at which a mininnn DiBR of 1.30 is predic.ted at the maxinum possible thermal power for the number of reactor coolant pumps in cperation or the local quality at the point of mininum INBR is equal to +22%, whichever condition is more restrictive. These curves include the potential effects of fuel rod bow and l

tuel densification. .

B babcock & WilCOM J MCDermott Cornparty

SAITIY LIMI'IS BASES For the curve of BASES Figure 2.1, a pressure-taperature point above an-1 to the left of the curve would result in a [NBR greater than 1.30 or a local quality at the point of mininann ENBR less than +22% for that particular reactor coolant punp situatien. 'Ihe 1.30 ENBR curve for three punp operation is less restrictive than the four punp culve.

2.1.3 REACIOR COOIN7P SYSIDi PRESSURE

'Ihe restriction of this Safety Limit protects the integrity of the Reactor Coolant System fren overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant fran reaching the containment atmosphere.

'Ibe reactor pressure vessel arri pressurizer are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maximum.ttansient pressure of 110%, 2750 psig, of design pressure. 'Ihe Reactor Ccolant System piping, valves and fittirgs, ere designed to ANSI B 31.7, 1968 Editicri, which permits a maxirum transient pressure of 110%, 2750 psig, of ccmponent design pressure. 'Ihe Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements.

'Ihe entire Reactor Coolant System is hydrotested at 3125 psig,125% of design pressure, to deconstrate integrity prior to initial operaticm.

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l LDUTING SAFEIY SYSTEM SEITINGS '

M RC Hich 'IWoerature

'Ibe RC high tenperature trip $6180 F prevents the reactor outlet tenperature frcrn eyrwviiry the design limits ard acts as a backup trip for all power e.xcursion transients.

Flux - AFlux/ Flow

'Ihe power level trip setpoint produced by the reactor coolant system flow ic based on a flux-to-flow ratio which has been established to am - *te flow decreasing transients frun high power where protection is not provided by the high flux / number of reactor coolant punps on trips.

'Ihe power level trip setpoint produced by the power-to-flow ratio provides both high power level ard low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. 'Ihe power level setpoint produced by the power-to-flow ratio provides overpower ENB protection for all nodes of pump operation. For every flow rate there is a maxinum permissible power level, ard for every power level there is a minimm permissible low flow rate. Examples of typical power level aM low flow rate crxtbinations for the pmp situations of Table 2.2-1 that wculd result in a trip are as follows:

1

1. Trip would occur when four reactor coolant punps are operatirg if power is 108.0% aM reactor coolant flow rate is 100% of full flow rate, or

ficw rate is 92.59% of full flow rate aM power level is 100%.

2. Trip would occur when three reactor coolant punps are operating if, power is 80.68% ard reactor ccolant. flow rate is 74.7% of full flow rate, or flow rate is 69.44% of full flow rate and power is 75%. Note that the l

value of 80.6% in Figure 2.2-1 was truncated frcan the calculated value

of 80.68%.

1 For safety calculations the instrumentation errors for the power level were

, used. Full flow rate in the above two exauples is defined as the flow I

calculated by the heat balance at 100% power. At the time of the l calibration the RCS flow will be greater than or equal to the value in Table 3.2-2.

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Figure 8-4. Pressure / Temperature Limits at Maximum Allowable Power for Minimum DNBR (Tech. Spec. Bases Figure 2.1) 2300 2200 -

(636.3,2159.8)

(633.4,2129.8) /

2100 - /

/

4 Pump /

.5" r/

E 2000 - ACCEPTABLE

/

. OPERATION c

o-

^

(621.4,1929.8)

/

f (625.7,1959.8) 1900 - / UNACCEPTABLE

/ OPERATION

/

f

= 3 Pump ,

1800 -

j (608.2, !(614.3,1759.8) i 1729.8) 1700 1 I I I 595 605 615 625' 635 645 Reactor Outlet Temperature, F '

Required Measured

)

Flow to ensure Pumps Flow, gDm Power ComDliance, gDm 4 380,000 112% 389,50u 3 283,860 90.5% 290,957 B 2-8 Baf: dock & WilCOX 8-10 a ucoermott company

' REACTIVITY cot 7IROL SYSTDG B3 RATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.8 As a minimum, one of the following borated water sources shall be OPERABE:

a. A boric acid addition system with:

1. A minimum available borated water volume of 600 gallons,

2. Between 7875 and 13,125 ppm of boron, and

3. A minimum solution teq:erature of 1050F.

b. 'Ihe borated water storage tank (BWST) with:

1. A minimum available berated water volume of 3,000 gallons, l

2. A minimum boron concentration of 1800 ppm, and

3. A minimum solution terperature of 350F.

APPLICABILITY: MODE 5 and 6.

ACTION:

With no borated water sources OPERABE, suspend all operations involving 00RE ALTERATION or positive reactivity changes until at least one borated water source is restored to OPERABE status.

SURVEILIANCE REOUIRDE?7rS l

4.1.2.8 'Ihe above required borated water source Gull be der.istrated OPERABE:

l

a. At least once per 7 days by:

1. Verifying the baron concentration of the water,

2. Verifying the available borated water volume of the source, and l 3/4 1-14 8-11 Babcock & Wilcox FMcDermott company

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

I REACTIVITY CWIPOL SYSIDE IDRATED WATER SCURCES - OPERATING LIMITING CONDITION FOR OPERATICH 1

3.1.2.9 Each of the followirg borated water sources shall be OPERABIE:

J

a. The boric acid addition system with:

1

1. A minimum available borated water volume in ao::ordance l l

with Figure 3.1-1,

2. Between 7875 and 13,125 ppn of boren, and

3. A minirum solution tenperature of 105 0F.

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

1. An available borated water volume of between 482,778 ard l 550,000 gallons,

2. Between 1800 and 2200 pga of boron, and

3. A minimum solution tenperature of 350F.

APPIJCABIIlTY: EDES 1, 2, 3 ard 4.

