ML20006E365

Cycle 7 - Reload Rept.

ML20006E365
Person / Time
Site:	Davis Besse
Issue date:	11/30/1989
From:	FRAMATOME COGEMA FUELS (FORMERLY B&W FUEL CO.)
To:
Shared Package
ML20006E361	List: ML20006E360 ML20006E364 ML20006E365 ML20006E367
References
BAW-2096, NUDOCS 9002220746
Download: ML20006E365 (67)

Text

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lI 18 DAVIS-BESSE NUCLEAR POWER STATICN 1

UNIT 1,' CYCLE 7 - RELOAD REPORP  !

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1. INITODUCTION AND SUM 4ARY . . . . ................. 1-1 I 2. OPERATING HISTU E ........................ 2-1 i

3. GENERAL DESGIPTION ....................... 3-1

4. - IVEL SYSTEM IESIGN . . . . .................... 4-1 4.1. Fuel Assembly Mechanical Design .............. 4-1 4.1.1. . Mark BBA Fuel Assembly naw ription . . . . . . . . . 4-1 4.2. Fuel Rod Design ...................... 4-2 Claddirg Collapse ................. 4-2

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4.2.1.

4.2.2. Claddirg Stress ..................

4-3 4.2.3. Claddirq Strain .................. 4-3 4-4

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4.3. shamal Design . . . . . . . . . . . . . . . . . . . . . . .

4.4. Material ocupatibility . . . . . . . . . . . . . . . . . . . 4-4 i

-4.5. Operatirq Experience . . . . . . . . . . . . . . . . . . . . 4-5

5. NUCI. EAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 .

5.1. Physics Characteristics .................. 5-1 '

i' 5.2. Charges in Nuclear Design ................. 5-2

6. THERMAIr-HYNAULIC DESIGN . . . . .................. 6-1

7. ACCIIENT AND TRANSIDir ANALYSIS ................. 7-1

,' 7.1.

7.2.

- Safeey - ysis Accident Evaluation 7-1 7-1

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8. PROPOSED 20DIFICATIONS 'IO TECHNICAL SPECIFICATIONS . . . . . . . 8-1.

I 9. STARIUP PROGRAM - PHYSICS TESTING ................ 9-1 l

J 9.1. Precritical Tests ..................... 9-1 9.1.1. Control Rod Trip ' Dest ............... 9-1 9.1.2. RC Flow- . . . . . . . . . . . . . . . . . . . . . . 9-1 '

c 9.2. Zero Power Physics Tests . . . . . . . . . . . . . . . . . . 9-1 L 9.2.1. Critical Boron Crs.=udation . . . . . . . . . . . . 9-1 9.2.2. %WTature Reactivity Coefficient . . . . . . . . . 9-2 9.2.3. Control Rod Group / Boron Reactivity Worth . . . . . . 9-2 l Baw Fedcompas

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, 00tmNrs (Cont'd)

, og, g 9.3. Power Escalation 'Ibsts . . . . . . . . . . . . . . . . . . . 9-3 9.3.1 Core Symmetry Test . . . . . . . . . . ....... 9-3 9.3.2. Core Power Distribution Verification at

[ Irfam=4iate Power Isvel (IPL) and~100%FP With Nominal Control Rod Position . ....... . 9-3 9.3.3. Incore Vs. Excore Detector Imbalance I Correlation Verification at the IPL ....... . 9-4 9.3.4. '!%mperature Reactivity Coefficient at =100%FP ... 9-4 9.3.5. Power Doppler Reactivity Coefficient at =100%FP .. 9-5 Prrv'achnt for Use if Acceptance Criteria Not Met . .. . . .

9.4. . 9-5

10. REFERENCES . ........................... 10-1 I Tint of Tables Table 4-1. Fuel Design Parameters .................... 4-6 5-1. . Davis-Ramaa Unit 1, Cycle 7 Physics Parameters ....... 5-3 L

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l. 5-2. . Shutdown Margin Calculation for Davis-Raaaa, Cycle 7

..... 5-4 6-1. '!hermal-Hydraulic Design Conditions . . . . . . . . . . . . . . 6-3

,g, 7-1. Couparison of Fey Parameters for Accident Analysis ..... . 7-3 Eg 7-2. Bounding Values for Allowable IDCA Peak Linear Heat Rates . . ... . . . . . . . . . . . . . . . . . . . 7-3 ,

8-1. Quadrant Power Tilt Limits (Tech. Spec. Table 3.2-1) ..... 8-27 E.

g Tint of Flaures I 5

Figure 3-1. Davis-Raaaa cycle 7 Core Ioading Diagram ........... 3-2 .

3-2. Enrichment ard Burnup Distribution for I Davis-Raaaa 1, Cycle 7 .. . . . .-. . . . . . . .-. . . . . . . 3-3 ,1 3-3. Control Rod Iccations for Davis-Ramaa 1, Cycle 7 ...... . 3-4 -)

'3-4. -Davis-Raama Cycle 7 BPRA Enrichment and Distribution ..... 3-5 l y 5-1. BOC (4 EFPD), Cycle 7 'No-Dimensional Relative Power Distribution - Full Power, Egillibrium Xenon, All Rods LI: Out, APSRs Inserted . .................... 5-5 )

8-1. Regulating Group Position Limits, O to 30 +10/-0 EFPD, 1 Four RC Punps -- Davis-Besse 1, Cycle 7 L (Tech. Spec. Figure 3.1-2a) . . . . . . . . . . . . . . . . . . 8-7 8-2. Regulating Group Position Limits, 30 +10/-0 to 75 +10/-0 EFPD, Four RC Punps - Davis-ham 1, Cycle 7

<E ('Ibch. Spec. Figure 3.1-2b) . . . . . . . . . . . . . . . . . . 8-8 I E 8-3. Regulating Group Position Limits, 75 +10/-0 to 335 i10 EFPD, I l Four RC Punps - Davis-Besse 1, Cycle 7 (Tech. Spec. Figure 3.1-2c) 8-9 l

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List of Flaures (Con't) l B.99

I-8-4. Regulating Group Position Limits After 335 110 EPPD, Four RC Punps - Davis-Besse 1, Cycle 7 ]

(Tech. Spec. Figure 3.1-2d) . ................ 8-10 .

' I. 8-5. Regulating Group Position Limits, O to 30 +10/-0 EFPD, I

'Ihree RC Punps, - Davis-he 1, Cycle 7 l 1

(Tech. Spec. Figure 3.1-3a) . . ................ 8-11 8-6. Regulating Group Position Limits, 30 +10/-0 to 75 +10/-0 EFPD,. I

'Ihree RC Punps, - Davis-Besse 1, Cycle 7  !

('Dach. Spec. Figure 3.1-3b) . . . . . . . . . . . . . . . . . . 8-12 8-7. Regulating Group Position Limits, 75 +10/-0 to 335 d10 EFPD,

' 5- 'Ihr,ee RC Punps, - Davis-hma 1, Cycle 7

('Dach. Spec. Figure 3.1-3c) . . . . . . . . . . . . . . . . . . 8-13 8-8. Regulating Group Position Limits After 335 i10 EFPD, ,

'Ihree RC Punps -- Davis-Besse 1, Cycle 7

('Ibch. Spec. Figure 3.1-3d) . . . . . . . . . . . . . . . . . . 8-14 8-9. APSR Position Limits, O to 335 i10 EFPC, Four RCPunps - Davis-hea 1, Cycle 7 I

(' Inch. Spec. Figure 3.1-Sa) . . . . . . . . . . . . . . .... 8-15 8-10. APSR Position Limits After 335 10 EFPD, 'Ihree or Four RC Pumps, APSRs Withdrawn - Davis-he 1, Cycle 7 IL .

(Tech. Spec. Figure 3.1-5b) . . . . . . . . . . . . . . . . . . 8-16 8-11. APSR Position Limits, O to 335 i10 EFPD, 'Ihree RC Punps - Davis-he 1, Cycle 7 (Tech. Spec. Figure 3.1-5c) . . . . . . . . . . . . . . . . . . 8-17 8-12. AXIAL POWER IMBAIANCE Limits, O to 30 +10/-0 EFPD, Four RC Punps - Davis-he 1, Cycle 7 (Tech. Spec. Figure 3.2-la) . . . . . . . . . . . . . . . . . . 8-19 I 8-13. AXIAL POWER IMBAIANCE Limits, 30 +10/-0 to 75 +10/-0 EFPD, Four RC Punps - Davis-Besse 1, Cycle 7 (Tech. Spec. Figure 3.2-lb) . . . . . . . . . . . . . . . . . .

8-20

.I~ 8-14.- AXIAL POWER IMBAIANCE Limits, 75 +10/-0 to 335 i 10 EFPD, Four RC Punps - Davis-he 1, Cycle-7 (Tech. Spec. Figure 3.2-1c) . . . .............. 8-21 I' 8-15. AXIAL ICWER IMBAIANCE Limits, Afte> 135 110 EFPD, Four RC Punps - Davis-Besse 1, Cycle 7

('Dach. Spec. Figure 3. 2-1d) . . . . . . . . . . . . . . . . . . 8-22 8-16. AXIAL POWER IMBAIANCE Limits, O to 30 +10/-0 EFPD,

': 'Ihree RC Punps - Davis-Besse 1, Cycle 7

-(Tech. Spec. Figure 3.2-2a) . . . . . . . . . . . . . . . . . . 8-23 8-17. AXIAL POWER IMBAIANCE Limits, 30 +10/-0 to 75 +10/-0 5- EFPD, 'Ihree RC Punps - Davis-Besse 1, Cycle 7

('Dech. Spec. Figure 3.2-2b) . . . . . . . . . . . . . . . . . . 8-24

. 8-18. AXIAL POWER IMBAIANCE Limits, 75 +10/-0 to 335 10 EFPD,

'Ihree RC Punps - Davis-Besse 1, Cycle 7 (Tech. Spec. Figure 3.2-2c) . . ................ 8-25 8-19. AXIAL POWER IMBAIANCE Limits, After 335 10 EFPD,

g 'Ihree RC Punps - Davis-Besse 1, Cycle 7

3. ('Ibch. Spec. Figure 3. 2-2d) . . . . . . . . . . . . . . . . . . 8-26

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INIRODUCI' ION AND Sut99RY l

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'Ihis report justifies operation of Davis-Daaaa Nuclear Power Station Unit 1  !

at the rated core power of 2772 NWt for cycle 7. h e required analyses are included as outlined in the Nuclear Regulatory Comission (NRC) document,

" Guidance for Prr= = d License Amendments Relating to Refueling," June 1975.

'Ihis report utilizes the analytical techniques and design bases that have been submitted to the NRC ard approved by that agency.

Cycle 7 reactor and fuel parameters related to power capability are sumarized in this report and capared to cycle 6. All accidents analyzed in the ' Davis-Besse Final Safety Analysis Report1 (FSAR) or the Updated Safety Analysis Report 2 (USAR), as applicable, have been reviewed for cycle 7 ope:1stion. 'Ibe only transients analyzed were the loss of forced reactor I- coolant flow and locked rotor events. R ese were performed to verify that cycle 7 parameters were bounded by previous analyses. In all cases, the cycle 7 parameters are bounded.

Cycle 7 is the first cycle of Davis-Da==a 1 that will contain fuel assemblies with Zircaloy irdamac11 ate spacer grids. _ 2e incoming batch 9 fuel is of this type and is designated as Mark B8A. Se Mark B8A fuel assembly is an improved Mark B5A fuel assembly which has been used in previous cycles. h e improvements are di r ==ael in section 4.

- ' h e analyses supporting Zircaloy-grid fuel assemblies were submitted to the Se NRC Safety Evaluation 4 of reference 3 states that NRC in reference 3.

the report can be referenced for reload analysis for all lowered-loop B&W designed 177 fuel assembly plants. Davis-Besse 1 is a raised-loop plant and requires additional discussion concerning applicability of reference 3.

