ML20090J256

Cycle 5 - Reload Rept

ML20090J256
Person / Time
Site:	Davis Besse
Issue date:	04/30/1984
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20090J254	List: ML20090J252 ML20090J256
References
BAW-1827, TAC-55531, NUDOCS 8407300258
Download: ML20090J256 (77)

Text

..

BAM-1827 April 1984 l

DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 5 . RELOAD REPORT 1

1 r

1 I

go pgg7p'a*88?jgg Babcock &Wilcox P a McDermott company

l BAW-1827 l

April 1984 i

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.i DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 5 -- RELOAD REPORT 1

J 4

BABC0CK & WILC0X Utility Power Generation Division

- P. O. Box 1260 Lynchburg, Virginia 24505 t

Babcock &WHcom a McDermott company

I CONTENTS Page

- 1. INTRODUCTION AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . . 1-1

2. OPERATING ;!ISTORY ,....................... 2-1 ,

1 3.- GENERAL DESCRIPTION ....................... 3-1 i

4. FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Fuel Assembly Mechanical Design .............. 4-1 4.2. Fuel Rod Design ...................... 4-1 4.2.1. Cladding Collapse ................. 4-2 4.2.2. Cladding Stress .................. 4-2 4.2.3. Cladding Strain .................. 4-2 4.3. Th e rmal Des i g n . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4. Material Compatibili ty . . . . . . . . . . . . . . . . . . . 4-3 4.5. Ope rati ng Experience . . . . . . . . . . . . . . . . . . . . 4-3

5. NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Physics Characteristics .................. 5-1 5.2. Changes in Nuclear Design ................. 5-2

6. THERMAL-HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . 6-1

7. ACCIDENT AND TRANSIENT ANALYSIS ................. 7-1 7.1. General Safety Analysis .................. 7-1 7.2. Accident Evaluation .................... 7-1

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . . . . . . 8-1

9. STARTUP PROGRAM - PHYSICS TESTING ................ 9-1

-9.1. Precritical Tests ................'..... 9-1 9.1.1. Control Rod Trip Test ............... 9-1 9.1.2. Reactor Coolant Fl ow . . . . . . . . . . . . . . . . 9-1 9.2. Zero Power Physics Tests . . . . . . . . . . . . . . . . . . 9-2 9.2.1. Critical Boron Concentration . . . . . . . . . . . . 9-2 9.2.2. ~ Temperature Reactivity Coefficient . . . . . . . . . 9-2 9.2.3. Control Rod Group Reactivity Worth . . . . . . . . . 9-2 9.2.4. Ejected Control Rod Reactivity Worth . . . . . . . . 9-3

- iii - NMM a McDermott company

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CONTENTS (Cont'd)

Page 9.3. Power Escal ation' Tests . . . . . . . . . . . . . . . . . . . 9-3 9.3.1. Core Power Distribution Verification ats 40, s75, and S100 FP With Nominal Control Rod Position . . . 9-3

.9.3.2. Incore Versus Excore Detector Imbalance j Correlation Verification ats40% FP ........ 9-5 9.3.3. Temperature Reactivity Coefficient at s100% FP . . . 9-5 9.3.4. Power Doppler Reactivity Coefficient at s100% FP . . 9-5 9.4. Procedure for Use When Acceptance Criteria Are Not Met . . . 9-6 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 List of Tables Table 4-1. Fuel Des i gn Pa r ame te rs - . . . . . . . . . . . . . . . . . . . . . 4-4 4-2. Fuel Thermal Analysi s Parameters . . . . . . . . . . . . . . . . 4-5 5-1. Davis-Besse Unit 1, Cycle 5 Physics Parameters . . . . . . . . . 5-3 5-2. Shutdown Margin Calculation for Davis-Besse Unit 1, Cycle 5 . . 5-5 1 6-1.. Davis-Besse Cycles 4 and 5 Thermal-Hydraulic Design Conditions ....................... 6-2 7-1. - Comparison of Key Parameters for Accident Analysis . . . . . . . 7-3 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates . . . 7-3 8-1. Reactor Protection System Instrumentation Trip Setpoints . . . . 8-2 8-2.- Quadrant Power Til t Limits . . . . . . . . . . . . . . . . . . . 8-8 List of Figures Figure 3-1. Davis-Besse Cycle 5 Full Core Loading Diagram ......... 3-3 3-2. Enrichment and Burnup Distribution for Davis-Besse Unit 1, Cycle 5 ............................ 3-4

3-3. Control Rod Locations for Davis-Besse Unit 1, Cycle 5 . . . . . 3-5 3-4. Davis-Besse Cycle 5 BPRA Enrichment and Distribution . . . . . . 3-6 4-1 Mark B5 Upper End Fitting ................... 4-6 4-2. Hol ddown Spring Retai ner . . . . . . . . . . . . . . . . . . . . . 4-7 4-3. Mark-B5 BPRA Spider ...................... 4-8 4-4. Mark-85 BPRA Spider / Upper End Fitting / Reactor

,. , Internals Interaction ..................... 4-9 5-1. BOC (4 EFPD), Cycle 5 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, APSRs Inserted . . 5-6

(

,g, M M & M COE a McDermott company t

.c___-,__ __, _ _ _ _ _ _ _ . _ _ . _ . _ _ .. _ _ _ . - . . . _ . _ _ _ _ _ _ _ - , ~

l Figures (Cont'd)

Figure Page 8-1. Reactor Core Safety Limit . . . . . .............. 8-10 8-2. Trip Setpoint for Flux - AFlux/ Flow . . . . . . . . . . . . . . 8-11 8-3. Regulating Group Position Limits, O to 25+10/-0 EFPD, Four l RC Pumps - Davis-Besse 1, Cycle 5 . . . . . . . . . . . . . . . 8-12 8-4. Regulating Group Position Limits, 25+10/-0 to 200t10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 5 ............ 8-13 8-5. Regulating Group Position Limits, 200 10 to 330 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 5 ......... 8-14 8-6. Regulating Group Position Limits, 330 10 to 390 10 EFPD, Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 5 . . . 8-15 8-7. Regulating Group Position Limits, O to 25+10/-0 EFPD, Three RC Pumps - Davi s-Besse 1, Cycle 5 . . . . . . . . . . . . 8-16 8-8. Regulating Group Position Limits, 25+10/-0 to 200 10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycle 5 . . . . . . . . . . . . 8-17 8-9. Regulating Group Position Limits, 200 10 to 330 !10 EFPD, Three RC Pumps - Davis-Bes se 1, Cycle 5 . . . . . . . . . . . . 8-18 8-10. Regulating Group Position Limits, 330 10 to 390 t10 EFPD, Three RC Pumps, APSRs Withdrawn - Davis- Besse 1, Cycle 5 . . . 8-19 8-11. APSR Position Limits, O to 25+10/-0 EFPD, Four RC Pumps -