ACTION: a. With the boric acid addition system inoperable, restore the storage system to OPERABIE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY ard borated to a SHUIIDWN MARGIN equivalent to 1% ak/k at 200 0F within the next '6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the boric acid addition system to OPERABIE status within the next 7 days or be in CDID SHUIDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the borated water rtorage tank inoperable, restore the tank to OPERABIE status within cne hour or be in at least Hctr STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in CDID SHUIDOWN within the followirg 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3/4 1-16 Babcock & Wilcox 8-12 ,uco,, moi, comp,ny

REACTIVITY 00?7 TROL SYSTmS SURVEILIJWCE REOUIRDOTTS 4.1.2.9 Each borated water source shall be demonstrated OPERABIE:

a. At least once per 7 days by:

1. Verifying the boron concentration in each water source,

2. Verifying the available boratal water volume of each water l source, and

3. Verifying the boric acid addition system solution tenperature.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by veri ing the Her tenperature when the outside air temperature is < 35 .

l l

~

t 1

3/4 1-17 8-13 Babcock &Wilcox a McDermott company

1 Figure 8-5. Minimum Boric Acid Tank Available Volume as l )

Function of Stored Boric Acid Concentration-Davis-Besse 1 (Tech. Spec. Figure 3.1-1) l 8500 8000

\\

E 7500 -

S 5 7000

\s ACCEPTABLE OPERATION 6500 ,

o 6000 3 N 5500 U UNACCEPTABLE e 5000 OPERATION 2 N 4500

? '

< 4000 _

3500 _

3000 i l I I j

7000 8000 9000 10,000 11,000 12,000 13,000 14,000 Concentration of Boric Acid Solution, ppm B 3/4 1-18 Babcock & Wilcox 8-14 * *D"mott wmpany

REACTIVfIY CENIROL SYSTDE RBITIATING ROD INSERPION LIMPIS LIMITING CONDITICH FOR OPERATION 3.1.3.6 'Ihe regulating red groups shall be limited in physical insertion as shown on Figures 3.1-la, and -2b, ard 3.1-3a, and -3b. A red group overlap l of 25 d5% shall be maintained between sequential withdrawn groups 5, 6, ard 7.

APPLICABILTIY: MXES 1* ard 2*f.

I ACTION j With the regulating rod groups inserted beyard the above insertion limits (in a region other than acceptable operation), or with any group sequence or overlap outside the specified limits, ex pt for surveillance testing pursuant to Specification 4.1.3.1.2, either:

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

b. Reduce 'IHERMAL FOWER to less than or equal to that fraction of RATED

'IHERMAL IVWER which is allcwed by the rod group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

c. Be in ar, least HCTI STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

H213: If in unacmptable ragion, also see Section 3/4.1.1.1.

i i

See Special 'Ibst Exception 3.10.1 and 3.10.2.

l

With kegg 2 1.0.

3/4 1-26 8-15 Babcock & Wilcox 4 MCDermott Comparty

Figure 8-6. Regulating Group Position Limits, 0 to 325 10 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-2a)

(300,102)

(258,102) , ,

100 - Power Level (270,102)

Cutoff = 100%

(270,92)

SHUTDOWN l E MARGIN (250,S0) g 80 -

LIMIT a

b UNACCEPTABLE w

i: OPERATION OPERATION J 60 -

RESTRICTED S

(180,50) o

$ 40 -

e f

(128,28.5)

{

2 20 -

ACCEPTABLE OPERATION 4 (0.0,5.0) 1 I ' '

0 0 100 200 300 Rod Index (% Withdrawn) j GR 5 i i i 0 75 100 l

GR 6 ' I i i 0 25 75 100 i '

GR 7 I 0 25 100 I 3/4 1-28 8-16 Bat > cock & WHcom a McDermott company

Figure 8-7. Regulating Group Position Limits Af ter 325 10 EFPO, Four RC Pumps, APSRS Withdrawn -- Davis-Besse 1, Cycle 6, _(Tech. Spec. Figure 3.1-2b)

(266,102) (300,102) 100 - Power Level '(270,102)

Cutoff = 100%

(270,92)

SHUT 00WN MARGIN LIMIT (250,80) h 80 2

a f

e OPERATION E 60 -

UNACCEPTABLE OPERATION

& (176,50) (180,50)

{t' 40 -

E (136,28.5)

I 2 20 -

ACCEPTABLE OPERATION i (0.0,5.0) , , ,

0 100 200 300 Rod Index (% Withdrawn)

GR 5 i ' '

0 75 100 GR 6 i 0 25 75 100 GR 7 0 25 100 3/4 1-28a 8-17 , g g

Figure 8-8. Regulating Group Position Limits. O to 325 10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-3a) 100 -

(258,77) (300,77)

S 80 -

$ 0 G I SHUTDOWN MARGIN J(270,69.5) y LIMIT

- 60 -

(250,60.5) a W

UNACCEPTABLE OPERATION OPRATION RESTRICTED t 40 - (170,38) ,

y (180,38) 2 E

~

(128,21.8) ACCEPTABLE OPERATION 0 i e i 0 100 200 300 Rod Index (% Withdrawn)

GR 5 ' ' '

0 75 100 GR 6 i i i i 0 25 75 100 GR 7 8 i !

0 25 100 3/4 1-29 Babcock & Wilcom 8-18 a McDermore company l-

1 J

Figure 8- 9. Regulatirig Group Position Limits Af ter 325 10 EFPO, Three RC Pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-3b) 100 -

1 2 80 - (266,77) (300,77) 7 "g SHUTDOWN (270, 7)

c. MARGIN g UNACCEPTABLE LIMIT (270,69.5) u OPERATION 60 -

250,60.5) e U

N OPERATION (176,38) RESTRICTED 40 -

5 (180,38)

! E u

j 20 - (136,21.8)

ACCEPTABLE

{ OPERATION 0

0 100 200 300 Rod Index (% Withdrawn)

GR 5 ' ' '

)

) 0 75 100 GR 6 '

0 25 75 100 GR 7 '

0 25 100 3/4 1-29a Babcock & Wilcox 8-19 a ucoermoit company l

i l

Figure 8-10. Control Rod Core Locations and Group l Assignments -- Davis Besse 1, Cycle 6 l (Tech Spec. Fiqure 3.1-4)