Se nuclear, thermal-hydraulic, and non-IDCA safety analysis are unaffected by the differences between raised-loop and lowered-loop plants. Se analyses presented in reference 3 are equally applicable to either configuration. The major areas that musc be addressed because of this 1-1 l: B&W FuelCompany

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charge are the seismic-IOCA loads analysis and the loss-of-coolant-accident analysis. 'Ibese are addramW in sections 4.1 and 7.2 respectively.

The Technical Specifications have been reviewed and modified whare required for cycle 7 operation. RaaM on the analyses performed, taking into account the amertjency core cooling system (ECCS) Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Davis-Manaa Unit -

1 1, ' cycle 7 can be operated safely at its licensed oore power level of 2772 g.

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2. OPERATING HIS'ICRY l

'Ibe reference cycle for the nuclear ard thermal-hydraulic analyses of Davis-naaaa Unit 1 is the currently operatirg cycle 6, which achieved criticality on Da'amhar 5, 1988. Power escalation began on nm,amhar 15, 1988 and full power (2772 Mft) was attained on nm,amhar 30, 1988.

During cycle 6 operation, no operating anomalies occurred that would adver-sely affect fuel performance during cycle 7. 'Ibe cycle 6 and cycle 7 licensed lengths are 415 and 425 effective full power days (EFPD), res- )

pectively. Cycle 7 was analyzed out to 415 EFPD and the applicability of .

the cycle 6 reactor protection system (RPS) limits arri setpoints to cycle 7 has been verified out to 425 EFPD. '1he cycle 7 operating. limits have also been verified out to 425 EFPD. 'Jhe APSRs will be pulled at 325 i 10 EFPD to

'Ihe APSR pull coupled with a power I increase the lifetime of cycle 6.

coastdown will result in a potential cycle 6 length of 387 EFPD. 'Ihe cycle 7 design also includes an APSR pull and power coastdown.

'Ihe cycle 7 design minimizes the number of fuel assemblies that are . cross core shuffled to reduce the potential for quadrant tilt anplification. 'Ihe cycle 7 shuffle pattern is di s maai in section 3.

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3. GDGPAL DTSCPitPTICH I

The Davis-ham Unit i reactor oore is described in detail in chapter 4 of the USAR2 for the unit. The cycle 7 core consists of 177 fuel assemblies (FAs), eat of which is a 15x15 array otr.%ining 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. All TAs in cycle 7 are 'nade up of batches 7, 8, and 9 whim have a constant ncaninal fuel loading of 468.25 kg of uranium. The fuel consists of dished-end cylirdrical pellets of uranium dioxide clad in cold-worked Zircaloy 4. The undensified nominal active fuel lengths, theoretical densities, fuel ard fuel rod dimensions, and other related fuel parameters are listed in Table 4-1.

Figure 3-1 is the core loading diagram for Davis-Besse Unit 1, cycle 7. One batch 2A assenbly, 48 bat & 6 assemblies and 11 bat & 7A assenblies will be I dis &artyed at the and of cycle 6. The fuel assenblies in bat & es 7B and 8 will be shuffled to their cycle 7 locations, with bat & 8 on the core periphery. Bat &as 7 ard 8 have initial enrichments of 3.19 and 3.13 wt %,

respectively. The feed batch, consisting of 60 batch 9 asseenblies with uranium enrichment of 3.38 wt %, will be imarted in the core interior in a syninetric checkerboard pattern with the bat & 7 TAs. Figure 3-2 is a  ;

quarter-core map showing each assenbly's expected burnup at the beginning of cycle (BOC) 7 ard its in.ttial enrichment.

Cycle 7 is operated in a feed-and-bleed mode. The core reactivity is con-l

-trolled by 53 full-length Ag-In-cd control rod assemblies (CsAs), 60

burnable poison rod assenblies (BPRAs), ard soluble boron. In addition to the full-length control rods, eight Inoonel-600 axial power shaping rods (gray APSRs) are provided for additional control of the axial power distribution. The cycle 7 locations of the control rods and the group designations are irdicated in Figure 3-3. The core locations and the rod group designations of the 61 control rods in cycles 6 ard 7 are the same.

The cycle 7 locations and enrichments of the BPPAs are shcun in Figure 3-4.

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gl rigure 3 1. Davis-Besse Cycle 7 Core Loading Diagram I'  !

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I E6 L3 na F NB F Nit L13 K10 t E

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F K2 8 9 F

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F M

A10 F 8

K14 9

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8 9 M 9 M 9 M 9 M 9 M 9 8 i 0

F9 F 84 F M F 812 F 010 F Die F F7 l

g 8 8 9 M 9 M 8 3 9 M 9 2 8 CIO 89 F C3 F 86 R3 811 F Cl3 F 87 C6 8 8 9 M 9 4 9 M 9 M 9 4 9 8 8 f I (10 011 F F2 F N1 F M F AB F F14 F 06 (6 ,

8 9 4 9 78 9 4 8 4 9 4 9 4 9 8 8

De F F1 F  !! F A7 04 Gil F (14 F Fil F 07 8 8 9 78 8 4 8 4 8 M 8 78 9 8 8

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N13 F 02 G1 kl3 M14 03 Ell C4 ble F M6 N3 g 8 9 4 9 M 9 M 8 M 9 4 9 M 9 8 E

1. 9 F L1 F It F El Cl! RB F N14 F Lil F N7 f

8 8 9 78 9 78 9 M 3 4 l 9 9 9 8 8 f

. L MID hl! F Lt F R$ F A7 F Hll F Ll4 F 11 6 15 8 8 9 M 9 78 8 M 9 78 9 8 8 M 010 P9 F 03 F P6 013 Pil F 013 F P7 06 8 9 78 9 3 9 4 9 75 9 75 9 s L9 F 18 2 ' F P6 F P4 F P10 F Plf F L7 i l'

78 9 8 9 M 9 4 9 8 9 3 .

O Cl3 F G2 F R6 F RIO F Cl4 F E3 P B 8 8 9 8 9 8 8 8 G6 F3 (4 F L'8 F (12 F13 l GIO .

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F6 G4 C8 Gl! Fil

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C Cycle 6 Location (F e Fresh Assembly)

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i N 9ure 3-2. Enrichment and Burnup Distribution for Davis-Besse 1. Cycle 7 I 8 9 10 11 12 13 14 15 l

3.19 3.13 3.19 3.13 ' 19 3.38 3.13 3.13 [

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29731 12004 22126 12001 22201 0 17242 16417 l 3.13 3.19 3.38 3.19 3.38 3.19 3.38 3.13 l K

12004 22119 0 24593 0 22141 0 17136 i

3.19 3.38 3.19 3.38 3.19 3.38 3.13 3.13 22126 0 24322 0 24740 0 16722 17409 j

3.13 3.19 3.3B 3.19 3.38 3.13 3.13 M

I 12001 24583 0 23789 0 14110- 16228  ;

3.19 3.38 3.19 3.38 3.19 3.38 3.13 l 22201 0 24742 0 22178 0 17032 f I 3.38 3.19 3.38 3.13 3.38 3.19 0 22127 0 14104 0 30109 3.13 3.38 3.13 3.13 3.13 P

I' 17224 0 16702 16212 17014 -

3.13 3.13 3.13 16408 17115 17340 I

x.xx Initial Enrichment l '

xxxxx BOC Burnup, mwd /mtV

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Figure 3-3. Control Rod Locations for Davis-Besse 1. Cycle 7 I! ,

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E 2 5 5 2 5' i F 4 8 6 3 6 8 4 G 5 1 1 5 H W- 6 7 3 4 3 7 6 -Y. ,

K 5 1 1 5 L 4 8 6 3 6 8 4 it i 2 5 5 2 N l 7 8 7 8 7 ,

0 2 5 5 2 1 l P l l A 6 4 R .

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i 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 g X Group Number Group ___ No. of Rods Function 1 4 Safety 2 8 Safety 3 4 Safety 4 9 Safety 5 12 Control 6 8 Control l 7 8 Control 3 l

8 8_ APSRs 5..

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I I Figure 3-4. Davis-Besse Cycle 7 BPRA Enrichment and Distribution I 3 9 10 11 12 13 14 15 H 14 5 K 1.4 1.7 0.2

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I x.x BPRA Concentration, wt% B C in A1 0 .

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4. WEL SYSTIM DESIGN 4.1 Ibel Asammb1v Mechanical Desicrn

.I m e fuel assembly type and pertinent fuel design parameters for Davis-namaa 1 cycle 7 are listed in Table 4-1. Batch 7B and 8 are the Mark BSA design. .

Batch 9 is the Mark BSA design. i 4.1.1 Mark BBA Fuel Assenbly Description he Mark B8A fuel assembly 19 an inprwed Mark B5A fuel ast.ambly with design I features to permit easy recons-titution, to allw for high burnup, to prevent misalignment during fuel shuffle, and to prwide protection against debris fretting Anaga to the fuel rods. 7b permit easy r m astitution, the upper

.g 5 end fitting is designed to be renwable in the field. S e top spacer grid has been enhanced to prwide extra stiffness, and the top spacer skirt has .3 been removed. To prwide for high burnups, the lower and fitting has been shortened by approximately 0.7 inches. Se guide and instrument +mham were lengthened by the same amount, and the fuel rod lengthened by .4 inches.

This prwides additional growth rocn for the fuel rod.

To prevent misalignment when inserting fuel assemblies into the core, anti- ,

straddle bars have been a%r1 to the lower and fitting. % ese bars prevent  ;

the fuel assembly from being fully inserted unless it is in proper position  ;

over the lower oore grid pad. ,

To protect against debris induced fretting failure of the fuel rod the following design changes were made. The lower end plug solid portion was extended in length. S e lower spacer grid location was lowered so that the 1.t' solid and plug extends through the lower spacer grid. The intention of the design is to trap any debris capable of fuel rod fretting below the bottm spacer grid where the solid icwer end plug will prevent failure.

The use of the solid lower end plug in the Mark BBA fuel assenbly fuel rods j

results in a loss of gas volume (plenum volume) inside the fuel rod when l corpared to the Mark BSA fuel assembly fuel rods. The reduction in plenum 4-1 l

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volume would result in higher intamal pressure with burnup otmpared to the Ii f stardard design. 'Ihe decrease in gas volume was offset in part by t largthening the fuel rod by 0.4 irches. 'Ic obtain similar mechanical ard thermal performance between the two fuel rod designs, the fill gas pressure was reduced in the Mark BBA fuel assembly rods. Both fuel rod designs were analyzed for cycle 7.

'Ibe Mark BBA fuel assenblies incorporate Zircaloy rather than Inoonel for j the six intermediate spacer grids. Zircaloy has a substantially ar. slier neutron absorption cross section than inoonel. As a result, the separative work requirements are reduced ard the overall uranium utilization is improved. 'Ibe justification for use of Zircaloy rather than Inoonel for the g intan=44 ate spacer grids is contained in reference 3. 'Ibe NRC safety W ,

evaluation 4 of that report requires that a licensee who is incorporating that design submit a plant-specifia analysis of combized seismic and IDCA loads accordirg to Appendix A to Stardard Review Plan 4.2. 'Ihis analysis is i

currently being performed and will be ocmpleted prior to the introduction of Zircaloy grids into Davis-namaa Unit 1.

Sixty BTRAs will be used with the 60 bat & 9 fuel assemblies.