Davis-Besse 1, Cycle 5 ...... .............. 8-20 8-12. APSR Position Limits, 25+10/-0 to 200 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 5 ................ 8-21 8-13. APSR Position Limits, 200 t10 to 330 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 5 . . . . . . . . . . . . . . . 8-22

, 8-14. APSR Position Limits, 330 t10 to 390 10 EFPD, Three or Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 5 . . . . 8-23 j 8-15. APSR Position Limits, 0 to 25+10/-0 EFPD, Three RC Pumps -

! Davis-Besse 1, Cycle 5 ....... ............. 8-24 l 8-16. APSR Position Limits, 25+10/-0 to 200 10 EFPD, Three RC l Pumps - Davis-Besse 1, Cycle 5 ................ 8-25 8-17. APSR Position Limits, 200 t10 to 330 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 5 ................ 8-26 8-18. Axial Power Imbalance Limits, 0 to 25+10/-0 EFPD, Four RC Pumps - Davi s-Besse 1, Cycl e 5 . . . . . . . . . . . . . . . 8-27 8-19. Axial Power Imbalance Limits, 25+10/-0 to 200 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 5 ............ 8-28 8-20. Axial Power Imbalance Limits, 200 10 to 330 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 5 ............ 8-29 8-21. Axial Power Imbalance Limits, 330 10 to 390 10 EFPD, Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 5 . . . . 8-30 8-22. Axial Power Imbalance Limits, 0 to 25+10/-0 EFPD, Three RC Pumps - Davi s-Besse 1, Cycle 5 . . . . . . . . . . . . 8-31 8-23. Axial Power Imbalance Limits, 25+10/-0 to 200 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 5 . . . . . . . . . . . . 8-32

, -v- Babcock &WHcom a McDermott company

l l

Figures (Cont'd)

Figure Page 8-24. Axial Power Imbalance Limits, 200 10 to 330 10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 5 . . . . . . . . . . . . . . . 8-33 I 8-25. Axial Power Imbalance Limits, 330 10 to 390 i10 EFPD, Three RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 5 . . . 8-34 l 8-26. Control Rod Core Locations and Group Assignments -

Davis-Besse 1, Cycle 5 ................. . . . 8-35

- yi -

Babcock & WIIcox.

a McDermott company

1. INTRODUCTION AND

SUMMARY

This report justifies operation of the Davis-Besse Nuclear Power Station Ur.i t 1 at the rated core power of 2772 MWt for cycle 5. The required analyses are included as outlined in the Nuclear Regulatory Commission (NRC) document, " Guidance for Proposed License Amendments Relating to Re-fueling," June 1975. This report utilizes the analytical techniques and design bases documented in several reports that have been submitted to the NRC and approved by that agency.

Cycle 5 reactor and fuel parameters related to power capability are summa-rized in this report and compared to cycle 4. All accidents analyzed in the Davis-Besse Final Safety Analysis Reportl (FSAR) have been reviewed for cycle 5 operation, and in cases where cycle 5 characteristics were conserva-tive compared to cycle 1, no new analyses were performed.

Retainers2 and neutron sources will remain in the core. The effects on con-tinued ytration without orifice rod assemblies (0 ras) and with the re-tainers have been accounted for in the analysis performed for cycle 5.

The Technical Specifications have been reviewed and modified where required for cycle 5 operation. Based on the analyses performed, taking into ac-count the emergency core cooling system (ECCS) Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Davis-Besse Unit 1, cycle 5 can be operated safely at its licensed core power level of 2772 MWt.

i 1-1 Babcock &Wilcos a McDermott company

2. OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Davis-Besse Unit 1 is the currently operating cycle 4, which achieved criti-

. cality on September 27, 1983. Power escalation began on September 29, 1983 and full power (2772 MWt) was reached on November 2,1983. During cycle 4 operation, no operating anomalies occurred that would adversely affect fuel performance during cycle 5. The scheduled durations of cycles 4 and 5 are 280 and 390 effective full power days (EFPD), respectively.

The APSRs were pulled at 200 EFPD to increase the lifetime of cycle 4. The APSR pull coupled with a power coastdown resulted in a potential cycle 4 length of approximately 280 EFPD. The cycle 5 design also includes an APSR pull and power coastdown.

The cycle 5 design minimizes the number of fuel assemblies that are cross core shuffled to reduce the potential for quadrant tilt amplification. The cycle 5 shuffle pattern is discussed in section 3.

. 2-1 "# E EU8 a McDermott company

3. GENERAL DESCRIPTION The Davis-Besse Unit I reactor core is described in datail in chapter 4 of 1

the FSAR for the unit. The cycle 5 core consists of 177 fuel assemblies (FAs), each of which is a 15x15 array containing 208 fuel rods,16 control rod guide' tubes, and one incore instrument guide tube. All FAs in batches 4, 5, 6, and 7 have a constant nominal fuel loading of 468.25 kg of urani-um . - The batch 1E assembly has a fuel loading of 472.24 kg of uranium. The fuel consists of dished-end cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy 4. The undensified nominal active fuel lengths, theo-retical densities, fuel and fuel rod dimensions, and other related fuel parameters may be found in Tables 4-1 and 4-2 of this report.

Figure 3-1 is the core loading diagram for Davis-Besse Unit 1, cycle 5.

. Seventeen batch 1D assemblies, 20 batch 28 assemblies, and 28 batch 4A as-semblies will be discharged at the end of cycle 4. The fuel assemblies in batches 48, 5A, . 5B, and 6 will be shuffled to their cycle 5 locations.

Batches 4B and 5A have an initial uranium-235 enrichment of 3.04 wt %.

Batches 58 and 6 have an initial enrichment of 2.99 wt %. One batch IE as-sembly with an initial enrichment of 1.98 wt % will be reinserted in cycle

5. A feed batch . consisting of 64 batch 7 assemblies with uranium enrich-ment of 3.19 wt % will be inserted in the core interior. Batch 6 will oc-cupy the . core periphery. Figure 3-2 is a _ quarter-core map showing each assembly's burnup at the beginning of cycle (B0C) 5 and its initial enrich-

ment.

- Cycle 5 is operated in a feed-and-bleed mode. The core reactivity is con-trolled by 53 full-length Ag-In-Cd control rod assemblies (CRAs), 64 burn-able poison rod assemblies (BPRAs), and soluble boron. In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The cycle e

3-1 hWha a MCDermott comparty

5 locations of the 61 control rods and the group designations are indicated in Figure 3-3. The core locations of 61 control rods for cycle 5 are iden-tical to those of reference cycle 4. However, the cycle 5 rod group desig-nations differ from the cycle 4 designations. The cycle 5 locations and enrichments of the BPRAs are shown in Figure 3-4.