X

/ N N

l A

4 6 4 3

2 5 5 2 C

7 8 7 8 7 D

2 5 5 2 E

4 8 6 3 6 8 4 F

5 1 1 5 G 3 4 3 7 6 Y <

H w- 6 7 3 5 1 1 5 K

4 8 6 3 6 8 4 L l 2 5 5 2 M i i

7 8 7 8 7 N l 2 5 5 2 O l 4 6 4 P l l R I I (

Z 4 5 6 7 8 9 10 11 12 13 14 15 1 2 3 Grouc No. of Rods Function 1 4 Safety 2 8 Safety Safety 3 4 4 9 Safety 5 12 Control l

6 8 Control 7 8 Control i 8 8 APSRs Total 61 3/4 1-31 8-20 Babcock & Wilcex e McDermott company

- _- .~. ,-. ,

REACTIVTIY CONITOL SYSTDS AXIAL POWER SHAPING TOD INSERTIOT IlMPIS LIMITING CDNDITIQi FOR OPERATIOT 3.1.3.9 'Ihe axial power shaping Iod group shall be limited in physical insertion as shown on Figures 3.1-Sa, -5b, and -Sc. l APPLICABIIIIY: ! ODES 1 and 2*.

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

)

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

b. Reduce 'IHERMAL ICWER to less than or equal to that fraction of RATED

'IHERMAL ICWER wh.ich is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or f c. Be in at least ICT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEIIINE REQUIRDE7PS 4.1.3.9 'Ihe position of the axial power shaping red group shall be j determined to be within the insertion limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i except when the axial power shaping rod insertion limit alarm is inoperable, then verify the group to be within the insertion limit at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1

With Keff 2 1.0.

3/4 1-34 8-21 Babcock & Wilcox a McDermott company I

(

Figure 8- 11. APSR Position Limits, O to 325 :10 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-Sa) i RESTRICTED REGION c 3 100 - (0,102) (100,102)

(

80 -

2 W

2 <

a b 60 -

y PERMISSIBLE e OPERATING REGION N

5 (

o 40 -

2 <

8 e

5 <

~

i E

g 20 -

i 1

(

0 i i ' ' ' ' ' ' ' '

0 10 20 30 40 50 60 70 80 90 100 1 APSR Position (% Withdrawn) d 1

{

3/4 1-35 8-22 Bat > cock & W11com A MCDermott company

Figure 8-12. APSR Postion Limits After 325 10 EFPD, Three or Four RC Pumps, APSRs Withdrawn --

Davis-Besse 1 Cycle 6 (Tech. Spec. Figure 3.1-Sb) 100

$ 80 -

E c.

N g 60 -

APSR INSERTION NOT ALLOWED A IN THIS TIME INTERVAL S

E cc

% 40 -

a 5

E E

20 -

I e

i ' ' ' ' ' ' ' '

0 O 10 20 30 40 SO 60 70 80 90 100 APSR Position (% Withdrawn) 3/4 1-36 8-23 Babcock & Wilcox a MCDermott company l

Figure 8-13. APSR Position Limits. O to 325 10 EFPD, -

Three RC Pumps -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.1-Sc) 100 -

l

(

_x 80 .- RESTRICTED REGION W (0,77) (100,77) 2 l a l l

f e

1 y 60 -

8 E

a w

$ 40 -

PERMISS18LE 5 OPERATING REGION O

2

~

l

( b f c- 20 -

(

( 0 80 0 10 20 30 40 50 60 70 90 100 APSR Position (% Withdrawn)

{

l l

3/4 1-37 8-24 Babcock & WIfcom a McDermott Comparty

3/4.2 POWER DIS'IRIHJI' ION LIMITS AXIAL BJWER IMBAIRE LIMITING CONDITIQi FOR OPERATIOi 3.2.1 AXIAL BJWER IMBAING shall be maintained within the limits shown on Figures 3.2-1 and 3.2-2.

) APPLICABIIIIY: MDDE 1 above 40% of RATED THERMAL BJWER.*

\

ACTION With AXIAL ICWER IMBAIRG eWiry the limits specified above, either:

a. Restore the AXIAL BJWER IMBAING to within its limits within 15 minutes, or

b. Within one hour reduce power until imbalance limits are met or to 40% of RATED THERMAL FCWER or less.

SURVEILING RKUIRD4DTIS 4.2.1 The AXIAL RJWER IMBAIANG shall be determined to be within limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above 40% of RATED N RJWER except when the AXIAL POWER IMBAIRG alam is inoperable, then Mculate the AXIAL

{ BJWER IMBAIANG ac least once per hcur.

)

l l

l

See Special 7bst exception 3.10.1.

3/4 2-1 8-25 Babcock & Wilcox J MCDefmott Comparty

i Figure 8-14 AXIAL POWER IMBALANCE Limits Four RC Pumps -- Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.2-1)

{

-- 110

(-20,102) (15,102) f

[ -

100

(-25,92) (15,92)

- 90 2 '

(-28,80), W-- 80 d20,80) 2 (

j - 70 -

g (

! E -.60 8

(-28,50)o 3-- 50 o (20,50) I

-40 <

5 RESTRICTED PERMISSIBLE $ -- 30 s REGION OPERATING S. (

REGION ' -

- 20 2 -

- 10 i i i i ' ' '

e i.