4.2 Tbel Rod Desian

'Ibe fuel rod design and mechanical evaluation are di=>==ai in this section. ,

'Ihere are two fuel rod designs. 'Ibe standard Mark B design is used with the Mark BSA fuel assenbly in batches 7B and 8. 'Ihe other design used with the Mark BBA fuel assembly in batch 9 has several design changes from the standard design. 'Ihese changes are the elimination of tubular spacers ,

inside the rod, a reduction in fill gas pressure, and the use of a long solid lower erd plug for debris resistance. 'Ibe design changes result in a fuel rod with a debris resistance feature and similar mechanical and thermal performance to the standard design.

4.2.1 ClaMhx! Collanaa E,

'Ihe operating power history for the most limitirq fuel assembly was determined for each of the three fuel batches. 'Ihe history for each batch was ocmpared to that used in the generic creep collapse analysis. Batches g 7B and 8 slightly exoseded the generic envelope. A new envelope was W fornulated ard a new creep collapse analysis was performed. Both the new 4-2 I

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B specific analysis aM the generic analysis are based on the methodology described in reference 5.

We analysis predicted a creep collapse life longer that 35,000 EFE for batches 7B and 8. mis is longer than the =wi== batch residence in cycle 7 which is 28,700 EFW for batd) 7B.

For batch 9 with the IcWer fill gas pressure, the creep collapse analysis follcued the method frun reference 6. Se operational conditions and mechanical characteristics of the batd19 fuel assemblies were ocmpared to an envelope form 11ated by BhTC (reference 6) and approved by the imC (reference 7) . All values of the Mark BBA fuel amamblies are bounded by Scane as-built data .

I the congding parameters of the limiting envelope.

for the Mark B8A assemblies (presently unavailable) were approximated from as-built values of past BhTC fuel and then ccmpared to the limiting envelope. his is reasonable as the tolerances effecting these as-built values have not changed frun past BWFC fuel designs. Se creep collapse life of the batch 9 fuel rods based on references 6 and 7 is 55,000 mwd /mtU.

4.2.2 ClaMin

Stress A conservative fuel rod stress analysis envelopes the Davis-hama Unit 1 cycle 7 stress values for the Mark BSA and Mark BBA fuel assetly fuel rod designs. %e methods used for the analysis of cycle 7 have been used in the previous cycles.

4.2.3 Claddina Strgia Se fuel rod design criteria specify a limit of 1% on cladding plastic tensile circumferential strain. W e fuel pellet is designed to ensure that plastic cladding strain is less than 1% at the design local pcllet burnup and heat generation rate. Se design values are higher than the worst-case values Davis-Besse Unit 1, cycle 7 fuel is expected to experience. he strain analysis is based on enveloping the upper tolerance limits for the fuel pellet diameter and density, and the 1cuer tolerance limit for cladding inside diameter.

)I Both the Mark BSA and Mark BBA fuel assembly fuel rod designs were analyzed and found to be limited to less than 1% strain urder design conditions.

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E; 4.3 mar ==1 Desian me fuel designs utilized in the cycle 7 core are thermally similar. me g' design of the batch 9 Mark B8A assemblies is such that the thermal 5 performance of this fuel is equivalent to the fuel design used in the I reairder of the core. Se reduction in plenum volume resulting frm the  :

debris-resistant batta and plug was offset by a reduced fill gas pressure.

Fuel thermal analyses were performed with the 2002 8 fuel pin performance code. N m inal undensified input parameters used in the analysis are l present.ed in Table 4-1. Densification effects were accounted for in the ,

TA002 code densification model.

Linear heat rate (DR) to fuel melt capability for all fuel was determined ,

with the BCD2 fuel pin performance code. W e analyses performed for cycle 7 demonstrate that 20.5 )M/ft is a conservative limit to preclude centerline fuel melt (CFM) for all fuel batches.

me maximum fuel pin burnup at the end of cycle (EOC) 7 is predicted to be less than 43,100 M*l/mtU (batch 7B) . Nel rod internal pressure has been evaluated with BCO2 for the highest burnap fuel rod ard is predicted to be ,

less than the reactor coolant system pressure of 2200 psia at the core outlet. ,

3 4 Material omontibility The compatibility of all possible fuel-cladding-coolant-assembly g' interactions for batch 9 fuel assemblies is identical to that of present 5 fuel.

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4.5. Ooaratim Dcoerience NWR & Wilcox cperating experience with the Mark B 15x15 fuel assembly I has verified the adequacy of its design. 'Ihe folicwing experience has been accumalated for eight B&W 177 fuel assenbly plants using the Mark B fue',

assembly:

Cunulative Current Max FA Barnuo,MRi/mtU(a) Net Electric '

Reactor Cvele Discinarged Outout.21h (b)

I Incore Ooonee 1 12 37,128 58,310 78,666,485-I Ooonee 2 Ooonee 3 11 11 30,982 33,026 42,820 39,701 73,898,798 73,029,790

'Ihree Mile Island 7 27,566 33,863 41,163,976 Arkansas Nuclear 9 32,348 57,318 60,732,134 '

One, Unit 1

'I Rancho Seco Crystal River 3 7

7 34,123 35,410 38,268 35,350 42,140,543 48,449,202 l

Davis-hama 6 26,848 40,300 32,582,548 (a)As of June 30, 1989.

(b)As of Hardt 31, 1989.

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i Table 4-1. Fuel Desian Parameters i Batch ,

7B B 9  !

Fuel ammanhly type Mark BSA Mark BSA Mark BBA' No. of assenblies 53 64 60 l

Fuel rod OD, in. 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 Flexible spacer type Spring Spring Spring -

Tubular spaomr Zr-4 Zr-4 NA Undensified active 143.2 143.2 143.2 ,

fuel length, in.

Fuel pellet (mean) j dia., in. 0.3686 0.3686 0.3686 ,

Ibel pellet initial 95.0 95.0 95.0 density, %'ID maan Initial fuel enrictnant 3.19 3.13 3.38 wt % 235U Average burnup 23,940 15,924 0  ;

BOC, M d/stU Exposure tima li IDC, . (EFMI) 28,700 19,200 10,000 mJ Cladding collapse >35,000 >35,000 NA g .'

time, EFPli g Mav4== assembly l burnup, md/mtU 42,471 31,213 18,644 Cladding collapse NA NA 55,000 hirmp, Md/mtU.

Naninal linear h heat rate at gI 2772 Wt, kW/ft 6.14 6.14 6.14 5'l Mininnn linear 20.5 20.5 20.5 heat rate to melt, W/ft Ij 4-6 l B&W FudCompany lq- ,

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5. NOCLEAR DESI @

5.1. Ihysics Characteristics Table 5-1 cartpares the core Ihysics parameters for the cycle 6 and 7 de-signs. The values for cycles 6 and 7 were both generated with the NOODLE code9 . Differences in core physics parameters are to be e@ected between  !

the cycles due to the changes in fuel and burnable poison enridiments which create changes in radial flux and barriup distributions. Figure 5-1 illustrates a ivpresentative relative power distribution for the BOC 7 at  !

full power with equilibrium xenon, all rods out and gray APSRs inserted.

The ejected rod worths in Table 5-1 are the maxinn calculated values.

Calculated ejected rod worths and their adherence to criteria are considered

• at all times in life and at all power levels in the development of the rod I position limits presented in section 8. The adequacy of the shutdown margin with cycle 7 rod worths is shown in Table 5-2. The following conservatisms were applied for the shutdown margin calculations:

1. Poison material depletion allcwance.

2. 10% uncertainty on net rod worth.

3. A mwi=n flux redistribution penalty.

4. A maxinn power deficit with mininn boron.

The maxinn flux redistribution was taken into account to ensure that the effects of operational maneuvering transients were included in the shutdown margin analysis.

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I' 5.2. Charnes in Nuclear Desian Ii 7here are no significant oore design charges for cycle 7. 'Ihe calculaticaml -

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 cr procedural charges exist with regard .

to axial or radial power shape, xenon, or tilt control. The stability and control of the core with APSRs withdrawn has been enalyzed. 'Ibe calculated stability index without APSRs is -0.044 h-1, Wch h-ates the axial stability of the core.

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Tahkad-1. Davis -- Unit 1, evele 7 Physics IE--- emts  ;

cvele 6 cvele 7 Design cycle' length, EFPD 405 415  !

Cycle burnqp, )M&'atU 13,545 13,880 l

'I Average oors burnup - EOC, PM$'atU Initial core 1 , stU 24,335 82.9 26,806 82.0 1

l Critical boron (8 -_ BOCr No Xe, ppn '

l HZP 1,451 1,506 HFP 1,285 1,323 Critical boron (a) . EOC, Eq. Xe, ppm j HZP 204 HFP 10(b) 200(b) to Control rod worths - HFP, BOC, % Ak/k -1 Group 6 1.18 1.14 1 1.05 1.03 l Group 7 f Group 8 0.21 0.19 Ocx1 trol rod worths - HFP, IDC, % Ak/k Group 7 1.14 1.09 i Group 8 NA NA Max ejected rod worth - HZP, % Ak/k BOC, Groqps 5-8 inserted (Ir10) 0.36 0.32 EOC, Groups 5-7 inserted (Ir10) 0.40 0.38 I- Max stuck rod worth - HZP, % Ak/k

.BOC (M-13) 0.66 0.60 BOC (M-11) 0.79 0.81 Power deficit-- HZP to HFP, Eq. Xe, % Ak/k BOC (4 EFPD) -1.71 -1.71  ;

BOC -2.51 -2.55 Doppler coeff(- HFP, '10-3 % Ak/r/0F BOC, No Xe c) Group 8 inserted -1.55 -1.58 ,

BOC, Eq. Xe, O ppm, Group 8 withdrawn -1.84 -1.86  ;

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Moderator  % ak/k/ F BOC, No coeff)-

Xelc HFP,10-2 -0.59 -0.70  !

EOC, Eq. Xe, O ppm -2.84 -2.93 Boron worth - HFP, pptV% ak/k '

BOC (1323 ppm) vs (1285 ppm)(c) 124 129 EDC 107 110 Xenon worth - HFP, % Ak/k .h BOC (4 EFPD, equilibritzn) - 2.63 2.61 a I:' EOC (aqiilihvium) '2.78 2.78 Effective delayed neutrun' fraction - HFP BOC 0.00626 0.00618 EOC 0.00518 0.00520 1

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

(b) Power coastdown to EOC at 10 ppn.

(c) Cycle 7' values were calculated at 1323 ppnb; cycle 6 values were calculated at 1285 ppmb.

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Table 5-2. Sht#% Namin N1culatieri for Davis-W% Cvele 7 E3C, % Ak/k BDC, 345 EFPD 415 EFPD

% Ak/k Groun 8 in Grouc 8 out Avm41mbia Rcri hkarth

'Ibtal rod worth, HZP 7.22 7.53 7.54 Worth reduction due to burtiup -0.42 -0.42 -0.42 of poi m material i M== stuck rod, HZP -0.60 -0.74 .2,.lL1 Net worth 6.20 6.37 6.31 Imss 10% unoartainty Q.i.fi2 -0.64 -0.63 i

'Ibtal available worth 5.58 5.73 5.68 Raquired Rod Worth i Power deficit, HFP to HZP 1.71 2.45 2.55 Max allowable inserted rod worth 0.28 0.42 0.44 i Th x redistribution- -

O.30 0.72 _.Q,il2 Total required worth 2.29 3.59 3.71 Sht#% Namin j

'Ibtal available ininus total 3.29 2.14 1.97 '

required rod worth i Notet Required shutdown margin is 1.00% ak/k.