3-2 N MIcom a McDermott company

Figure 3-1. Davis-Besse Cycle 5 Full Core loading Diagram FUEL TRANSFER CANAL--->

X 6 6 6 6 6 A _

M2 K1 R8 K15 M14 6 6 6 7 5B 7 6 6 6 B L2 L1 N2 F C9 F N14 L15 L14 6 7 5A 7 SB 7 58 7 SA 7 6 I C 013 F E7 F C4 F C12 F E9 F 03 6 7 SB 7 4B 7 SB 7 48 7 58 7 6 D B10 F G7 F M4 F B8 F M12 F G9 F B6 6 5A 7 5B 7 48 7 48 7 SB 7 5A 6 E

A10 G5 F E5 F L5 F Lil F E11 F Gil A6 6 6 7 48 7 SB 7 SB 7 5B 7 4B 7 6 6 F 811 912 F D11 F C7 F D8 F G13 F 5 F B4 85 6 7 B5 7 4B 7 5B 5B 58 7 4B 7 5B 7 6 G A9 F D3 F E10 F D4 L8 D12 F E6 F D13 F A7 6 SB 7 SB 7 SB SB 5B SB 7 SB 7 58 6 W-H - H6 H1 H15 G3 F H2 F H4 H10 e H12 F H14 F K13 6 7 SB 7 4B 7 5B 5B SB 7 48 7 SB 7 6 K

R9 F N3 F M10 F N4 F8 N12 F M6 F N13 F R7 6 6 7 48 7 SB 7 5B 7 5B 7 4B 7 6 6 L P11 P12 F N11 F K3 F N8 F 09 F N5 F P4 P5 6 5A 7 SB 7 4B 7 48 7 5B 7 5A 6 M RIO K5 F M5 F F5 F Fil F M11 F K11 R6 N

6 7 5B 7 48 7 5B 7 48 7 SB 7 6 P10 F K7 F E4 F P8 F E12 F K9 F P6 6 7 5A 7 SB 7 58 7 SA 7 6 0

C13 F M7 F 04 F 012 F M9 F C3 p 6 6 6 7 5B 7 6 6 6 F2 F1 02 F 07 F D14 FIS F14 6 6 6 6 6 R E2 G1 A8 GIS E14 I

Z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Batch Cycle 4 Location Cy 1 = reinserted from cycle 1 3-3

c' 1

Figure 3-2. Enrichment and Burnup Distribution for Davis-Besse 1, Cycle 5 8 9 10 11 12 13 14 15 1.98 2.99 2.99 3.19 2.99 3.19 2.99 2.99 H

12644 17710 22063 0 18631 0 20249 8600 2.99 2.99 3.19 3.04 3.19 2.99 3.19 2.99 K

17738 18518 0 25725 0 18235 0 8250 2.99 3.19 2.99 3.19 3.04 3.19 2.99 2.99 L

22055 0 20211 0 26978 0 6450 8993 3.19 3.04 3.19 2.99 3.19 3.04 2.99 M

0 25676 0 18367 0 17637 6904 2.99 3.19 3.04 3.19 2.99 3.19 2.99

~

N 18631 0 26956 0 17133 0 10719 3.19 2.99 3.19 3.04 3,19 2.99 0

0 18252 0 17622 0 7216 2.99 3.19 2.99 2.99 2.99 P

20249 0 6460 6905 10720 2.99 2.99 2.99 R

8602 8251 9000 x.xx Initial Enrichment xx,xxx 80C Burnup, mwd /mtU 3-4

Figure 3-3. Control Rod Locations for Davis-Besse 1, Cycle 5 X

l A .

B 3 7 3 C 2 6 6 2 0 7 8 5 8 7 E 2 5 5 2 F' 3 8 1 7 1 8 3 G 6 4 4 6 N- 7 '5 7 1 7 5 7 -Y H

K 6 4 4 6 L 3 8 1 7 1 8 3 M 2 5 5 2 N 7- 8 5 8 7 0 2 6 6 2 P 3 7 3 R

l Z

1 2 3 4 9 6 7 8 9 10 11 12 13 14 15 No. of Group rods Functions 1 5 Safety 2 8 Safety X Group Number 3 8 Safety 4 4 Safety 5 8 Control 6 8 Control

~7 12 Control 8 8' APSRs 3-5 Total # ET-

1 4$-

Figure 3-4. Davis-Besse Cycle 5 BPRA Enrichment and Distribution 8 9 10 11 12 13- 14 15 H 0.8 1.1

^

K 0.8 1.1 0.5 L 0.8 0.8 0.5 M 0.8 0.8 0.8 N 1.1 0.8 0.2 0 1.1 0.5 0.2 P 0.5 R

l t

X.X BPRA Concentration, wt% 4B C in Al23 0*

3-6

4. FUEL SYSTEM DESIGN 4.1. Fuel Assembly Mechanical Design The types of fuel assemblies (FAs) and pertinent fuel parameters for Davis-Besse 1, cycle 5 are listed in Table 4-1. The Batch 7 FAs are the MK-85 design, while the other reinserted FAs are the Mk-B4 design. The Mk-B5 FAs are -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) hol ddown. Retainer assemblies will be used on two FAs (Mk-B4 design) that contain the regenerative neutron sources. The justifi-cation for -the design and use of the retainer assemblies is described in references 2 and 3.

Sixty-four BPRAs will be used with the sixty-four Batch 7 FAs. The reten-tion mechanism for the Mk-B5 design BPRAs is built into each assembly and retainer _ assemblies are not required.

The Mk-B5 upper end fitting (Figure 4-1) provides four open sl ots that align and guide the movement of the h:1ddown spring, spring retainer (Fig-ure 4-2), and the new Mk-B5 BPRA spider (Figure 4-3). The holddown spring is preloaded' through stop pins welded to ears on each side of the upper end fitting. Incore, as shown in Figure 4-4, the BPRA spider feet are captured between the holddown spring retainer and the upper grid pads on the reactor internals. This arrangement retains the fixed control components at all de-sign flow conditions.. The Mk-85 design also contains a redesigned holddown spring made from Inconel 718 material which provides added margin over the R-B4 spring design. The Mk-B5 upper end fitting has been tested extensive-ly, both in air and in .over 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of simulated reactor environment, to determine analytical input and to assure good incore performance.