-50 -40 -30 -20 -10 0 10 20 30 40 50 AXIAL POWER IMBALANCE (%)

3/4 2-2 ,

8-26 Babcock & Wilcox a McDermott company

l f

Figure 8-15. AXI AL POWER IMBALANCE Limits, Three RC Pumps --

Davis-Besse 1, Cycle 6 (Tech. Spec. Figure 3.2-2) l

(

-.110 100 90

)

(-15,77)- -

(11.25,77)

(-18.75,69.5) -- 70 t (11.25,69.5) 2

(-21,60.5) 4 -- 60 > (15,60.5) a 5

22 -- 50 e

p

)

(-21.38) o O " 40 e o (15,38) 5 5 f 3-- 30 $

RESTRICTED g y=

REGION g - 20 w b- ggg

{ S G" y-- 10 h e t I f I ail t I a w f f f a 40 -30 220 -10 0 10 20 30 40 50

{ AXIAL POWER IMBALANCE (%)

3/4 Babcock & WHcox 8-27 4 McDermott (ompany

l PJWER DISTRIBUTION LIMTIS l OUATANT ICWER TILT LIMITING CONDITION FOR OPERATION 3.2.4 THE QUATANT PJWER TILT shall not eW the Steady State Limit of Table 3.2-1.

l APPLICABILITY: LODE 1 above 15% of PATED THERMAL BJWER.*

ACTION:

a. With the QUAmANT FOWER TILT determined to eW the Steady State Limit but less than or equal to the Transient Ilmit of Table 3.2-1.

1. Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a) Either reduce the QUANANT BJWER TILT to within its Steady State Limit, or b) Reduce THERMAL FOWER so as not to exceed 7HEINAL PJWER, including power level cutoff, allowable for the reactor coolant punp ocanbination less at least 2% for each 1% of QUANANT PJWER TILT in excess of the Steady State Limit arxl within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reduce the High Flux Trip Setpoint and the Flux-4 Flux-Flow Trip Setpoint at least 2% for each 1% of QUANANT POWER TILT in excess of the Steady State Limit.

2. Verify that the QUANANT FOWER TILT is within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after eWbg the Steady [

State Limit or reduce 7HmMAL F0WER to less than 60% of i THERMAL ICWER allowable for the reactor coolant punp ombination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 5 65.5% of THERMAL IWER allowable for the reactor coolant punp cx2nbination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Identify and correct the cause of the out of limit condition prior to increasing 7HERMAL IOWER; subsequent PJWER OPERATION above 60% of THERMAL BJWER allowable for the reactor coolant punp ombination may proceed provided that the QUAmART POWER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL ICWER.

See Special 7bst Exception 3.10.1.

3/4 2-9 8-28 Babcock & Wilcox a McDermott company

I I

NWER DISTRIBEIQ' LIMITS LIMITING CEtIDITIOT FOR OPERATIOT (Continued)

b. With the QUAIPR E ICWER TILT deteImined to exceed the Transient Limit but less than the Maxinun Limit of Table 3.2-1 l f due to misalignment of either a safety, regulating or axial pcuer shaping rod:

1. Radu THERMAL ECWER at least 2% for each 1% of irdicated

{ QUAEFRE I%'ER TILT in ems of the Steady State Limit within 30 minutes.

2. Verify that the QUAEFR C POWEP TILT is within its Transient Limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after eynaaM ng the Transient Limit or reduce THERMAL IG'ER to less than 60%

of THERMAL FOWER allowable for the reactor coolant punp ccrbination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 5 65.5% of THERMAL FOWER allowable for the reactor coolant pmp ocobination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Identify and correct the cause of the out of limit cordition prior to increasing THERMAL FOWER; c:n W t IMER OPERATIQ1 above 60% of THERMAL POWER allowable for the reactor coolant pmp cmbination may proceed provided that the QUAIPRE POWER TILT is verified within its Steady

( State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL IWER.

I c. With the QUAEPRE I%'ER TILT determined to exceed the Transient Limit but less than the Maxinum Limit of Table 3.2-1 l due to causes other than the micalignment of either a safety, regulating or axial power shaping rod:

1. Reduce THERNAL POWER to less than 60% of THERMAL ICWER allcuable for the reactor coolant punp ccabination within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 5 65.5%

of THERMAL IG'ER allcvable for the reactor coolant pmp cccbination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2. Identify ard correct the cause of the out of limit condition prior to increasing THERMAL EWER; subsequent PCEER OPERATIOi above 60% of THDEAL ICWER allcuable for the reactor coolant pmp cmbination may procecd provided

. that the QUADPRE IWER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERRL POWER.

3/4 2-10 .

8-29 Babcock & Wilcox a McDermott company

ECWER DISIRTHJI' ION LIMITS LIMITING CONDITION FOR OPEPATION (Continued) _

ACI' ION: (Contimled)

d. With the QUATAtC POWER TILT determined to ex ed the Maxinum Limit of Table 3.2-1, reduce 'IHERMAL POWER to 5 15% of PATED l

'IHERMAL ICWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

EE/EITIR1CE RDDUIREMENIS 4.2.4 'Ihe QUATNE ECWER TILT shall be determined to be within the limits at least once every 7 days during operation above 15% of RATED 'IHERMAL POWER except when the QUATANT ECWER TILT alann is inoperable, then the QUAmANT ECWER TILT shall be calculated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . (

)

l ,

l l

\

3/4 2-11 8-30 Babcock & Wilcox a McDermott company l

Table 8-2. Quadrant Power Tilt Limits (Tech. Spec. Table 3.2-1) l Steady state Steady state limit for limit for 7HERMAL THERMAL Tcdnsient Maximum POWER < 50% PJWER > 50% __ljpit lirit QUAERANT POWER TILT as measured by:

Symetrical incore 6.83 4.12 10.03 20.0 detector system Power range channels 4.05 1.96 6.96 20.0 l Mininum incore detector 2.80 1.90 4.40 20.0 system s

)

1 I -

l 1

I i

l l

3/4 2-12 8-31 Bath:ock & Wilcom a McDermott company l

ICWER DISTRIBL7I'IOT IM TJ1B PA N LIMITDC CDt1DITIOT FOR OPERATIOi 3.2.5 The following DiB related parameters shall be maintained within the limits shcun on Table 3.2-2. l

a. Reactor Coolant Hot Isg Tenperature f

b. Reactor Coolant Pressure

c. Reactor Coolant Flcw Rate APPLICABILITY: } ODE 1 ACTIQi:

If parameter a or b above exrwrk its limit, restore the parameter to within f its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THEIBRL ICWER to less than 5% of RATED THERMAL IONER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If parameter e W s its limit, either:

1. Restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

2. Limit THERMAL ICWER at least 2% below RATED THERMAL PCHER for each 1%

parameter c is outside its limit for four punp operation with5 the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or limit THERMAL 10WER at least 2% below 75% of RATED THEMAL 10WER for each 1% parameter c is outside its limit for 3 punp operation within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . .