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I Figure 5-1. BOC (4 EFPD), Cycle 7 Two-Dimensional Relative Power Distribution - Full i Out, APSRs Inserted l( qwer. Equilibrium Xenon, All Rods I 3 9 10 11 12 13 to is l M .981 1.214 1.121 1.249 1.128 1.292 1.049 .560 I K 1.216 1.113 1.292 1.086 1.261 1.120 1.173 .532 8 1 L 1.121 1.290 1.107 1.262 1.062 1.230 .852 .377 M 1.248 1.084 1.258 1.104 1.279 1.102 .637 I N 1.127 1.260 1.061 1.280 1.069 1.069 .436 ,

I 1.104 1.072 .450 0 1.293 1.121 1.231 ,

1 P 1.050 1.174 .853 .639 .438 R .561 .533 .378 f x Inserted Rod Group Number L x.xxx Relative Power Density lI (a) Calculated results from two-dimensional pin-by-pin PDQ07.

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'Ibe Mark B8A fuel assenblies inserted for cycle 7 have Zircaloy intermediate spacer grids. 'Ihe thermal-hydraulic design evaluation supporting cycle 7 operation utilized the methods ard models described in references 10,11 and I 12 as supplemented by reference 3, Which implements the BWC (reference 13)

OF oorrelation for analysis of Zircaloy-grid fuel assemblies. 'Ihe analyses presented in section 5 of reference 3 demonstrate that changes in the flw parameters resulting fra the incorporation of Zircaloy spacer grids do not significantly inpact the thermal-hydraulic characteristics of a Zircaloy-grid ,

oore relative to the standard Inoonel-grid (Mark B) core. Inplementation of the Zircaloy-grid fuel assenblies into existing reactors, however, is I performed on a batch basis, with the transition cycles having both Zircaloy-grid and standard Mark B fuel assemblies.

'Ibe Mark B8A fuel assembly has a slightly higher pressure drop than the standard Mark B assenbly due to the higher flw resistance of the Zircaloy spacer grids. 'Ihe presence of Mark BBA fuel assemblies in a predcminantly Mark B core, will, therefore, tard to divert some flw from the more restrictive Mark B8A anaamblies to the Mark B fuel. As a result, the Mark B8A fuel assenblies in a mixed core will experience slightly less coolant I' flw than in a hmogeneous Mark BBA core. 'lhis reduced flw results in a reduced thermal margin for the Mark BBA assemblies relative to a full Mark B8A core. 'Ibe amount of coolant flw reduction is dependent on the number of Mark B8A ==aamblies (with the smaller number of Mark B8A assemblies being more limiting) . A " transition core penalty" nust, therefore, be considered for the introduction of Zircaloy-grid assemblies into a standard Mark B core.

For cycle 7 of Davis-Besse Unit 1, this transition penalty is offset by the consideration of a core bypass flw fraction in the thermal-hydraulic model that is higher than the actual value.

6-1 l Baw Fusicompany 1

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BE Niesigned reactors, including Davis-Besse Unit 1, currently operate  !

without orifice rod assemblies in the control rod guide tubes (CRGPs). The core bypass fraction is dependent on the number of unplugged guide tubes, 1 Whit:h is in turn der--dent on the number of burnable poison rods (BPRAs) ard control rod assemolies (CPAs), since these mycnents restrict flw through the CEGrs. For the thermal-hydraulic analysis, the most limiting case is j L that with the higher bypass flow fraction, or smaller number of BPRAs. )

The design basis chosen for cycle 7 thermal-hydraulic analyses was a full Ziroaloy v id core, containing 37 BPRAs, for Which the core bypass f1w is 8.9%. This design configuration was used to calculate the 1.54 DiBR (112% )

FP) shown on Table 6-1 for cycle 7. The actual cycle 7 core configuration consists of 60 fresh Mark BBA fuel assemblies and 117 standard Mark B fuel j

assemblies, and 60 BPRAs, resulting in a core bypass flow of 8.3%. The DiBR l l for this configuration, using the same core conditions presented in Table 6-2, is 1.56. Ocuparison of the DiBRs for the design ard actual core configurations shows that the design configuration is conservative for cycle f 7 DiBR analyses and that a transition core penalty is, therefore, not e=U. Table 6-1 provides a sumary ocmparison of the DiB analysis i parameters for cycles 6 and 7.

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Table 6-1. Maximmn Desian Carxiitions. Ovelas 6 ard 7 Cvele 6 Ovele 7 Rated thernal power level, M9t 2772 2772 Ncninal core exit pressure, psia 2200 2200 Minima core exit pressure, psia 2135 2135 Reactor coolant flw, gpn 380,000 380,000 Nerninal vammal inlet coolant I tenperature (100% FP), OF 557.4 557.4 Ncaninal vessel outlet coolant tartparature (100% FP), CF 606.6 606.6 Core bypass f1w, 4 (a) 8.6 8.9 INBR hiing Crossflw Crossflw I' Design radial x local powerpeakingfactor,FN A 1.71 1.71 Design axial flux shape 1.65 ct@ cosine 1.65 ctM cocine Hot channel factors

. Enthalpy rise 1.011 1.011 I Heat flux Flow area 1.014 0.98 1.014 0.97 (D)

Active fuel length, in. 143.2 143.2

. I.l Avgheatfluxgt100% power, 105 Bt4/h-ft 1.86 1.86 Maxheatfluxgt100% power, I -105 Btu /h-ft 5.25 5.25 QIF correlation B&W-2 BWC CHF correlation INB limit 1.3 1.18 mn -

I at 102% power at 112% power 2.07 1.79 1.78 (b) 1.54 (b)

(a)Used in the analysis.

(b) Calculated for the instrument guide tube subchannel Which is limiting for the Mark B8A fuel assernblies.

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7. ACCIDENT AND 'IRANSIDTP ANALYSIS 7.1. General safety Analvsis rach USAR accident analysis has been examined with respect to &anges in the cycle 7 parameters to determine the effects of the cycle 7 reload and to >

ensure that thermal performance during hypothetical transients is not degraded. 'Ibe effects of fuel densification on the USAR accident results have been evaluated and are reported in reference 14.

'Ibe radiological dose w_;gwr.i:s of the FSAR Chapter 15 accidents have been evaluated using conservative radionuclide source terms that bound the cycle specific source term for Davis-namaa 1 cycle 7. 'Ihe dose calculations were performed consistant with the assunptions described in the Davis-Basse 1 FSAR r but used the more conservative source terms (whid bound future reload I. cycles). 'Ihe results of the dose evaluations showed that offsite '

radiological doses for each accident were below the respective acceptance i criteria values in the current NRC Standard Review Plan (NURB3-0800).

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Accident Evaluation I

7.2.

'Ihe key parameters that have the greatest effect on determining the outcome ,

of a transient can typically be classified in three major areas: (1) oore I thermal, - (2) thermal-hydraulic, and (3) kinetics parameters including the reactivity faa**ck ocefficients and oortrol rod worths.

Fuel thermal analysis parameters from ead batch in cycle 7 are given in '

Table 4-1. 'Ihe cycle 6 and cycle 7 thermal-hydraulic mav4== design conditions are presented in Table 6-1. A comparison of the key kinetics parameters from the USAR and cycle 7 is provided in Table 7-1.

A generic loss-of-coolant accident (IDCA) analysis for the B&W 177-FA raised-loop nuclear steam system (NSS) has been performed. 'Ibe Final Acceptance Criteria B&W ECCS Evaluation Model techniques ard ===tions (as described 7-1

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in BAW-10104P, Rev. 515) were used in the analysis. 'Ibe application of the Evaluaticn Mcdell6 includes the inpacts of the NUREG-0630 fuel pin rypture ]

curves and the MICSLT refloodirg heat transfer coefficient calculations. In '

addition, the IMC OF oorrelation was used to determine the time of INB. 'Ibe ocznbination of average fuel tenparatures as a function of linear heat rate (IJR) and the lifetime pin pressure data for the Mark BBA fuel used in the 1DCA limits analysis is bourded by those calo11ated for the B&W 177-FA  ;

raised-loop plant for previous reload evaluations.16 A tabulation showing the allowable 1DCA IRRs for Davis-Besse Unit 1, cycle 7 fuel is provided in Table 7-2. 'Ibese limits have generally been reduced frun previous cycles due to the inpacts of NUREG-0630, FIECSLT, and the IMC OF oorrelation.

It is concluded by the examination of cycle 7 core thermal, thernal- '

hydraulic, and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ability to safely S g,

operate the Davis-Ramaa Unit 1 plant durirg cycle 7. Considering the previously accepted design basis used in the FSAR and snhaaquent cycles, the transient evaluation of cycle 7 is considered to be bourded by previously accepted analyses. 'Ibe initial canditions of the trensients in cycle 7 are l bounded by the USAR ar4/or the fuel densification report.

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B Taole 7-1. hviam of Fev Paranstars for Accident Analysis I ~

densif'n I Par =

• report value cycle 7 value BOL(a) Doppler oceff,10'3, % Ak/k/DF -1.28 -1.58 EOL(b) Doppler coeff,10'3, % Ak/k/T -1.45(c) ~1.86 BOL moderator coeff,10'2, % Ak/k/0F +0.13 -0.70 EOL moderator coeff, 10'2, % Ak/k/0F -3.0 -2.93 Total rod grtup worth (d) (HZP), % Ak/k 10.0 7.22 Boron reactivity worth (HFP), pptV1% Ak/k 100. 129 Max ejected rod worth (HFP), % Ak/k O.65 0.26 Max &M rod worth (HFP), % Ak/k O.65 0.20 Initial boron conc (HFP), ppn 1407 1323 I (a)BOL denotes beginning of life.

I (b)EOL denotes and of life (c).1,77 x 1o-3 % Ak/k/ 4 was used for steam line failure analysis.

(d) exclude APSRs (Groqp 8)

Table 7-2. Mmoina vali== for A11=hle IDCA Peak Linear Heat Rates Allowable Allowable Allowable Core Peak LHR, Paak IHR, Peak IHR Elevation, O to 30 EPPD, 30 to 75 EFPD, After 75 EFPD, ft kW/ft kW/ft _ kW/ft 2 16.0 16.5 16.5 4 15.75 17.2 17.2 6 16.5 18.0 18.4 I 9 17.25 17.5 17.5 10 17.0 17.0 17.0 7-3

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8. FROICSED )CDIFICATIONS 'IO TEONICAL SPECIFICATICNS l 2he 'hachnical Specifications have been revised for cycle 7 operation to f acocunt for changes in power peaking and control ro:1 worths. 'Ihe core operating limits are based on an ECCS bcunding analysis that was perfonned to determine the allcwable I.CG linear heat rate limits for the B&W 177 fuel asMmbly raised-loop plant. 'Ihe analysis irwrWrated the NURm-0630 I

i cladding rupture data, TACD2 Mk BBA fuel assenbly data, the IHC OfF oorrelation, and the FIREET reflooding heat transfer coefficient corre-latico. Figures 8-1 through 8-19 are revisions to the previous cycle opera-ting limits. Based on these operating limits the final acceptance criteria ECCS limits will not be exceeded arxl the thermal design criteria will not be violated.

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I 2.1 SAM:TY 12MITS l I

BASES i 2.1.1 ard 2.1.2 REACKR CDRE The restrictions of this safety limit prevent overheatirg of the fuel clackiirg ard possible claddirg perforation which would result in the release 3 of fission products to the reactor coolant. Overheatirg of the fuel claddirg g is prevented by restrictirg fuel option to within the nucleate boilirg regime where the heat transfer coefficient is large and the claddiM. surface tenperature is slightly above the coolant saturation tanparature.

Operation above _ the upper boundary of the nucleate boilirg regime would result in excessive cladding tanparatures because of the onset of departure g fran nucleate boilirg (INB) and the resultant sharp reduction in heat 3-transfer coefficient. INB is not a directly measurable parameter during operation and therefore THE:RMAL POWER and Reactor Coolant Temperature and a

,^ Pressure have been related to INB using critical heat flux (CHF) corre- g

! lations. The local INB heat flux ratio, INBR, defined as the ratio of the heat flux that would cause INB at a particular core location to the local heat flux, is indicative of the margin to INB.