4.2. Fuel Rod Design The fuel rod design and mechanical evaluation are discussed below.

4-1 m g, a McDermott company

4.2.1. Cladding Collapse i

The power histories were reviewed for each fuel assembly in cycle 5. For <

l each of the six batches the most limiting assembly power history was deter-  !

mined. The ' most limiting assembly is the one with the highest burnup.

These six power histories were compared to a generic analysis, or were used )

as input to a creep collapse analysis to ensure that creep ovalization will not affect the fuel performance during Davis-Besse 1 cycle 5. Both specif-ic and generic creep ova:ization analyses are based on reference 4 and are applicable to the designs for batches 1E, 48, 5A, 5B, 6, and 7.

Creep collapse analyses predict collapse times longer than 35,000 EFPH.

The longest incore exposure time for cycle 5 is 29,249 EFPH for batch 4B (Table 4-1).

4.2.2. Cladding Stress The Davis-Besse Unit 1, cycle 5 stress parameters are enveloped by a

~

conservative fuel rod stress analysis. The methods used for the analysis of cycle 5 have been used in the previous cycles.

4.2.3. Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding plastic ten-sile circumferential strain. The pellet is designed to ensure that plastic 4 cladding strain is less than 1% at design local pellet burnup and heat gen-eration rate. The design values are higher than the worst-case values the Davis-Besse Unit 1, cycle 5 fuel is expected to see. The strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density, and the lower tolerance for the cladding inside diameter (ID).

4.3. Thermal Design i All fuel in the cycle 5 core is thermally similar. The analyses for the i-incoming batch 7 fuel and for the reinserted fuel from batches 58 and 6 have been perfomed with the TAC 02 5 code using the analysis methodology described in reference 6. This methodology uses nominal undensified input I parameters provided in Table 4-2. Densification effects are accounted for in . the TACO 2 densification model. Reinserted fuel from batches 1E, 4B, and 5A was evaluated using TAFY3 analyses performed for prior cycles.

Babcock &Wilcon a McDermott company 5

The thermal design evaluation for the cycle 5 core is summarized in Table 4-2. Linear heat rate (LHR) capabilities are based on centerline fuel melt (CFM) with core protection limits based on a 20.4 kW/ft LHR to CFM. The TAC 02 analyses performed for batches 5B, 6, and 7 demonstrate that 20.5 kW/ft is the CFM limit for this fuel. Using TAFY3, the fuel internal pres-sure has been evaluated for the highest burnup fuel rod and is predicted to be less than the nominal reactor coolant system pressure of 2200 psia. The maxisium burnup of any fuel rod during cycle 5 is less than 45,000 mwd /mtU.

4.4. Material Compatibility The . compatibility of all possible fuel-cladding-coolant-assembly interac-tions for batch 7 FAs is identical to that of Present fuel.

4.5. Operating Experience Operating experience with the Mark-B 15x15 FA has verified the adequacy of its design. As of January 15,.1984, the following experience has been ac-cumulated for eight Babcock & Wilcox (B&W) 177-FA plants using the Mark-B FA:

Max FA burnup,(a) mwd /mtU um a Current ec r c Reactor cycle Incore Discharged output, b) MWh Oconee 1 8 31,629 50,598 47,528,595 Oconee 2 7 24,221 36,800 42,405,541 Oconee 3 7 33,943 35,463 43,931,451 Three Mile 5 25,000 32,400 23,840,053 Island 1 Arkansas Nuclear 6 28,914 36,540 37,642,004 Unit 1 Rancho Seco 6 28,195 38,268 32,926,908 4

Crystal River 3 5 20,350 29,900 25,926,908 Davis-Besse 4 25,364 32,790 18,515,778 (a)As of January 15, 1984.

(b)As of November 30, 1983.

4-3 gatscock&WIIcom a McDermott company

, . . - , . . .._.. _ _ _ , . = . . _ _ _ _ . _ . . - . . _ _ . , . _ _ _ _ . . . . . . _ _ _ _ . . . . . _ . , _ _ . _ _ . _ . . . - _ . . . . _ . . . .

t Table 4-1. Fuel Design Parameters Batch 1E 48 5A 5B 6 7 FA type Mk-B4 Mk-B4 Mk-B4 Mk-B4 Mk-B4 Mk-85 Number of assemblies 1 16 8 40 48 64 Fuel rod OD, in. 0.430 0.430 0.430 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 0.377 0.377 0.377 Flexible spacer type Spring Spring Spri ng Spring Spri ng Spring Rigid spacer type Zirc-4 Zirc-4 Zirc-4 Zirc-4 Zirc-4 Zirc-4 Undensified active fuel length, in. 143.5 143.44 143.44 143.20 143.20 143.20 Fuel pellet (mean) dia., in. 0.3675 0.3697 0.3697 0.3686 0.3686 0.3686 Fuel pellet initial density, % TD mean 96 94 94 95 95 95 Initial fuel enrich-ment, wt % 235U 1.98 3.04 3.04 2.99 2.99 3.19 Estimated residence time, EFPH 18,336 29,249 22,145 22,145 15,600 9,360 Cladding collapse time, EFPH >35,000 >35,000 >35,000 >35,000 >35,000 >35,000 4-4 Babcock & Wilcox a McDermott company

Table 4-2. Fuel Thermal Analysis Parameters Batch 1E 4B/5A 5B/6 7 Number of assemblies 1 16/8 40/48 64 Initial density, % TD 96 94 95 95 Pellet diameter, in. 0.3675 0.3697 0.3686 0.3686 Nominal stack height, in. 143.5 143.44 143.2 143.2 Enrichment, wt % 235U 1.98 3.04 2.99 3.19 LHR capability, kW/ft to CFM 20.4 20.4 20.5 20.5 Densified fuel parameters (a)

TAFY3 Code Analysis Only Pellet diameter, in. 0.3651 0.3648 0.3649(b) 0.3649(b)

Fuel stack height, in. 143.14 141.65 142.13 142.13 Average fuel temperature, 'F 1340 1355 1464(c) 1464(c)

Nominal LHR, kW/ft at 2772 MWt 6.14 6.21 6.19 6.19 (a)Densification to 96.5% T' assumed for TAFY3 analysis.

(b)This data is provided for comparative purposes only and does not repre-sent parameter values used in TACO 2 analyses.

(c)BOL, TAC 02 code.

4-5 Babcock &WHcom a McDermott company

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Figure 4-1. Mark B5 Upper End Fitting (Side View)

SLOT STOP

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• There are two stop pin holes on each side of the upper end fitting. One contains a stop pin and the other is a spare.