SURVEILIR1CE RDXTIRDO7IS 4.2.5.1 Each of the parameters of Table 3.2-2 stall be verified to be l r within their limits at least once per 12 hcurs. _ /

i 4.2.5.2 The Reactor Coolant Systs total flow rate shall be detemined to be within its limit by measurement at least once per 18 nonths.

1 1

I 3/4 2-13 1 8-32 Babcock & Wilcox a ucoermore company

Table 8-3. INB Mamin (7bch. Spec. Table 3.2-2) l Required Required Measured Parameters Measured Parameters dm de Four Reactor Three Reactor Cbolant Punps Coolant Punps Parameter Openting Operating l

Reactor Cmlant Hot Iag 5 610 $ 610(1)

'nsperature TaoF Reactor Coolant Pressure, psig(2) 2 2062.7 2 2058.7(1)

Reactor Coolant Flow Rate, gpn(3) 2 389,500 2 290,957 l (1) Applicable to the loop with 2 Reactor Coolant Punps Operating.

(2) Limit not applicable during either a 7HERMAL IVWER ranp increase in

{ excess of 5% of RATED THERMAL 10WER per minute or a 7HERMAL ICWER step increase of greater than 10% of RATED THERMAL POWER.

, (3)These mininn required measured flows include a flow rate un rtainty of 2.5%, and are based on a minin.un of 52 lunped bumable poison red ,

assemblies in place in tre core.

l 3/4 2-14 Babcock & Wilcom 8-33 , uco,, moi,co,rp ny l

Table 8-4.

Reactor Protection Systen W% tion Surveillance Rmdrenents (Tedi. Spec. Table 4.3-1) mannel Modes in Whidi Gannel dannel Ebnctional Surveillance Functional Unit meck Q libration Test Reauired

1. Manual Reactor Trip N.A. N.A. S/U(1) N.A.

, 2. High Flux S D(2), and Q(7) M 1,2

3. RC High Terrperature S R M 1,2

4. Flux - AFlux - Flow S(4) M(3) a:d Q(7,8) M 1,2

5. RC low' Presstu e S R M 1,2

6. RC High Pressure S R M _ 1,2 RC 6 ture w 7. S R N M 1,2

? " 8. High Flux / Number of Reactor Coolant S R M 1,2 My Punos On v

9. Containment High Pressure S R '

M 1,2

'10. Informadiate Ranga, Neutron Flux S R(7) and Bate S/U(5) (1) 1,2 and*

~

11. Sourcc. Range, Neutron Flux and Rate S R(7) M and S/U(1)(5) '2,3,4 and 5

12. Control Rod Drive Trip Breakers N.A. N.A. M armi S/U(1) 1,2 ard*

1 13. Reactor Trip Module Logic N.A. N.A. M 1,2 and*

14. Shutdown Bypass High Pressure S R M 2 * * , 3 * * ,4 ** , 5*

E e-ga

~

g= .

~

=0 85 ~

l sr;2 & -

i *O .

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Table 4.3-1 (Continued)

WTATIQi (1) - If not perfonned in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL IGE.

(3) - When TdERMAL IWER [TP) is above 50% of RATED THERMAL IGER [RTP) and at steady state, conpare cut-of-core measured AXIAL IWER IMIaIANCE [APIo) to incore measured AXIAL IGER IMBAIANC (APII ) as follows:

M [APIo - APII ) = Offset Error TP Recalibrate if the absolute value of the Offset Error is 2 2.5%.

(4) - AXIAL PCWER IMBAIANC and locp ficw indications only.

(5) - Verify at least one decade cuerlap if not verified in previous 7 days.

I (7) - Neutron detectors may be excluded frm CHANNEL CALIBRA". ION.

(8) - Flow rate measurement sensors may be excitried frm O{ANNEL CALIBRATION. However, each flow measurement sensor shall be calibrated at least once per 18 nonths.

- With any control rod drive trip breaker closed.

- - When Shutdown Bypass is actuated.

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3/4 3-8 Babcock & Wilcox 8-35 ,uco,, moi, comp,ny

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3/4.4. REACIOR COOIRTP SYSTEM l l

3/4.4.1. COOIRTP LOOPS AND CDOIRTP CIRWIATICH

)

STARIUP AND KWER OPERATION 1 LIMITING 00NDITICN FDR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant punps in each loop shall be in operation.

APPLICABIII1Y: } ODES 1 and 2*.

ACTION:

a. With one reactor coolant punp not in operation, STARIUP and KMER OPERATION may be initiated aM may proceed provided 'IHERMAL POWER is i restrictM to less than 80.6% of RATED 'IHEPHAL KMER and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the setpoints for the following trips have been reduccd in accordance with Specification 2.2.1 for operation with three reactor coolant punps f operating:

1. High Flux j

2. Flux-AFlux-Flow (

SURVEIIIANCE RKUIRDETPS l

4.4.1.1 'Ihe above required reactor coolant loops shall be verified to be in operation and ciru11ating reactor coolant at least once per 12 ;

hours.

4.4.1.2 'Ihe Reactor Protection System trip setpoints for the instrunentation channels specified in the ACTION statenent above shall be verified to be in accordance with Specification 2.2.1 for the applicable number of reactor coolant punps operating either:

a. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after switching to a three punp ccmbination if l the switch is made while operating, or

b. Prior to reactor criticality if the switch is made while shutdown.

See Special h st Exceptian 3.10.3.

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v4aj m u ......

A MCDermott Comparty

DERGDiCY CORE COOLING SYSIDE BOPATED NA'IER S'IDRAGE TANK LIMITING CONDITICN FDR OPERATION 3.5.4 'Ihe borated water storage tank (BWST) shall be operable with:

a. An available borated rater volume of between 482,778 and l 550,000 gallons,

b. Between 1800 and 2200 ppu of boron, and

c. A mininum water temperature of 350F.