The B&W-2 and BWC OIF correlations have been developed to predict INB for J axially uniform and non-uniform heat flux distributions. The B&W-2 3; correlation applies to Mark-B fuel and the BWC correlation applies to all gi i B&W fuel with Zircaloy spacer grids. The mininum value of the INBR durirg )

steady state operation, normal operational transients, and anticipated trantdants is limited to 1.30 (B&W-2) and 1.38 (BWC) . These values corre-

,I spond to a 95 percent probability at a 95 percent confidence level that INB l will n:.t occur and is chosen as an appropriate margin to INB for all )

operatiry conditions. ,

The curve presented in Figure 2.1-1 represents the conditions at which a minimam INLR equal to or greater than the correlation limit is predicted for l gl the =vi== possible thermal power of 112% when the reactor coolant flow is gI 380,000 GPM, which is approximately 108% of design flow rate for four )

operating reactor coolant punps. (7he mininum required measured flow is 4 389,500 GIN.) This curve is based on the followirg hot channel factors with potential fuel densification and fuel rod bewirg effects: 1 F = 2.83; g (H = 1. 1; = 1.65 The design limit power peaking factors are the most restrictive calculated at full power for the range frun all control rods fully withdrawn to mininum allowable control rod withdrawal, ard form the core INBR design basis.

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'Ihe curves of Figure 2.1-2 are h=M cri the more Instrictive of two thermal l I ' limits and account. for the effects of potential fuel densification and potential fuel rod bow.

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1. 'Ibe INER limit pMv=4 by a ruclear power peaking factor of I Fg = 2.83 or- the ocabination of the radial peak, axial peak, and position of tne axial peak that yields no less than the INBR limit.

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2. 'Ibe ocatination of radial and axial peak that causes central fuel melting at the hot spot. 'Ibe limit is 20.5 kW/ft for s

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all fuel in the core.

Power peaking is not a directly r*=*nele quantity and therefore limits have been established on the basis of the reactor power inbalance prMv=d by the I ~ ~ -

' 'Ihe specified flw rates for the two curves of Figure 2.1-2 cott+4 to the 1

analyzed minimum flew rates with four punps and three punps, respectively.

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'Ibe curve of Figure 2.1-1 is the most restrictive of all possible reactor '

coolant puvv4== thermal power canbinations shown in IASES Figure 2.1. l l

'Ibe curves of IESES Figures 2.1 tasar-rit the conditions at which a mininum i INBR equal to the INBR limit is predicted at the mav4== possible thermal l ,

power for the number of reactor coolant punps in operation or the local I' quality at the point of mininum INBR is equal to the wituspunning INB correlation quality limit (+22% (B&W-2) or +26%

is more restrictive.

(BWC)) , whichever condition l 1

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- l BASES For the curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DER greater than 1.30 (B&W-2) or 1.18 (BWC) and a local quality at the point of minimum DER less than +22%

(B&W-2) or 426% (EHC) for that particular reactor coolant punp situation.

h DER curve for three punp cperation is less restrictive than the four punp curve.

2.1.3 REACIOR 000IANT SYSTf24 PRESSURE

'Ibe restriction of this Safety Limit protects the integrity of the Reactor Coolant System frm overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant frm reaching the containment l atmosphere. W

'Ibe reactor pressure vessel and pressurizer are designed to Section III of g the ASME Boiler and Pressure Va==1 code which permits a =vl== transient g pressure of 110%, 2750 psig, of design pressure. h Reactor coolant Systen piping, valves arx! fitt:.ngs, are designed to ANSI B 31.7, 196B Edition, which permits a =v4== transient pressure of 110%, 2750 psig, of ocuponent design pressure. 'Ihe Safety Limit of 2750 psig is therefore consistent with the design critaria and anwv iated code requirements.

h entire Reactor coolant System is hydrotested at 3125 psig,125% of design pressure, to demonstrate integrity prior to initial operation.

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LIMrrne SArrry sysTD1 serrnus I

The AXIAL PCHER IMEAIAN3 bourdaries are established in ortler to prevent I reactor thermal limits from being exceeded. These thermal limits are either power peaking )M/ft limits or INHR limits. The AXIAL KMER IMBAIANT reduces the power level trip produced by a flux-to-flw ratio such that the I

boundaries of Figure 2.2-1 are pr % M .

I RC Pressure - Im. Hidh and Pr===ure Tamarature i'

I 1he high and low trips are p2rvided to limit the pressure range in which reactor operation is permitted.

I I During a slw reactivity insartion startup accident frun low power or a slw reactivity insertion frun high power, the RC high pressure setpoint is reached before the high flux trip setpoint. 7he trip setpoint for RC high pressure, 2355 psig, has baen established to maintain the systen pressure I

i belw the safety limit, 2750 psig, for any design transient. The RC high pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, ard is therefore set lower than the set pressure for I these valves, $2525 psig. The RC high pressure trip also backs up the high flux trip.

The RC low pressure, 1983.4 psig, and RC pressure-taperature (12.60 tout" '

I 5662.2) psig, trip setpoints have been established to maintain the INB ratio greater than or equal to the minimum allowable INB ratio for those design l accidents that result in a pressure reduction. It also prwvents reactor I operation at pressures belw the valid range of DNB oorrelation limits, protecting against INB.

Hiah Flux /Nunber of Reactor Coolant Punos On ,

In conjunction with the flux - Aflux/ flow trip the high flux / number of reactor coolant punps on trip prevents the mininnan oore INBR frun decreasing below the mininum allowable [NBR ratio by tripping the reactor due to the l loss of reactor coolant punp(s). The punp monitors also restrict the power level for the number of punps in operation.

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REACTIVITt CCtmOL SYSMtS I; REGUIATING ROD INSIEPICtf LIMITS IlMITEC 00NDITICtf FOR OM:RATIOf 3.1.3.6 The regulating rod groups shall be limited in ytiysical insertion as sh Nn on Figuras 3.1-2a, 3.1-2b, 3.1-2c, and 3.1-2d, ard 3.1-3a, 3.1-3b, I!

3.1-3c, and 3.1-3d. A rod group overlap of 25 1 5% shall be maintained between sequential withdrawn groups 5, 6, and 7.

APPLICABIIJTV: PODElS 1* and 2*f.

ACTIO1: I:.

With the regulating rod groups inserted beyond the above insertion limits 3 (in a region other than acceptable operation), or with any group sequence or 5' overlap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either

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

b. Reduce 7HI:RMAL ICWER to less tlan or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or s

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

Egig If in unacceptable region, also see Section 3/4.1.1.1 I. .

I I.

I-

"See Special Test Exception 3.10.1 and 3.10.2. i fWith l'eff 21.0. -'

3/4 1 l 8-6 j

^~ ^ "~ ~ ~ ~ ~ ~ ~ ~' ~ ~ ~ ~ ~ ~ ^ ~ ~ ~ ~ ~ ~ ^ ~ ~ ~ ~ ' ' ~ ^

G;gg fA. 8 4

IiMR j-4 i-l' ] ,

Figure 8-1. Regulating Group Position Limits, 0 to 30 +10/-0 EFPD, Four RC Pumps -- Davis-Be:re 1. Cycle 7 (Tech Spec Figure 3.1-2a) J a

l POWER LEVEL (300,102)-

CUTOFF = 1001- (164,102) (273,102) - -

l 100 -

)

(268,92)-

, ' me

I ~ f

80 SHUTDOWN MARG'N (250,80)'

' LIMIT =

y ,

, OPERATION 4 , RESTRICTED UNACCEPTABLE

.g (234,60) j-w .60 OPERATION (104,60) lW - E-T* g

[' , :d

_$,40 -

'b

,I e -

E .t [ (79,28.5)

I q5 : 20- -

ACCEPTABLE .,

OPERATION

(0,3), * -

8 1 l0 - - - - - - - '

0. 100 .200 300 Rod Index (% Withdrawn)

- GR:.5 t i a-1 0 75 100 .

1 0 GR 6.1 a a > -

0 25 75 100 l GR 7 ' a 0 25 100 .

I- 4

.?

3 /4 1-28 8-7 gg gg

, + , . , . . .

v- - - - - . - . . _ . _ . . . _ . _ . _ . _ . _ _ __

g  :

' 5 W

4 Figure 8 2 - Regulating Group Position Limits 30 +10/-0 to 75 +10/-0 EFPD,-

Four RC- Pumps -- Davis-Besse 1,- Cycle - 7.

l]

(Tech Spec Floure 3.1-2b): l

' i L

POWER LEVEL.

CUTOFF 100t' (164'102)

(268,102) ,. (300102) 2 Il( '

100 ~

^ .

(263,92)  ; .

r SHUTDOWN g I

g 80 MARGIN (250,80) 5 LIMIT OPERATION E- RESTRICTED g!

3,

, k 60 - -

UNACCEPTABLE (104,60) (234,60) El L-- .c OPERATION g i , W-

! g  ;

'I' ,,

. .g 40 -

e I,

$ (79,28.5) ACCEPTABLE 4 w OPERATION g

-[' 20. -

3;

l

.(0,3)0' - - '- ' ' ' ' ' ' ' '

O 100 200 300 Rod Index (% Withdrawn)

GR 5. * * '

l 0 75- 100 GR 6 '. ' ' '

4 0 25 75 100 g '

GR 7 a 8 B

s: 0 25 100 I ,

Iu 3/4 -28a gggg g 4 .

f'

,.1 1

ll .

i Figure 8- 3 Regulating Group Position Limits, 75 +10/-0 to 335 2 10.EFPD, -l I'

Four RC' Pumps -- Davis-Besse 1. Cycle 7 i (Tech Spec Figure 3.1-2c) .

I -

.t t

I POWER LEVEL (228,102) 1 f

100 - CUT 0FF = 100%

• (263,92)- i E

80 - SHUTDOWN (250,80) .

MARGIN f LIMIT

' OPERATION c!

{ RESTRICTED j E! 60 - UNACCEPTABLE (186,60) (234,60) y OPERATION

~

,I W .

a 5 l40 -

'I 1.

.e s.,

(119,28.5)

ACCEPTABLE

-20 - - -

OPERATION 1

(0,3)', ' '! -

0

• 8 ' ' ' ' ' ' ' ' '

' 1

-0'

-I 100 Rod Index (% Withdrawn)

=

200 300 GR'5' ' e i 0 75 100 GR 6 i a a i I O 25 GR 7 75-0 100 25 100 I

I 3/4 1-28b MMM 8-9

. - .J

c- -- -- -- - --- -

,, 4:r , ,,

It .

W. ,_

l' .'

30 ,

~,.

l N't Figure 8-4.~ Regulating Group Position . Limits, After 335 t 10 EFPD, Four RC Pups -- Davis-Besse 1, Cycle 7 (Tech Spec Figure 3.1-2d) <

Il x

. POWER LEVEL (276,102) (300,102)

Il-w

-CUT 0FF = 100% -

, 100 - I

,w

~, (263,92) g

=. <

SHUTDOWN 80- (250,80)

MARGIN UNACCEPTABLE LIMIT j OPERATION l

8 -60 - '(200,60) X (234,60)- '

E i

g n

, OPERATION

$ 40 - RESTRICTED

.g

".:. L u -(162,28.5)

J

n. 20- -

ACCEPTABLE OPERATION- .

.(0.3):'

0 r i e i e i i- > 2 i i i

g 0~ 100 200 '300 5' Rod In,dex (1 Withdrawn) 0 75 100 - .

[ GR 6 i e i- a 0 25 75 100 GR 7 i a i l;,

0 25 100 i

E i

3/4 1-28e B&W FWlCW9MF 8- -

ll , Figure 8-5..j Regulating Group Position Liinits, O to 30 +10/-0 EFPD, p i Three RC Pumps -- Davis-Besse-1,. Cycle 7-L (Tech Spec figure 3.1-3a).

b .

l: .