4-6 Babcock & Wilcox

Figure 4-2. Holddown Spring Retainer

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Figure 4-3. Mark B5 BPRA Spider FOOT

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Figure 4-4. Mark BS BPRA Spider / Upper End Fitting /  :;X Reactor Internals Interaction W

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REACTOR INTERNALS h. .i1l.:

UPPER GRID PAD 2 T.i'

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5. NUCLEAR DESIGN N

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5.1. Physics Characteristics Table 5-1 compares the core physics parameters of cycle 4 with those of _g 8-10 -

cycle 5. These val ues were generated using PDQ07 for both cycles.

Di fferences in core physics parameters are to be expected between the 5 cycles due to the initial BPRA loading, the longer cycle 5 length, and the -d_

different shuffle pattern for cycle 5. Figure 5-1 illustrates a rep- ~@

resentative relative power distribution for BOC-5 at full power (FP) with equilibrium xenon and group 8 inserted. $%

Because of different isotopic distributions, cycle 5 control rod worths, ,-

ejected rod worths, and stuck rod worths differ from those of cycle 4. The 4 ejected rod worths in Table 5-1 are the maximum calculated values. D Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development 2 of the rod position limi ts presented in section 8. The adequacy of the $

e shutdown margin with cycle 5 rod worths is shown in Table' 5-2. The -Q following conservatisms were applied for the shutdown calculations: N

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth, p

3. Flux redistribution penalty. i e

Flux redistribution was taken into account since the shutdown analysis was

^

calculated using a two-dimensional model . The cycle 5 moderator coefficients and the power deficits from hot zero power (HZP) to hot full (RE power (HFP) are similar to those for cycle 4. The differential boron and @ =

xenon worths are al so similar in both cycles. The effective delayed W neutron fraction for cycle 5 shows a decrease with burnup (also shown in [

reference cycle 4). E

% _s

[

W3 5-1 Babcock & Wilcox S -

a McDermott company g -

5.2. Changes in Nuclear Design There is only one significant core design change between the reference cycle and the cycle 5 designs. The change is the increase in cycle life-time to 390 EFPD and the accompanying use of BPRAs to aid in reactivity control. The same calculational methods and design information were used to obtain the important nuclear design pa rameters. No significant oper-ational or procedural changes exist with regard to axial or radial power shape, xenon, or tilt control. The stability and control of the core with APSRs withdrawn have been analyzed. The calculated stability index without APSRs is -0.044h-1, which demonstrates the axial stability of the core.

5-2 Babcock & Wilcox J McDermott company

1 I

l Table 5-1. Davis-Besse Unit 1, Cycle 5 Physics Parameters Cycle 4 Cycle 5 Licensed cycle length, EFPD 280 390 Cycle burnup, mwd /mtU 9,350 13,043 Average core burnup - E0C, mwd /mtU 20,259 22,797 Initial core loading, mtU 83.0 82.9 Critical boren - B0C, No Xe, ppm HZP 1,250 1,485 HFP Group 8 inserted 1,042 1,255 Cri tical boron - E0C, Eq. Xe, ppm 2

Group 8 withdrawn (a) )(a)

Control rod worths - HFP, 80C, % ak/k Group 6 1.02 1.13 Group 7 1.73 1.42 Group 8 0.27 0.38 Control rod worths - HFP, E0C, % ak/k Group 7 1.68 1.46 Group 8 NA NA Max ejected rod worth - HZP, % Ak/k (location)

B0C Groups 5-8 inserted 0.85(b) 0.60 (N-12)

E0C Groups 5-7 inserted, Group 8 (N-12)(b) 0.85 0.55 withdrawn (N-12) (N-12)

Max stuck rod worth - HZP, % Ak/k (location)

BOC 1.70 0.80 (L-14) (N-12)

E0C 1.46 0.76 (N-12) (M-11)

Power deficit-HZP to HFP, Eq. Xe, % Ak/k BOC (4 EFPD) -1.77 -1.76 E0C -2.33 -2.48 Doppler coeff - HFP,10-3 BOC, No Xe, 1255 ppm,tc$ i GroupAk/k/*F 8 inserted -1.47 -1.50 E0C, Eq. Xe,10 ppm, Group 8 withdrawn -1.63 -1.76 Moderator coeff - HFP,10-2 % Ak/k/'F B0C, No Xe, 1255 ppm,Lc) Group 8 inserted -1.00 -0.81 E00, Eq Xe,10 ppm, Group 8 withdrawn -2.76 -2.86 Boron worth - HFP(clopm/% Ak/k B0C (1255 ppm) 108 123 E0C (10 ppm) 96 106 5-3 Babcock & Wilcox a McDermott company

Table 5-1. (Cont'd) ,

Cycle 4 C,vele 5 Xenon worth - HFP, % Ak/k B0C (4 EFPD) 2.67 2.63 E0C (equilibrium) 2.74 2.73 ,

Effective delayed neutron fraction - HFP B0C 0.00598 0.00631 E0C 0.00530 0.00524 (a) Power coastdown to E0C at 10 ppmb.

(b) Ejected rod worth at the rod insertion limit.

(c) Cycle 5 values were calculated at 1255 ppm. Cycle 4 values were calcu-lated at 1042 ppm.

5-4 NAMhz a MCOffmott Compar1y

Table 5-2. Shutdown Margin Calculation for Davis-Besse, Cycle 5 E0C, % Ak/k B00, 330 EFPD 390 EFPD

% Ak/k Bank 8 in Bank 8 out Available Rod Worth Total rod worth, HZP 7.31 7.57 7.44 Worth reduction due to burnup -0.17 -0.19 -0.18 of poison material Maximum stuck rod, HZP -0.80 -0.74 -0.76 Net worth 6.34 6.64 6.50 Less 10% uncertainty -0.63 -0.66 -0.65 Total available worth 5.71 5.98 5.85 Required Rod Worth ,9 Power deficit, HFP to HZP 1.76 2.52 2.48 i Max allowable inserted rod worth 0.42 0.49 0.52 Flux redistribution 0.65 1.18 1.20 Total required worth 2.83 4.19 4.20 -

t Shutdown Margin #y

+

Total available minus total 2.88 1.79 1.65 i required j Note: Required shutdown margin is 1.00% ak/k.

., n '~

5-5 Babcock & Wili:cm a McDermott ccmpan.

i Figure 5-1. B0C (4 EFPD), Cycle 5 Two-Dimensional Relative Power Distribution - Full Pgwgr, Equilibrium Xenon, APSRs Inserted (a) 8 9 10 11 12 13 14 15 H .794 .935 1.009 1.266 1.122 1.244 .940 .639 K .938 1.006 1.249 1.004 1.212 1.111 1.155 .643 8

i L 1.011 1.249 1.107 1.250 .876 1.263 1.061 .502 M 1.265 1.002 1.245 1.127 1.253 1.029 .791 8

N 1.120 1.210 .874 1.250 1.097 1.063 .505 1

0 1.243 1.109 1.261 1.029 1.064 .648 P .939 1.154 1.060 .790 .505

R .638 .642 .501 i

X Inserted Rod Group Number X.XX Relative Power Density (a) Calculated results from two-dimensional pin-by-pin PDQ07.