APPLICABILITY: FDDES 1, 2, 3 and 4.

ACI' ION:

With the borated water storage tank inoperable, restore the tank to OPERABI.E status within one hour or be in at least HCTT STRfDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ard in CDI.D SHUIDCEN within the followiry 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEIIlANCE RBDUIRD4L7FS 4.5.4 The fEST shall be dcawhated OPERABIE:

a. At least on per 7 days by:

1. Verifying the available borated water volume in the tank, l

2. Verifyiry the boron corcud. ration of the water.

b. At least once per 24 hcurs by verifying the water tenperature when outside air tecperature <35 0F.

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3/4 5-7 Pabcock & Wilcox 8-37 ,ycgy ,,,,,,,,,n,

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1 REACTIVITI COMPOL SYSTDG j BASm :

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3/4.1.1.4 MINIEN TEMPER 7'IURE FOR CRITICALITY

'Ihis specification ensures that the reactor will not be made critical with the reactor coolant system average tenperature less than 525 0F. 'Ihis limitation is required to ensure (1) the nederator temperature coefficient is within its analyzed temperature range, (2) the protective instrumentation /

is within its nomal operating range, (3) the pressurizer is capable of being in an OPERABI.E status with a steam bubble, ani (4) the reactor pressure vessel is above its mininen RI'g tenperature.

3/4.1.2. MPATION SYSTDG

'Ihe boron injection system ensures that regative reactivity control is 4 available during each mode of facility operation. 'Ibe ccr:ponents required to perfonn this function include (1) borated water sourms, (2) makeup or DIR punps, (3) separate flow paths, (4) boric acid punps, (5) associated j heat tracing systems, and (6) an emergency power supply from operable emergency busses.

With the RCS average tenperature above 2004, a minimum of two separate and redundant baron injection systems are provided to ensure single functional h

mP ility in the event an - tmad failure renders one of the systems incperable. Allowable out-of-service periods ensure that minor cwwwt repair or corrective action may be cmpleted without undue risk to overall facility safety frun injection system failures during the repair pericd.

'Ihe boration capability of either system is sufficient to provide a SHL7IDOWN MARGIN frun all operating conditions of 1.0% Ak/k after Xenon decay and cooldown to 2000F. 'Ihe maxirum boration capability requirement m' irs frun full power equilibrium xenon conditions and requires the equivalent of either 7373 gallons of 8742 ppn borated water frun the boric acid storage tanks or 52,726 gallons of 1800 ppn borated water frun the borated water storage tink.

'Ihe requirement for a minimum available volume of 482,778 gallons of borated l water in the borated water storage tank ensures the capability for borating the RCS to the desired level. 'Ihe specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4; therefore, the larger volume of borated water is specified.

With the RCS tenperature below 2004, one injection system is acceptable withcut single failure cr>nsideration on the basis of the B 3/4 1-2 8 38 Babcock & Wilcox A MCDermott CompJny

REACTIVITY CWIPOL SYSIDS BASES 3/4.1.2 BORATION SYSTDiS (Cbntinued) stable reactivity condition of the reactor and the additional rest rictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system W inoperable.

'Ihe boron capability required below 200 0 F is sufficient to provide a SHUIDOWN MARGIN of 1% Ak/k after xenon decay and cooldown frun 2000 F to 700F. 'Ihis coniition requires either 600 gallons of 7875 ppn borated water frun the boric acid storage system or 3,000 gallons of 1800 ppn borated water frun the borated water storage tank.

'Ihe bottan 4 inches of the borated water storage tank are not available, and I the instrumentation is calibrated to reflect the available volume. All boric acid tank volu:ne is available. Ihe limits on contained water volume, .

and boron concentration ensure a pH value of between 7.0 and 11.0 of the solution recirullated within containment after a design basis accident. 'Ihe pH band minimizes the evolution of iodine and minimizes the effect of

( chloride and caustic stress corrosion cracking on mechanical systems and ca ponents.

'Ibe OPERABIIII'Y of one boron injection system during REFUELING ensures that this system is available for reactivity wubvl while in MDDE 6.

3/4.1.3 M3VABIE CWIPOL ASSDmTTm

'Ihe specifications of this section (1) ensures that acceptable. power distribution limits are raintained, (2) ensure that the mininum SHUItXWi MARGIN is maintained, and (3) limit the potential effects of a red ejection accident. OPERABILITY of the control rod pcsition indicators is required to determine control red positions and thereby ensure capliance with the control rod alignment and insertion limits.

'Ihe ACI'ICti statenents which pennit limited variations frun the basic requime_nts are anvrmnied by adiitional restrictions which ensure that the original criteria are met. For exanple, miulignment of a safety or regulating rod requires a restriction in 'IHERMAL 1:OWER. 'Ihe reactivity worth of a misaligned red is limited for the remainder of the fuel cycle to prevent ermiing the assu:ptians used in the safety analysis.

'Ihe position of a rod declared inoperable due to misalignment should not be l included in ca puting the average group position for determining the

) OPERABILITY of IU9s with lesser misalignments.

B 3/4 1-3 Babcock & Wilcox 8-39 a ucoermoti company

D4ERGDiCY CDRE ODOLDJG SYSTDE BASES With the RCS temperature below 280 F, 0 one OPERABLE ECCS subsystem is acceptable without sirgle failure consideration on the basis of the stable reactivity cordition of the reactor and the limited core coolirq requirements.

'Ihe Surveillance Requirements provided to ensure OPERABIIITI of each caponent ensures, that, at a mininum, the assunptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. 'Ihe decay heat receval system leak rate surveillan requirements assure that the leakage rates assumed for the systen durirq the recirculation phase of the low pressure injection will not be ex = dad.

Surveillance requirements for throttle valve position stops and flow balance testirg provide assurance that prtper ECCS flows will be maintained in the event of a IOCA. Paintenance of prtper flow resistance and pressure drop in the piping system to each injection point is newy to: (1) prevent total pump flow from exMrg runaut conditions when the system is in its

~

minirum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assunptions used in the ECCS-IDCA analyses, ard (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that memad in the ECCS-IDCA analyses.