100 -

cr .- ,

UNACCEPTABLE OPERATION (273,77) (300,77)

~

f 80 --

SHUTDOWN: (164,77) ^

0

~  % MARGIN I: LIMIT e (268.69) y b- ,W: r 60 -: (250,60): :i

[ .

o OPERATION-

. RESTRICTED s- ,

g .

3 2

g (104,45.5) (234,45)

~

[ 40 -

., l . v k.-

a.:

20 -

(79,21.-9) ACCEPTABLE OPERATION-l (0,2.8),  :

o e i e i . 2 i i . . .

0 :100 -200 300-Rod Index (%' Withdrawn)~

GR . 5 :' i i n 4

. 0- 75 100

GR 6 L n i -

0 25 75 100 GR 7 6.

i a 0 25 100 3/4 1-29 B&W FuelCongsmy  :

lh: ,

8-11

? $ ( 1 1

?

s .

.a i.: e -J

~ Figure' 8-6. Regulating Group Position Limits, 30'+10/-0 to 75'+10/-0 EFPD, -

.Three RC Pumps -- Davis-Besse 1,' Cycle 7 (Tech Spec Figure 3.1-3b) .

-. .Il .t t

t

+

100

'.l I E 80 -

(164,^77) e.'

g UNACCEPTABLE: SHUTDOWN (263,69) s

.w. OPERATION MARGIN ,

iE . LIMIT  ;

o - "'

ic

. K.; 60 '(250,60); '

OPERATION

, -I RESTRICTED g

, y (104,45.5) (234,45)

L 5. 40 -- '

g -

N. -

2-y

. L. . --

ACCEPTABL2 h j

-gl 20 (79,21.9)

OPERATION il (0,2.8{ 'M

'O a a a a * * ' '

i i i i h

0 100 _ _

200 300 j' Rod Index (% Withdrawn) 1 GR 5 L. e i

ii 14 0 75 100 L GR 6 a

• f i '

0 25 75 100 gi GR 7

• i ' '

l,$

.r.

3/4.1-29a 8-12 B&W M M Q

s I, .

Figure 8-7. Regulating Group Position Limits, 75 +10/-0 to 335 t10 EFPD, l Three RC Pumps -- Davis-Besse 1, Cycle 7 (Tech Spec Figure 3.1-3c) l

-100  :

l 7 l80 -

(228,77) (268,77) ,

(300,77) w

, Il UNACCEPTABLE SHUTDOWN -

l N

'y -OPERATION (263,69)  ;

. $ _ 00 -

(250,60)

E a:

h (186,45.5)

~(234,45)

{s40-

- j 4

OPERATION RESTRICTED u ,

l< . l 20

a. -

(119,21.9):

ACCEPTABL!.

' OPERATION.

lI',(0,2.8).

i -

.+

i . . .

O i i n a-i i i i i

,i ' ;

.0

100 200 300-i l= Rod Index (% Withdrawn)

GR 5'n a n

'f:

0_ 75 100 GR 6 i ' i i +

l 0 25 75 100 GR 7

• I '

0 25 100 ll Li i!

3/4 1-29b 8-13 BIE M M  !

-v' p '

, Ak'

.m i

Figure 8-8 . Regulating Group Position Limits, After 335 110 EFPD, j;;l 'Three RC Pumps -- Davis-Besse 1. Cycle 7 .l (Tech Spec Figure 3.1-3d).

31 h 5 "-

L,

[

.100 - '

j' j g~

L - 80 - '( 0' .

(300,77) i

)

SHUTDOWN (263,69)' 4 5 MARGIN.

UNACCEPTABLE LIMIT:

4 60 -

(250.60)-'

l-

, g OPERATION '

a:

,. (200,45.5) . (234,'45)

L40 t E OPERATION 1-

! RESTRICTED

-(162,21.9)

.-Al 20 ~

1

' ACCEPTABLE g-OPERATION 3' (0_ 2p , , , , ., , , ., , , , ,

O 100 Rod Index (% Withdrawn);-

200 300 h

GR 5'a ' '

0 75- 100 ;h GR . 6 ' ' ' ' <

O' 25 E7a 75 100 n '

g. '

0 25 100  ;

I 3" 2 i.77 n c.,,,, l;

'qL a

I,w;t Figure-8-9 , APSR Position Limits. O to 335 *10 EFPD, 1

' E :. Four RC Pumps -- Davis-Besse 1, Cycle 7 l

-5: (Tech Spec Figure 3.1-Sa)

I RESTRICTED REGION (0,102) (100,102). 4 100 K

g- f- 80 -

! _[

W-g g. r' l3 8 60 -'

PERMISSIBLE J0PERATING L 'Q

"; REGION 1:

,: . ii. 40 -

1 L

.- 6 8

!.. g s

u f

l[ 20 -

I:

L 0 ' ' ' ' ' ' 8 ' '

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

l r i ,

3/4 1-35 EU i 8-15

~-

g

J w: ,E'.

y; Figure 8-10; APSR Position Limits,- After 335 10 EFPD, E Three or-Four RC Pumps, APSRs Withdrawn -- 'E.'

Davis-Besse 1. Cycle 7 (Tech Spec Figure 3.1-5b). -: '

-100 - E

.m g

T-w  :.

. ! 80 -

g

5

.: iE i

^

.. S 60 -

j '

-Q APSR INSERTION NOT ALLOWED-IN THIS TIME INTERVAL-

'a 1 g: 40 -

.)

e H

.g -

j c  ;

. %' ~E' 3

2 20-0 i i e i i i i i i i I ^

.0 10 20 30. 40 50 60 70 80 90 100 E

l APSR Position (% Withdrawn) 51 T

I I'

bit B&W Fed Compey l

,- q,

-j

[ i L

~-

1.. .

I L .

1

[ . Figure 8-11. APSR Position Limits, O. to- 335 t10 EFPD, ,,

Three RC Pumps -- Davis-Besse 1, Cycle 7 (Tech Spec Figure 3.1-Sc) p ,

+

f .100 --

RESTRICTED e (0,77) REGION (100,77) 80 -

0 2

a 3

[  ;

l. .- .

1

'f,-

k 60.-

' e.

i W. PERMISSIBLE I- OPERATING REGION a

h= y .j q

?

., 40 4

- i 15; g.

~E-w.

k.gn. I

s. i

.. . O i i e i i i i i- i i l

0. 10 20 30 40 50 60 70 80 90. 100 APSR Position (% Withdrawn) i la '!

t e

I i:

4 3/4 1-37 B&W FuelConquiy 8-17

i. .- . ,

3/4.2 ~ RJWER DIS'IRTHJrION LIMITS AXIAL POWER IMIRIANG LIMITING 00NDITICN FOR OPERATICN 3.2.1 AXIAL POWER IMBAIAN3 shall be maintained within the limits shown on Figums 3.2-la, 3.2-1b, 3.2-1c, and 3.2-1d, and 3.2-2a, 3.2-2b, 3.2-2c, and g 3.2-2d.

E' APPLICABIITN: LODE 1 above 40% of RATED THERMAL IOWER.* -

ACTION:

With AXIAL POWER IMBAIANG eMrg the limits specified above, either:

a. Restore the AXIAL POWER IMBAIAN3 to within its limits within 15 minutes, or ,

b. Within one hour mduce power until imbalance limits am met or to 40%

of RATED 'IHERMAL PJWER or less. i

-- I, 4.2.1 'Ihe AXIAL POWER IMBAIANG shall be determined to be within limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of RATED 'IHERMAL POWER except when l the AXIAL POWER IMBAIANG alam is inoperable, then calculate the AXIAL ICWER IMBAIANG at least once per hour. -

gL Il 3J I l

'l l

l "See Special 'Dast Exception 3.10.1. ,

I l 3/4 2-1 1 8-18

, B&W FuelCompany lj

I 1

]

lL 1

1 Figure 8-12~. AXIAL POWER IMBALANCE Limits. O to 30 +10/-0-EFPD,  ;

Four RC Pumps -- Davis-Besse 1, Cycle 7- .,

(Tech Spec Figure 3.2-ia)- l j

u

~- i'

. 110

(-22,102) - - - 100 i' (' '

(-25,92) <

(15,92)

. . 90

.. . (-30,80)< _. 80 ,(20,80)

1. a:

l-

RESTRICTED- y RESTRICTED J. REGION. a. .. 70 REGION

(-30,60)'o . . 60 o (20,60) <

l '

5 PERMISSIBLE .

OPERATING g. . d0

, REGION i

-: g g. . 40 p5; ,

y. . 30 -

b b.

~

20 s.

, g' '

5: a. . . 10  :

, -50 30 -20 -10 0 10 2b 30 40 50 AXIAL POWER IMBALANCE (%).

l:

$I; b

I' l4 3/4 2-2 UIMM 8-19

4 I

g-Figure 8-13. AXIkl POWER IMBALANCE Limits, 30 +10/-0.to 75 +10/-0 EFPD, Four RC Pumps'-- Davis-Besse 1, Cycle 7 T

. (Tech Spec. Figure 3.2-1b) k 110

(-21,102)

~

(15,102) .

(-25,92) < ( 5,92)

- 90

(-30,80)<

I_

5 . 80 p(20,80) g .

RESTRICTED RESTRICTED

-REGION

- 70 '

REGION

(-30,60) o $< 60 o (20,60) .

S '

y- 50

. 40

(

PERMISSIBLEy, -i OPERATING t' ,

2, - -0 REGION 20 B.

-10 m

.g-i t i i i l i i 1 1-40 -30 -20 -10 0 10 20 30 40 50 AXIAL POWER IMBALANCE (%)

V'4 A

' ~

I

=

_ Figure 8- 14 AXIAL POWER IMBALANCE Limits,75 +10/-0 to 335 t 10 EFPD, .

Four RC' Pumps.-- Davis-Besse 1, Cycle 7

.' (Tech Spec Figure 3.2-ic)

) . 110 t.

- (-18.8.102 ^) "

(15,102)

< - 100

(-25,92) (15,92)

. . 90 4 ,

. t

(-30,80), p. 80 ,(20,80) w 5 , ._ RESTRICTED  % , 70 RESTRICTED REGION REGION g h".

.g: (-30,60).< > g.. 60 - o(20,60)

$. 50 E '

PERMISSIBLE g .. 40

'g; OPERATING ,

5 REGION 5 E.. 30 -  ;

.E-I i

3- e. . 20 1 l' '

. . 10  !

i e l i 1 1 a j i f s t

.(-

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

1

g >

g LI

I '" -2' ""*""'"*

8-21

v i

Figure 8-15. AX1AL POWER IMBALANCE Limits, Afte' 135 10 EFPD, Four RC Pumps--- Davis-Besse 1, Cycle =7 -

,, (Tech Spec Figure 3.2-1d)-

,,110

-f 8

(-21,102) - -

100.

} (13,102) -lti

(-25,92) ,

<(15,92)

I

(-30,80)1 p--80 '(20,80) -i E

'-- 70 .

I

(-30,60) .- o -

- 60 0(20,60)

RESTRICTED' RESTRICTED '

REGION. 8- - 50 REGION E.

g . u --

g-- 40 PERMISSIBLE

.0PERATING b-- 30 '

REGION

[e

-- 20 '

-10 i

l' 1 i l 1 i 1 l'

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

I;1 E!

I t 8-22

?

if j':

ill El ,

h >

- Figure 8-16.: AXI AL POWER : IMBALANCE Limits. O to 30 +10/-0 EFPD,

g: Three RC Pumps'-- Davis-Besse 1. Cycle 7

~

i gy .(Tech Spec Figure 3.2-2a)-

J :-

. 110

.. . 100

. - 90 l i

. (-16.5,77p . . 80 (11.2,77)

g<# a

-(-18.7,69) g.

a.