5-6 i

i

6. THERMAL-HYDRAULIC DESIGN Tha fresh batch 7 fuel is hydraulically and geometrically similar to the other fuel loaded into the cycle 5 core. The introduction of the it-B5 upper end fitting does not affect either the core flow rate or the thermal-hydraulic performance. The introduction of BPRAs increases the core flow available for heat transfer by reducing the core bypass flow rate from 10.7 to 8.1%. This reducea bypass flow rate has been conservatively neglected for cycle S. Therefore, the cycle 5 thermal-hydraulic design is identical to that of cycle 4. The thermal-hydraulic evaluation supporting cycle 5 operation is based on the methods and models described in references 11 and

, 12. The thermal-hydraulic design conditions for cycles 4 and 5 are sum-marized in Table 6-1.

Previous fuel cycle evaluations included the calculation of a rod bow penal-ty for each fuel batch based on the highest fuel rod burnup in that batch.

A rod bow topical report 13, which addresses the mechanisms and resulting local conditions of rod bow, has been submitted to and approved by the NRC.

The topical report concludes that rod bow penalty is insignificant and is offset by the reduction in power production capability of the FAs with irradiation. Therefore, no departure from nucleate boiling ratio (DNBR) reduction due to fuel red bow need be cont.idered for cycle 5.

6-1 BakM Nicou a McDermott company

Table 6-1. Davis-Besse Cycles 4 and 5 Thermal-Hydraulic Design Conditions Design power level, MWt 2772 Nominal system pressure, psia 2200

. Reactor coolant flow, % design 110 Vessel inlet / outlet coolant temp.,100% power, F 557.7/606.3 Ref design radial-local power peaking factor 1.71 Ref design axial flux shape 1.5 cosine with tails Hot channel factors Enthalpy rise (Fq) 1.011 1.014 Heat flux (F5)

Flow area 0.98 Active fuel length See Table 4-2 Avg heat flux,100% power, Btu /h-ft 2 1.89 x 105 (a)

Max heat flux,100% power, Btu /h-ft2 4.85 x 105 (a)

Critical heat flux (CHF) correlation BAW-2 Minimum DNBR, (% power) 1.79 (112%)

(a)With thermally expanded fuel rod OD of 0.43075 inch.

6-2 md &Mhz a McDermott company

l

)

I

7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in the cycle 5 parameters to determine the effects of the cycle 5 reload and to ensure that thermal performance during hypothetical transients is not degraded. The effects of fuel densification on the FSAR accident results have been evaluated and are reported in reference 11.

The radiological dose consequences of the FSAR chapter 15 accidents based on cycle 5 iodine and noble gas inventories have been evaluated. These doses are either bounded by the FSAR values or are a small fraction of the 10 CFR 100 limits.

7 . Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: (1) core thermal , (2) thermal-hydraulic, and (3) kinetics parameters including the reactivity feedback coefficients and control rod worths.

Fuel thermal analysis parameters from each batch in cycle 5 are given in Table 4-2. The cycle 4 and cycle 5 thermal-hydraulic maximum design con-ditions are presented in Table 6 -1. A comparison of the key kinetics parameters from the FSAR and cycle 5 is provided in Table 7-1.

A generic loss-of-coolant accident (LOCA) analysis for B&W 177-FA raised-loop nuclear steam systems (NSSs) has been performed using the Final Ac-ceptance Criteria ECCS Evaluation Model.14 The combination of average fuel temperature as a function of linear heat rate (LHR) and the lifetime pin pressure data used in the LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the analysis and the LOCA limits Babcock &Wilcou 7-1 a McDermott company

i reported in reference 14 provide conservative results for the operation of Davis-Besse Unit 1, cycle 5 fuel. A tabulation showing the bounding values for allowable LOCA peak LHRs for Davis-Besse Unit 1, cycle 5 fuel are pro-vided in Table 7-2.

It is concluded by the examination of cycle 5 core thermal, thermal-hydrau-lic, and kinetics properties, with respect to acceptable previous cycle val-ues, that this core reload will not adversely affect the ability to safely operate the Davis-Besse Unit 1 plant during cycle 5. Considering the previ-ously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 5 is considered to be boundad by previous 1y accepted analyses. The initial conditions of the transients in cycle 5 are bounded by the FSAR and/or the fuel densification report.

I s

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(

l l

l I

7-2 Babcock &Wiscos

! a McDermott company f

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

Table 7-1. Comparison of Key Parameters for Accident Analysis

• FSAR and densif'n report Cycle 5 Parameter value value BOL(a) Doppler coeff,10-3, % Ak/k/*F -1.28 -1.50 E0L(b) Doppler coeff,10-3, % ak/k/*F -1.45(c) -1.76 BOL moderator coeff,10-2, % ak/k/*F +0.13 -0.81 E0L moderator coeff,10-2, % tk/k/*F -3.0 -2.86 All rod bank worth (HZP), % Lk/k 10.0 7.31 Boron reactivity worth (HFP), ppm /1% Ak/k 100 123 Max ejected rod worth (HFP), % tk/k 0.65 0.32 Max dropped rod worth (HFP), % ak/k 0.65 0.20 Initial boron conc (HFP), ppm 1407 1255 (a)BOL denotes beginning of 11fe.

(b)EOL denotes end of life.

(c)-1.77 x 10-3 % ik/k/'F was used for steam line failure analysis.

Table 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates Allowable Allowable Core peak LHR, peak LHR, elevation, first 25 EFPD, balance of cycle, ft kW/ft kW/ft 2 15.5 16.5 4 16.8 17.2 6 18.0 18.4 8 17.5 17.5 10 17.0 17.0 Babcock &WHcom 7-3 , ucoermoit comp.ny

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 5 operation to ac-count for changes in power peaking and control rod worths. The effects of NUREG-0630 have been incorporated into the operating limits. Figures 8-1 through 8-26 are revisions to the previous cycle Technical Specifications.

Based on these Technical Specifications the final acceptance criteria ECCS limits will not be exceeded and the thermal design criteria will not be violated.