3/4.5.4 BORATED WATER SIOPAGE 'IANK

'Ihe OPERABIIITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a IDCA. 'Ihe limits on BWST mininum volume ard boron concentration ensure that 1) sufficient water is available within containment to permit recirculation coolirg flcw to the core, and 2) the reactor will remain subcritical in the cold condition followirg mixing of the BWST ard the RCS water volumes with all control rods inserted except for the most reactive control assembly. 'Ibese assunptions are consistent with the IDCA analyses.

'Ihe bottan 4 indles of the borated water storage tank are not available, and the instrumentation is calibrated to reflect the available volume. 'Ihe limits on contained water volume, ard boron concentration ensure a pH value of between 7.0 and 11.0 of the solution sprayed within containment after a design basis accident. 'Ihe pH band minimizes the evolution of iodine ard minimizes the effect of chl.oride and caustic stress corrosion cracking on mechanical systems and cmponents.

B 3/4 Babcock & Wilcox 8 0 & Mtottmott Company

DESIGi FTATURES lESIGi PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 40 psig and a tenperature of 2640F.

3.3 REACIOR CDRE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 177 fuel a-blies with each fuel assembly containing 208 fuel rods clad with Zircaloy-4. Each fuel rod shall have a naminal ac'.ive fuel length of 144 inches aM contain a maxim 2m total weight of 2500 grams uranium. The initial core loading shall have a maximum enrichment of 3.0 weight percent U-235. Reload fuel shall be similar in physical design tc. the initial core loading and shall have a maxinum enrienment of 3.3 weight percent U-235.

s CDtTTROL RODS 5.3.2 The reactor core shall contain 53 safety and regulating ard 8 axial power shaping (APSR) control rods. The safety and regulating cxx1 trol rods shall contain a n minal 134 inches of absorber material. The n minal values of a**ar material shall be 80 permnt Silver,15 percent Indium and 5 l percent Cadmium. All control rods shall be clad with stainless steel tubing. The APSRs shall contain a nminal 63 inches of absorber material at their lower ends. The absorbe.c material for the APSRs shall be 100%

Inconel-600.

5.4 REACIOR CDOINTP SYSTEM ,

DESIGi PRESSURE AND TEMPERA'IURE i

5.4.1 The reactor coolant system is designed and shall be maintained:

a. In aamrdance with the code requirem2nts specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to applicable Surveillance Requirements.

b. For a pressure of 2500 psig, and

c. For a tenperature of 650 0F, except for the pressurizer ard pressurizer carge line which is 6700F.

5-4 Babcock & Wilcox 8-41 a uco,,most comp,ny

9. STARIUP PROGRAM - HlYSICS TESTDM he planned startup test program associated with core perforn~nce is cutlined belw. Rese tests verify .that core performance is within the assumptions of the safety analysis and provide informatien for continued safe operation of the unit.

9.1. Precritical Tests 9.1.1. Cbntrol Rod Trio Test Precritical control rod drop times are recorded for all control roda at hot full-flw conditions before zero pwer physics testing begins. Acceptance criteria state that the red drop time from fully withdrawn to 75% inserted shall be less than 1.58 seconds at the conditions above.

It should be noted that safety analysis calculations are based on a red drop from fully withdrawn to two-thinis inserted. Sirm the most accurate position indication is obtained from the zone reference switch at the 75%-inserted position, this position is used instead of the two-thinis inserted position for data gathering.

9.1.2. RC Flw Reactor coolant flow with four RC pumps running will be neasured at hot I standby conditions. Acceptance criteria require that the measured flow be within allowable limits. '

9.2. Zero Power Fhysics Tests 9.2.1. Critical Boron Concentration Once initial criticality is achieved, equilibrium boron is obtained ard the critical boron concentration detemined. W e critical boron concentration is calculated by correcting for any rod withdrawal required to achieve the all rods out equilibriun boron. me acceptance criterion placed on -

critical boron concentration is that the actual boren concentration must be within i 100 ppu boron of the predicted value.

9-1 Babcock & Wilcox a McDermott company

c. -

9.2.2. Te m r;ture Reactivity Coefficient he isothemal HZP temperature coefficient is measured at approximately the all-rods-cut configuration. During changes in temperature, reactivity feedback may be canpensated by control rod movement. Se change in reactivity is then calculatal by the summation of reactivity associated with the temperature change. Acceptance criteria state that the measured value shall not differ from the predicted value by more than i 0.4x10-2%

ok/k/ OF.

We moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predictai value of fuel Doppler coefficient of reactivity is subtracted to obtain the moderator coefficient. Eis value must not be in excess of the acceptance criteria limit of +0.9x10-2g ak/k/ OF.

9.2.3. Control Red Group / Boron Reactivity Worth Control rod group reactivity worths (groups 5, 6, and 7) are measured at hot zero pcuer corxiitions using the boron / rod swap methal. mis technique consists of establishing a deboration rate in the reactor coolant system and cocpensating for the reactivity changes frun this deboration by inserting control rod groups 7, 6, and 5 in incremental steps. Se reactivity changes that occur during these neasurenents are calculated based on reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the change in rod group position. Se differential rod worths of each of the controlling groups are then summed to obtain integral red group worths. Se acceptance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

credicted value - measured value measured value x 100 5 15

2. Sums of groups 5, 6, and 7:

credicted value - measured value

~

measured value x 100 5 10 he boron reactivity worth (differential boron worth) is measural by dividing the total inserted rod wcrth by the boron change made for the red 9-2 Babcock & Wilcox a McDermott company i

worth test. S e acceptance criterion for measured differential boron worth is as follows:

1. credicted value - measuicd value measured value x 100 $ 15 he predicted rod worths and differential boron worth are taken from the PIN.

9.3. Power Escalation Tests >

9.3.1. Core Symmetry Test he purpose of this test is to evaluate the symetry of the core at lw pwer during the initial power escalation following a refueling. Symetry evaluation is based on incore quadrant power tilts during escalation to the )

intemediate pwer level. Se core symetry is acceptable if the absolute values of the quadrant pwer tilts are less than the error adjusted alam limit.