70 - < (11.2,69)

-f (-22.5,60) <

RESTRICTED w

. . 60 > (15,60 )

RESTRICTED REGION iE . - 50 REGION

(-22.5,45) 0 8 o (15,45)

E. - 40 m y . 30 -;

i E

-lJ GE e  :

mgg g- - 20 E5G 1; g- E8e g- 10 j

a.

I I i i i i i i I

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

l l g .

y a

.I

I f 3/4 2-3 00U M M 8-23

' 4 S.. y ,

y _.i s t. a F

i ^

g

- u 5 r 0

I e .

8

( R (

1

- - E ~ 7 4 i - (

0 2 RS 2 - .

2 ET GR 2 1 (

8 -

( TA

- 5 II 5 1 ThX

, OC ,

er1 3 l 4 NT 6 7 5 ceA 0 5 E 0 , .

heL

) D ) 6 7

. 9 SRP 7

A-X o 4

)

7 p CO e W I 2 I "wg 2i"

)

cPE L

A 0 S"'gC3 -

Fn ioI uR P - "S! g

'M 3 O W 0 1

l u r - A s.B .

/ E e - ' L 4 R

'0 mgu S. b!- . 3y *g nga 3DN A .

I - . - - - - - - aC

- 82 - - -

- - M - - - - - . - - 2vE B 4 ,

- i 23 A 1 2 3 5 6 7 8 1 1 2sL 4a L 0 0 0 0 0 0 0 0 0 0 1 b - i A 0 0 ) Bm

. 1 N 0 i

ei C . st ,

E ss .

( o > ( (

e,

% 2 1 1 R 1 3

) 0 i E ( 1 1

, 0 .

1 . .

RS ET ,5 2 2 C+

, GR y 1 II 6 6 7 c0 3 I 5 OC 0 9 7 l /

0 ) NT ) ) )

e -

E 0 D 7 t

B 4 o .

& 0 e W ' 7 5

5 +

M 0 a 1 0

/

0 g .

,D E

F P

- =l I I 5' 5g ' ag I , l b

- .  ! ,  ! t,  ; ;1 l i l l)J .

m- a g ..

$- Figure 8-18. AX1 AL POWER IMBALANCE Limits, 75 +10/-0 to 335 210 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 7-(Tech Spec Figure 3.2-2c) l:

. 110

. .100

- - 90

(-14.1,77)_ -- ,, (11.2,77)

I (-18.7,69)-

E g.

n.

. 70 < (11.2,69)

(-22.5,60) < ". . 60 i(15,60)

I.- RESTRICTED E REsTRiCTE0 REGION REGIM

'b.- 50

(-22.5,45)- o - g < p (15,45)

{.

40

[. . 30

p. . 20

$$n s' E 5 G.

I i i i E 85 E j i m

k~

~10 I i l i 1

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

I)

.B 1

3/4 2-3b EMM 8-25

i h -

[

[ Ei p 1

~

Figure 8-19. AXI AL~ POWER IMBALANCE Limits, After 335 i 10 EFPD, Three RC Pumps'-- Davis-Besse 1. Cycle 7 .

(Tech Spec Figure 3.2-2d)

'IL G ,

1

, . 110

. 100 J

. .90 .

(-15.7,77) _ . .80 (9,7,77)

" 70

(-18.7,69)- g-c.

< (11.2,69)

(-22.5,60)i a. .60 > (15,60) l RESTRICTED' E-w RESTRICTED l

j REGION E. .50 REGION .

(-22.5,45) o S O (15,45) 40

{.

w

,. 30 .  ;

22 5 s

$!C E gg Eg. .20 i l' $ES aca

~

u l

g. .10 l >

1 l':.

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

[

I I}

3/4 2-3c 8-26 B&W FuelCompany ll -

Y ,

j

^

Table 8-1. Quadrant Power Tilt Limits -

1 (Tech Soec Table 3.2-1) l st.ady-state.

Limit for steady-state Limit for 1

i Quadrant Power Tilt- Thermal Thermal Transient Maximm -I as measured by:- Power 5 60% Pewer 2 60% Limit Limit j 'l symetrical incore 6.83 4.31 '10.03 It I'

20.0 l detector system Power Range cilannels 4'05

. 1.96 6.96 20.0 Minimum incore detector 2.80* 1.90* 4.40* 20.0* 'l systm ,

<It i  !

-t l:

i i

• Assumes detector strings with >60% depletion are excluded frcn the miniu m

.-incore system configuration.

e  ;

s s

3/4 2-12 8-27 r

B&W FuelCompany L:ll

I 3/4.2. BJWER DISTRIRJfION LIMITS N ES Il The specifications of this section provide assurance of fuel integrity during g Condition I - (normal operation) and II (incidents of moderate frequency) g events by: (a) maintaining the minimam DiBR in the core 2 the mininnn allowable INB ratio during normal operation and during short term transients, a.

(b) maintaining the peak linear power density 5 18.4 kW/ft during normal' g[

operation, and (c) maintaining the peak power density less than the limits given in the Mess: to specification 2.1 during short term transients. In addition, the above criteria mist be met in order to meet the ====ntions E, used for the loss-of-coolant accidents. B The power imbalance envelope defined in Figures 3.2-1 ard 3.2-2 and the insertion limit curves, Figures 3.1-2 and 3.1-3 are Maewi on IOCA analyses 3 '.

which have defined the =vimm_ linear heat rate such that the =v4== clad g

tenperature will not exceed the Final Acceptance Criteria of 2200 F following j a IDCA. Operation outside of the power imbalance envelope alone does not  ;

constitute a situation that would cause the Final Acceptance Criteria to be  !

ex w should a IJOCA ooCur. The power imbalance envelope repfasents-the bourdary of operation limited by the Final Acceptance Criteria only if the B control rods are at the insertion limits, as defined by Figures 3.1-2 and

~

5' 3.1-3 and if the steady-state limit QUANANT POWER TILT exists. Additional conservatism is introduced by application of:

a. Nuclear uncertainty factors.

b. Thermal calibration uncertainty.

c. Fuel densification effects,

d. Hot rod manufacturing toleran factors. '

e. Potential fuei rod bow effects.

The ACTION statements which permit limited variations fran the basic -

requitawi:nt.s are amnied by additional restrictions which ensures that the original criteria are met.

The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:

F Q Nuclear heat flux bot channel factor, is defined as the maxinum local fuel rod linear power density divided by the average fuel rod linear per density, a===4ng normal fuel pellet and rod dimensions. ,

B 3/4 2-1 3 8-28 g-B&W FuelCompany l

g i

~LI: - mm-I --

i LI

b. 'Ihe measurement of enthalpy rise hot channel factor, I incraami by 5 percent to account for measurement error.

(H' I l For Condition II events, the core is protected frun exceeding the values given in the ha- to specification 2.1 locally, and frm . going below the

m- mininum allowable INB ratio by autmatic protection on power, AXIAL IWER I

g- IMBAIANCE pressure and tenparature. Only conditions 1 through 3, above, are l mandatory since the AXIAL POWER IMBAIANCE is an explicit input to the reactor protection system. ]

h QUANANT IWER TILT limit assures that the radial power distribution j

. satisfies the design values used in the power capability analysis. Radial- '

power distribution maastuair.nts are made during startup testing and L

L

[g . periodically during power operation. q l 'Ihe QUAIRANT POWER TILT limit at which ' corrective action is required provides l l INB and linear heat generation rate protection with x-y plane power tilts.

l In the event the tilt is not corrected, the margin for uncertainty on F is l

l reinstated by reducing the power by 2 percent for each percent of tile in ,

excess of the limit.

'3/4.2.5 INB PARAMETERS j

'Ibe limits on the INB related W assure that each of the parameters  !

are maintained within the normal steady- state envelope of operation assumed  ;

in the transient and accident analyses. 'Ibe' limits are consistent with the 1 1, ESAR initial a==tions and have been analytically desnated adequate to l maintain a mininum INBR greater than the minimum allowable INB ratio l throughout each analyzed transient.

'Ihe 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changas and other expected transient operation. h 18 month periodic meastu=irnt of the RCS total flow rate using delta P i

instrumentation is adequate to detect' flow degradation ard ensure correlation l of the flow indication channels with measured flow such that the indicated 1 g percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l3 basis.

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. B 3/4 2-3 8-29 B&W FuelCcapany Ll t

J Iq 3/4.4 REACIOR CDOIANT SYS'ID4 'I i 3/4.4.1 REACIOR 000IANP I. OOPS I

! me plant is designed to operate with both reactor coolant loops in operation, and maintain DER above the minimm allowable DiB ratio during all l normal operations and anticipated transients. With one reactor coolant punp not in operation in one loop, DIERMAL POWER is restricted by the Nuclear Overpower had on RCS Flow and AXIAL FOWER IMBAIANCE, ensuring that the DER will be maintained above the minimm allowable DE ratio at the maximum l possible DIERMAL POWER for the nunber of reactor coolant punps in operation 3 or the local quality at the point of minbaum DNBR equal to the D4B corre- l 3 lation quality limit, whichever is more restrictive.

In PODE 3 when RCS pressure or tenperature is hi@c than the decay heat i removal system's design condition (i.e. 330 psig and 350 F), a single reactor

'~

coolant loop provides sufficient heat removal capability. Se remainder of PODE 3 as well as in PODES 4 arx15 either a single reactor coolant loop or a E' DIR loop will be. sufficient for decay heat removal; but single failure 5 considerations require that at least two loops be OPERABLE. Bus, if the reactor coolant loops are not OPERABIE, this specification requires two DIR loops to be OPERABII.

' Natural circulation flow or the operation of one DIR punp provides adequate _

flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

%e reactivity change rate associated with boron reduction will, therefore, be within the capacity of operator recognition and m d w l.

3/4.4.2 and 3/4.4.3 SAFEIY VALVES

• me pressurizer code safety valves operate to prevent the RCS frcan being pressurized above its Safety Limit of 2750 psig. Each safety valve is designed to relieve 336,000 lbs per hour of saturated steam at the valve's setpoint.

me relief capacity of a single safety valve is adaqute to relieve any overpressure condition which could occur during shutdown. In the event that g no safety valves are OPERABIE, an operating DIR loop, ww:cted to the RCS,- 3 provides overpressure relief capability and will prevent RCS overpressurization. During operation, all pressurizer code safety valves must be OPERABIE to prevent the RCS frun being pressurized above its safety limit of 2750 psig. S e combined relief capacity of all of these valves is greater than the maximm surge rate resulting from any transient. i

%e relief capacity of the decay heat renoval system relief valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that this relief valve is not OPERABIE, reactor coolant system pressure, pressurizer level and make up water inventory is limited and the capability of the high pressure injection system to  ;

B 3/4 4-1 3 8-30 g1 B&W FuelCompany l

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9. STARIUP PROGRAM - PHYSICS TESTING

'Ihe planned startup test program associated with core perfonnance is outlined

'below. 'Ibese tests verify that oore performance is within the a===ntions of

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the safety analysis and provide infonnation for continued safe operation of the unit.

9.1. Precritical 'IV.sts

[ 9.1.1. Control Rod TriD Test Precritical control rod drop times are recorded for all O.ahcl rods at hot

, full-flow conditions before zero power physics testing begins.- Acceptance criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.58 se.uds at the conditions above.

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

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L 9.1.2. RC Flow 1.

Reactor coolant flow with four RC pumps running will be measured at hot

! standby conditions. AcceW criteria require that the measured flow be LM~ within allowable limits.