8-1 NdE Eb8 a McDermott company

Table 8-1. Reactor Protection System Instrumentation Trip Setpoints Table 2.2-1 Functional unit Trip setpoint Allowable values

1. Manual reactor trip Not applicable. Not applicable.

2. High flux <104.94% of RATED THERMAL POWER with <104.94% of RATED THERMAL POWER with Tour pumps operating Tour pumps operatingJ

<79.7% of RATED THERMAL POWER with <79.7% of RATED THERM.U. POWER with l Three pumps operating Three pumps operating #

3. RC high temperature <618'F <618*F#

4. Fl ux -- aflux/flowIII Trip setpoint not to exceed the lim- Allowable values not to exceed the it line of Figure 2.2-1 limit line of Figure 2.2-1#

>1983.4 psig**

{"4 5. RC low pressureIII 11983.4 psig 11983.4 psig*

6. RC high pressure <2300 psig <2300.0 psig* <2300.0 psig**

7. RC pressure-temperature (l) >(12.60 To ut *F - 5662.2) psig 1(12.60 To ut F - 5662.2) psig#

8. High flux <55.1% of RATED THERMAL POWER with <55.1% of RATED THERMAL POWER with pumps on({) number of RC one pump operating in each loop one pump operating in each loop #

<0.0% of RATED THERMAL POWER with <0.0% of RATED THERMAL POWER with Two pumps operating in one loop and Two pumps operating in one loop and no pumps operating in the other loop no pumps operating in the other loop #

E <0.0 of RATED THERMAL POWER with no <0.0% of RATED THERMAL POWER with no pumps operating or only one pump op- liumps operating or only one pump op-2a eratingf gg 2

erating

9. Containment pressure high 3 psig <4 psigi EIN

-4

i SAFETY LIMITS BASES ,

r t The reactor trip envelope appears to approach the safety limits more close-ly than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core out-let pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal f limits and account for the effects of potential fuel densification and po-  ;

tential fuel rod bow. -

1. The 1.30 DNBR limit produced by a nuclear power peaking factor of F0" i 2.56 or the combination of the radial peak, axial peak, and position of l the axial peak that yields no less than a 1.30 DNBR. j r

2. The combination of radial and axial peak that causes central fuel melt- +

ing at the hot spot. The limits are 20.4 kW/f t for batches 1E, 48, and ,

5A and 20.5 kW/ft for batches 58, 6, and 7. j Power peaking is not a directly observable quantity and therefore limits I have been established on the basis of the reactor power imbalance produced j t by the power peaking. -

The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond to 1 the expected minimum flow rates with four pumps and three pumps, respective-ly.  ;

The curve of Figure 2.1-1 is the most restrictive of all possible reactor  ;

coolant pump-maximum thermal power combinations show1 in BASES Figure 2.1.  !

The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the num- [

, ber of reactor coolant pumps in operation or the local quality at the point  ;

of minimum DNBR is equal to +22%,'whichever conot t.lon is more restrictive. t These curves include the potential effects of fuel red bow and fuel densiff- '

cation. l The DNBR as calculated by the 84W-2 DNB correlation continually increases f from point of minimum DNBR, so that the exit DNBR is always higher. Extrap- l olation of the correlation beyond its published quality range of +22% is j justified on the basis of experimental data. i I

i  !

i B 2-2 l 8-3 Wha a McDermot conyany

LIMITING SAFETY SYSTEM SETTINGS BASES RC High Temperature The RC high temperature trip <618'F prevents the reactor outlet temperature from exceeding the design limTts and acts as a backup trip for all power ex-

, cursion transients.

Flux -- AFlux/ Flow The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accommodate flow decreasing transients from high power where protection is not provided by the high flux / number of reactor coolant pumps on trips.

The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, ana for every power level there is a minimum permissible low flow rate. Examples of typical power level and low flow rate combinations for the pump situations of Table 2.2-1 that would result in a trip are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 106.8% and reactor coolant flow rate is 100% of full flow rate, or flow rate is 93.63% of full flow rate and power level is 100%.

2. Trip would occur when three reactor coolant pumps are operating if power is 79.7% and reactor coolant flow rate is 74.7% of full flow rate, or flow rate is 70.22% of full flow race and power is 75%.

For safety calculations the instrumentation errors for the power level were used. Full flow rate in the above two examples is defined as the flow cal-culated by the heat balance at 100% power.

B 2-5 84 Babcock & WHeen A McDermott company

REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups shall b( limited in physical insertion as shown on Figures 3.1-2a -2b -2c, and -21 and 3.1-3a, -3b, -3c and -3d. l A rod group overlap of 2515% shall be maintained between sequential with-drawn groups 5, 6 and 7.

APPLICABILITY: MODES 1* and 2*#.

ACTION ,

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

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

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

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

NOTE: If in unacceptable region, also see Section 3/4.1.1.1.

• See Special Test Exception 3.10.1 and 3.10.2.

1. Withkefff,1.0.

DAVIS-BESSE, UNIT 1 3/4 1-26 8-5 Md aha a McDermott 4ompany

REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING R00 INSERTION LIMITS LIMITING CONDITION,FOR OP,ERATION , ,,

3.1.3.9 The axial power shaping rod group shall be limited in physical in-sertion as shown on Figures 3.1-Sa, -5b, -5c, -5d, -Se, -Sf, and -59 . l APPLICABILITY: MODES 1 and 2*.

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

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

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

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

SURVEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be deter-mined to be within the insertion limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• With Ke rf 3,1.0.

DA5/IS-BESSE, UNIT 1 3/4 1-34 8-6 M M M ***

A M(Detmoll(Ompdriy

i 3/4.2. POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-la, -lb, -1c, and -Id and 3.2-2a, -2b, -2c and -2d. l APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.*

ACTION With AXIAL POWER IMBALANCE exceeding the ifmits-specified above, either:

a. Restore the AXIAL POWER IMBALANCE to within i ts limits wi thir. 15 minutes, or

b. Within one hour reduce power until imbalance limits are met or to 40%

of RATED THERMAL POWER or less.

SURVEILLANCE REQUIREMENTS 4.2.1. The AXIAL POWER IMBALANCE shall be determined to be within ifmits 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 THERMAL POWER except when the AXIAL POWER IMSALANCE alann is inoperable then calculate the ,

AXIAL POWER IMBALANCE at least once per hour.

s a

n f, 1

I L

i

• See Special Test exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 8-7 Babcock &WIIcom a McDermott company

Table 8-2. Quadrant Power Tilt Limits (Tech. Spec. Table 3.2.2)

Steady state Transient Maximum limit limit limit Measurement independent 4.92 11.07 20.0 QUADRANT POWER TILT QUADRANT POWER TILT as measured by:

Symmetrical incore detector 3.37 8.52 20.0 system, 0-50 110 EFPD Symmetrical incore detector 3.02 8.52 20.0 system, after 50 t10 EFPD _

Power range channels 1.96 6.96 20.0 Minimum incore detector system 1.90 4.40 20.0 3/4 2-12 8-8 Babcock &Wilcox a McDermott company

c l' 2

-:p

+  ?.