9.3.2. Core Power Distribution Verification at Intermediate Power level (IPL) and 100% FP Withj1cminal Control Rod Position Core power distribution tests are performed at the IPL ard 100% full power (FP). Equilibrium xenon is established prior to both the IPL and 100% FP tests. Se test at the IPL is essentially a check of the power distribution in the core to identify any abnormalities before escalating 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 100t FP operation.

We followirg acceptance criteria are placed on the IPL and 100% FP tests:

1. We worst-case maximum Um must be less than the IDCA limit.

2. We minimum DE must te greater than the initial condition De limit.

3. Se value obtained from extrapolation of the miniatum DE to the next power plateau overpower trip setpoint must be greater than the initial condition DE limit, or the extrapolated value of irbalance must fall f

outside the RPS power /irbalance/ flow trip envelope.

f 4. Se value obtained from extrapolation of the worst-case maximum Um to the next power plateau overpower trip setpoint must be less than the 9-3 Babcock & WilcOX a McDermott company

fuel mit limit, or the extrapolated value of imbalance nust fall outside the RPS power / imbalance / flow trip envelope.

_ 5. We quadrant poser tilt shall not exceed the limits specified in the

'Ib::hnical Specifications.

6. We highest measured and predicted radial peaks shall be within the following limits:

predicted value - measured value measured value 100 nore positive than -5

7. We highest measured aM predicted total peaks shall be within the followirg limits:

credicted value - measured value 100 more positive than -7.5 measured value Items 1, 2, ard 5 ensure that the safety limits are maintained at the IPL and 100% FP.

Items 3 and 4 establish the criteria whereby escalation to full power may be acomplished without exceeding the safety limits specified by the safety analysis with regard to [EBR and linear heat rate.

Items 6 ard 7 are established to detemine if measured and predicted power distributions are consistent.

9.3.3. Incore Vs. Excore Detector Irbalance Correlation Verification at the IPL Imbalances, set up in the core by control ' red positioning, are read j simultaneously on the incore detectors ard excore power range detectors.

We excore detector offset versus inmre detector offset slope must be greater than 0.96. If this criterion is not met, gain amplifiers on the excore detector signal processing equignent are adjusted to provide the required gain.

9.3.4. Tenperature Reactivity Coefficient at s100% FP We average reactor coolant temperature is decreased ard then increased at constant reactor pcuer. We reactivity associated with each temperature .

. \

change is cbtained frcra the charge in the controlling rod group position. 1 Controlling rod group worth is neasured by the fast insert / withdraw methcd.

Se terperature reactivity coefficient is cal.culated from the measured 9-4 Babcock & Wilcox a McDermott company

b

)

changes in reactivity and tenperature. Acceptance criteria state that the moderator tenperature coefficient shall be negative, p.3.5. Power Docoler Reactivity Coefficient at s100% ]T We power Doppler reactivity coefficient is calculated fraa data recorded during control rod worth measumments at power using the fast insert / withdraw method.

Se fuel Ecppler reactivity coefficient is calculated in conjunction with the power Doppler coefficient measurement. W e power Doppler coefficient as measured above is nultiplied by a precalculated conversion factor to obtain the fuel Ibppler coefficient. his measured fuel Doppler coefficient nust be more negative than the acceptance criteria limit of -0.90 x 10-3%

ok/k/ OF.

9.4. Procedure for Use if Acceptance Criteria Not Met If acceptance criteria for any test are not met, an evaluation is performed before tae test program is continued. mis evaluation is perfomed by site test personnel with participation by Babcock & Wilcox technical personnel as required. R1rther specific actions depend on evaluation results. Wese

- actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for ancnalies, or design personnel

- perfoming detailed analyses of potential safety problems because of

- parameter deviation. Power is not escalated until evaluation shows that p

plant safety will not be comprcnised by such escalation.

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h 9-5 Babcock & WilCOM a McDermott company f -- - -

10. REFEPHJCES

1. Davis-Besse Unit 1, Final Safety Analysis Report, Docket No. 50-346.

2. Davis-Besse Nuclear Power Station No. 1, Updated Safety Analysis Report, Docket No. 50-346.

3. Program to Detemine In-Reactor Performnce of B&W Fuels - Claddirg CrEcp Colla $?, PAW-10084P, Rev. 2, B3Wk arri Wilcox, Lynchburg, VA, October 197E..

l

4. TACD2: Fuel Performnce Analysis, IRW-10141P-A, Rev. 1, B' h k &

Wilcox, Lynchburg, Virginia, June 1983.

5. B,Wk & Wilcox Version of PD2 User's Manual, BAW-10117P-A, Babcock &

Wilcox, Lynchburg, Virginia, January 1977.

l l 6. NOODLE - A Multi-Dimensional TwoH3rcup Reactor Sin 11ator, JWW-10152A, Babcock & Wilcox, Lynchhlrg, Virginia, June 1985.

l 7. Ctreparison of Core Physics Calculations with Measurements, BAW-10120, Rhk & Wilcox, Dfnchburg, Virginia, June 1978.

l

8. merml-Hydraulic Crossflow Applications, BAW-1829, April 1984.

l 9. LYNX 1 Reactor Fuel Assembly herml-Hydraulic Analysis Code, PAW-10129-A, July 1985.

10. LYNX 2 Subchannel h ermal Analysis Program, BAW-10130-A, July 1985.

11. LYNXT Core Transient Werml-Hydraulic Program, RAW-10156-A, February 1986.

12. Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000-A, May 1976.

13. Divis-Besse Unit 1 Fuel Densification Report, BAW-1401, Babcock &

Wilcox, Lynchburg, Virginia, April 1975.

14. FCCS Evaluation of D&W's 177-FA Raised-Loop NSS, BAW-10105, Rev. 1, R'Wk & Wilcox, Lynchburg, Virginia, July 1975.

10-1 Babcock & Wilcox a McDermott company

ML20154H368

Text

SUMMARY

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