1 9.2. Zero Power Physics Tests 9.2.1. Critical Boron Corr.Em bation

once initial criticality is achieved, equilibrium boren is obtained and the critical baron concentration determined. 'Ihe critical boron wishation

-is calculated by correctirg for any rod withdrawal required to achieve the all rods out equilibrium boron. 'Ihe acceptance criterion placed on critical boron concentration is that the actual boron evic.eanution must be within 100 ppm boron of the predicted value.

I. 9-1 g B&W FuelCompany

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9.2.2. Tennerature Reactivity Ocefficient

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'Ibe isothermal HZP taperature coefficient is measured at approximately the all-rods-cut configuration. During charges in tenperature, reactivity faarhek ney be w p ssated by control rod movement. 'Ibe change in -

reactivity is then calculated by the sumation of reactivity associated with the tenperature change. Acceptance criteria state that the measured value shall not differ fra the predicted value by more than i 0.4x10-2% ok/k/0F. s

- Ibe moderator coefficient of reactivity is calculated in conjunction with the tenperature coefficient measurement. After the tenperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is subtracted to obtain the moderator coefficient. 'Ihis value must not be in excess of the acceptance criteria limit of +0.9x10~2% Ak/k/0F.

1 9.2.3. Control Rod Grouo/ Boron Reactivity Worth Iniividual control rod group reactivity worths (groups 5, 6, and 7) are measured at hot zero power conditions using the bororVrod swap method. 'Ihis technique consists of deborating the reactor coolant system and coripensating L

for the reactivity changes frun this deboration by inserting individual control red groups 7, 6, and 5 in irc = = ital steps. 'Ihe reactivity changes that occur during these measurements are calculated based an reactimeter data, and differential red worths are obtained from the measured reactivity worth versus the change in rod group position. 'Ibe differential rod worths l of each of- the controlling groups are then s-a to obtain integral rod group worths. 'Ihe acceptance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

Dredicted value - measured value M ue x 100 515

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

l D edicted value - measured _value x 100 < 10 E measured value -

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B&W FuelConipany  !

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('Ibe borcri reactivity worth (differential boron worth) is measured by dividing .

' the total inserted rod wcrth by the borun change made for the rod worth test. ,

'Ihe acceptance criterion for measured diffeInntial borun Worth is as follows: [

1. oredicted value - measured value measured value x 100 5 15

'Ibe predicted rod worths and differential boren vorth are taken from the PIM. "

'9.3. Power Escalation Tests , 9.3.1. Cbre Synr@rv Test

, 'Ihe purpoise of this test is to evaluate the symmetry of the core at low power

-during the initial power escalation following a refueling. Symmetry evaluation is based on incore' quadrant power tilts during escalation to the intamadiate power level. 'Ibe core symetry is acceptable if the absolute ,

values of the quadrant power tilts are less than the error adjusted alarm '

limit.

9.3.2. Core Power Distribution Verification at Intermediate Power Invel *

(IPL) and 100% FP With &=inal Control Rod Position Core power distribution tests are performed at the IPL and .100% full power

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?(FP). . ' Equilibrium xenan is established prior to both.the IPL and 100% FP tests. 'Ihe test at the IPL is essentially a check of the power distribution in the core to identify any abnormelities before escalating to the 100% FP

- y plateau. Peaking factor criteria are applied to the. IPL oore power distribution results to determine if ^ additional tests or analyses are-required prior to 100% FP cperation.

.Ihe following acceptance criteria are placed on the IPL and 100% FP tests:

1. The -4== IHR must be less than the IDC* limit.

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I 2.

'Ihe minimum INBR mist be greater than the initial cordition INBR limit.

l 3.- 'Ibe value obtained fra extrapolation of the minimem INBR to the next l

power plateau overpower trip setpoint nist be c.3ater than the calculated 112% FP INBR value, or theextrapolatedvaluhof imbalance alst fall outside the RPS power / imbalance / flow trip envelope. .

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4. The value obtained frm extrapolation of the worst-case max 11pm IHR to the next power plateau overpower trip setpoint must be less than the 9-3

l. B&W FuelCompany l

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l fuel melt : limit, or the extrapolated value of imbalance mst fall-

- outside' the RPS power / imbalance / flow trip erwelope.

5. . 'Ibe quadrant power tilt shall not eM the limits. specified in the 5-Todinical Specifications. .

6. 'Ibe highest measured and predicted radial peaks shall be within the following limits: q

" " " x 100 more positive than -5 l

7. 'Ihe highest measured ard predicted total peaks shall be within the I" followirs limits:

omdW nlm - measumd ulue x 100 more positive than -7.5 measured value j

= Items 1, 2, and 5 ensure that the initial condition IOCA, 3nitial condition ~ -

DNBR, and quadrant power tilt limits respectively are maintained at the IPL l

and 100% FP.

Items 3 and 4 establish the criteria whereby arMation to full power may be acconplished without avMing the safety limits Jpecified by the safety l analysis t ~ gard to INBR and linear heat rate. .

l Items 6 at m established to determine if measured and predicted power '

distributions are consistent.

9.3.3. Incore Vs. Fxcore Detector Ita'.ance Correlation Verification at the IPL 1 Imbalances, set up in the core by control rod positioning, are read simultaneously on the incore detectors and excore pw range detectors. 'Ibe ;l excore detector offset versus incore detector offset slope must be greater l than 0.96. If this criterion is not met, gain amplifiers on the excore -l detector signal prmaaaing equipnent are adjusted to provide the required slope. I 9.3.4. Tencerature Reactivity Coefficient at =101%,_E

'Ihe average reactor coolant tenperature is decreased and then increased at .(

constant reactor power. 'Ihe reactivity associated with each temperature change is obtained from the change in the controlling rod group position.

Controlling rod group worth is measured by the fast insert / withdraw method.

9-4 saw Fuelcompany l

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I he temperature reactivity coefficient is calculated from the measured charges in reactivity ard tamparature. Acomptance criteria state that the moderator temperature coefficient shall be negative.

9.3.5. Power Doooler Reactivity Ocefficient at =100% FP The power Doppler reactivity ocefficient is calculated frm data recorded during control rod worth measurements at power usirg the fast insert / withdraw method.

S e fuel Doppler reactivity coefficient is calculated in conjunction with the power Dgpler coefficient measurinant. De pcur Doppler coefficient as measured above is alltiplied by a precalculated conversion factor to obtain the fuel Dappler coefficient, h is mansared fal Doppler coefficient m st be more negative than the acceptance criteria limit of -0.90 x 10'3% WOF.

9.4. Pmoedure for Use if Accootance criteria Not Met If acceptance criteria for any test are not met, an evaluation is performed before the test program is continued. mis evaluation is performed by site test personnel with participation by B&W Nuclear %chnologies technical personnel as required. nirther specific actions depend on evaluation results. These aedons can include repeating the tests with more detailed test prerequisites and/or steps, added tests to seard for anomalies, or design personnel perfornirg detailed analyses of potential safety problems because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be canprmised by such escalation.

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I I 9-5 l B&W FedCompeg

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-10. REITRDCES

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

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

3. Rancho Seco Cycle 7 Aeload Report - Volume 1 - Mark BZ Fuel Assembly j Design Report, BAW-1781P, April 1983.

4. Rancho Seco Nuclear Generating Station - Evaluation of Mark BZ Fuel j Assembly - Design, U. S. Nuclear Regulatory Ocnniassion, Washington, D. >

C., November 16, 1984.

5. hupeuu to Determine In-Reactor Performance of B&W Ibels - Cladding I Creep Collapse, BAW-10084P. Rev. 2, hWk arrt Wilcox, Dfnchburg, VA,  ;

IL October 1978.  ;

6. latter, J.H. Taylor (B&W) to C.O. 'Dxanaa (NRC),

Subject:

Creep Collapse f Analysis for B&W Ibel, JHT/86-011A, Dated January 31, 1986.

7. Imtter, Dennis M. Crutchfield (NRC) to J.H. Taylor (B&W), Subjects I

Acceptance for Referencing of a Special Licensing Report, Dated December '

5, 1986.

8. TACO 2: Fuel Performance Analysis, BAW-10141P-A, Rev. 1, . hWk &

Wilcox,16'nchburg, Virginia, June 1983.

9. N00 DIE - A Multi-Dimensional Two-Groqp Reactor Simulator, BAW-10152A, hWk & Wilcox, Lynchburg, Virginia, June 1985.

10. LYNXP Core Transient 'Ibermal-Hydraulic hwnun, BAW-10156-A, February 1986.

11. Davis-Besse Nuclear Ibwer Station Unit 1, Cycle 6 - Reload Report, DE-29.28, April 1988. ]

12. 'Ibermal-Hydraulic. Crossflow Applications, BAW-1829, April 1984. )

ll 13. BWC Correlation of Crtitical Heat Heat Flux, BAW-10143P-A, hWk &

!5 Wilcox, Lynchburg, Virginia, April 1985. >

l l I 14. Davis-ham Unit 1 Fuel Densification Report, BAW-1401, hWk & Wilcox, Lynchburg, Virginia, April 1975.

I 10-1 l

Lg B&W FudCompet c .

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15. NW's 2005 Evaluation Model. BhW-10104P, Bec. _5, mWA & . Wi1eox, )

Iynctiburg, Viginia, April 1986. )

16. Ects Evaluatice of NW's 177-rA Raised-Incy NSS, BAW-10105, Rev. 1, hwk & Wiloax, Iynciturg, Viginia, July 1975.

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10. RE:FERDGS ]

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

2. Davis-Besse Nuclear Power Station No.1, Updated Es.f4ty Arudysis Report, Docket No. 50-346.

3. Rancho Seco Cycle 7 Raload Report - Volume 1 - !@uk E .h1 Asseqbly Design Report, BAW-1781P, April 1983.  ;

4. Rancho Seco Nuclear Generating Station - Evaluaties of Mark E3 Fuel Assembly Design, U. S. Nuclear Regulatory ocanniconion, **bAinson, D.

C., Ncwenbar 16, 1984.

5. Ps %uun to Determine In-Reactor Performance of B&W nerls - Cladding Creep Collapse, BhW-10084P,.. Rev. 2, hWk ani Wilecw, IapKhot:rg, VA, October 1978.

6. Istter, J.H. Taylor (B&W) to C.O. 'thcanas (NRC),

Subject:

Ctw1p Collapse

, Analysis for B&W Fbel, JHT/86-011A, Dated January 31, it%

s

7. Ietter, Dennis M. Crutchfield (NRC) to J.H. Tayler (N#), subject:

Acceptance for Referencing of a Special Licensing Rg:c::t, Dated December  ;

5, 1986.

8. TA002: Fuel Performance Analysis, BAW-10141P-A. AsG h*k&

Wilcox, Dynchburg, Virginia, June 1983. '

9. N00DT.E - A Multi-Dimensional 'IWo Grup Reactor Simulator, BAW-10152A, h*k & Wilcox, Dfnchburg, Virginia, June 1985.

10. IXNXT Oore Transient '!hermal-}tfdraulic Fwam, BAW-10156-A, Pinbruary 1986.

11. Davis-Besse Nuclear Power Station Unit 1, Cycle 6 - Reload Report, BMf- r 2028, April'1988.

12. ' thermal-Hydraulic Crcssflow Applications, BAW-1829, April 1984.

13. BWC Correlation of Crtitical Heat Heat Flux, BAW-10143P-A, h*k&

Wilocx, Dfnchburg, Virginia, April 1985.

14. Davis-Basse Unit 1 Pbel Densification Report, BAW-1401, h

• k & Wilcox, Dfnchburg, Virginia, April.1975.

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15. N W's EOCS Evaluation W del. BhW-10104P, Rev. 5, h W k & Wi1cox, .

! O 14'nctiburg, Virginia, April 1986.

16. ECCS Evaluation of N W's 177-TA Raimmi-Icop NSS, BAW-10105, nev. 1, j hWk & Wilcox, Lynchburg, Virginia, July 1975. i

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