3/4.4. Rt: ACTOR COOLANT SYSTEM ., f 3/4.4.1. COOLANT LOOPS AND COOLANT CIRCULATION Y STARTUP AND POWER OPERATION .- f-

-t LIMITING CONDITION FOR OPERATION _

[

3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each 1-loop shall be in operation. e_p i

APPLICABILITY: MODES 1 and 2*. y l{

ACTION: - .!

=*

a. With one reactor coolant pump not in operation, STARTUP and POWER OPERA- ' '

TICN may be initiated and may proceed provided THERMAL POWER is re- .

stricted to less than 79.7% of RATED THERMAL POWER and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l .-

1(

the setpoints for the following trips have been reduced to the values -i specified in Specification 2.2.1 for operation with three reactor cool- =y -

ant pumps operating:

1. High Flux ,

2. Flux-AFlux-Fl ow ..-- ,-  ;.

T SURVEILLAhCE REQUIREMENTS .

4.4.1.1 The above required reactor coolant loops shall be verified to be J2p in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ~~$

-y 4.4.1.2 The reactor protective instrumentation channels specified in the ,

applicable ACTION statement above shall be verified to have had their trip ,

setpoints changed to the values specified in Specificstion 2.2.1 for the ap-plicable number of reactor coolant pumps operating either: ..

_"{

a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after switching to a different pump combination if the $

switch is made while operating, or __

1

b. Prior to reactor criticali ty if the switch is made whi e sh stdown. z

_{o y -.

?I ;fI

' (

1

• See Special Test Exception 3.10.3. [ _.

j DAVIS-BESSE, UNIT 1 3/4 4-1 ,[

sF 8-g Babcock &Wilcom a a McDermo:t company J -' .

-w-w ht. 'l . -

'h h- i.h

Figure 8-1. Reactor Core Safety Limit (Tech. Spec. Figure 2.1-2)

% RATED THERMAL POWER i

1

- 120 (44,112)

(-48,112.0)Q

(-49,100.0)b -

- 100 (49,100)

(-48,89,1) P (44,89.1)

F 3 PUMP LIMIT

-80 )[4g,77,7)

(-49,77.1) ()

ACCEPTABLE

- 60 OPERATION UNACCEPTABLE FOR SPECIFIED UNACCEPTABLE OPERATION RC PUMP OPERATION COMBINATION - - 40

- 20 I l l l 1 t i I

-60 -40 -20 0 20 40 60 Axial Power Imbalance, %

PUMPS OPERATING REACTOR COOLANT FLOW, GPM 4 387,200 3 290,100 8-10 Babcock &WHcom a McDermott company

Figure 8-2. Trip Setpoint for Flux -- AFlux/ Flow (Tech. Spec. Figure 2.2-1)

Curve shows trip setpoint for a 25% flow reduction for three pump operation (290,100 gpm). The actual setpoint will be directly proportional to the actual flow with three pumps.

% RATED THERMAL POWER UNACCEPTABLE

- 120 OPERATION UNACCEPTABLE OPEPATION

(-18.2,106.8) (18.2,106.8)

M =1.000 1 ._.P;; -1.000 I

LIMIT 3 I

(-34.0,91.0) 34.0,91.0) l I i

I 80 (18.2,179.7)

(-18. 2,l79.7 ) -

l 3l PUMP  !

! l

(-34.0,63.9) l LIflT 34.0,63.9) l

! 60 l I

l l l ACCEP ABLE OPE RATION F@

l SPECI IED RC F I

JMP COMBIfATION I l -

- 40  !

! l I l I

I l l 1 I 20 l I I l l l

1 l l l t I i i I l l1 l 1 1 I

-60 -40 -20 0 20 40 60 Axial Power Imbalance, %

8-11 Babcock & WHcom a McDermott company

i

- \

Figure 8-3. Regulating Group Position Limits, 0 to 25+10/-0 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-2a)

(275,102)

(229,102) -

100 -

Power Level Cutoff = 100% (300,102)

(270,92) 2 W 80 - SHUTDOWN 2 MARGIN (250,80) g LIMIT E

W UNACCEPTABLE RESTRICTED 60 OPERCION OPERATION S

g (159,50)

(225,50) o

, 40 -

5 E

20 -

b ACCEPTABLE (86,15) OPERATION

[

(0,7.6) 0 I I I 0 100 200 300 Rod Index (% Withdrawn)

GR S i i i l 0 75 100 GR 61 1 I I 0 25 75 100 l

I I GR 7 '

0 25 100 l

1 8-12

Figure 8-4. Regulating Group Position Limits, 25+10/-0 to 200t10 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5

' Tech. Spec. Figure 3.1-2b)

(229,102) (27),102[

100 - Power Level (300,102)

Cutoff = 100% (270,92) n 5

{ 80 -

SHUTDOWN MARGIN (225,80)

$ LIMIT

$ UNACCEPTABLE N 60 - OPERATION 8

(159,50) (200,50) 40 -

p 8

u

$, ACCEPTABLE OPERATION b 20

[ (86,15)

(0,7.6) '

0 O 100 200 300 Rod Index (% Withdrawn)

GP, 5 0 75 100 GR 6 0 25 75 100 GR 7 'O 25 100 8-13 l'

Figure 8-5. Regulating Group Position Limits, 200 10 to 330 110 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-2c)

(275,102)

. (267,102) 100 -

Power Level (300,102)

Cutoff = 100%

OPERATION RESTRICTED (270,92) 5 g 80 -

( 9,0)

SHUTDOWN a MARGIN f LIMIT 5

? 60 - I o UNACCEPTABLE

{

a OPERATION (200,50) o -

, 40 5

E ACCEPTABLE 2 OPERATION g 20 j (126,15)

(0,5.7) , , i 0 300 0 100 200 Rod Index (% Withdtawn)

GR 5 75 100 0 25 75 100 GR 7 0 25 100 8-14

Figure 8-6. Regulating Group Position Limits, 330 i10 to 390 i10 EFPD, Four RC Pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-2d)

( 1,102)=

100 -

Power Level (30b,102)

Cutoff = 100%

S SHUTDOWN 80 - MARGIN LIMIT a

I 60 -

e Q

UNACCEPTABLE (206'50)

OPERATION o

, 40-5 E.

d'

~

ACCEPTABLE OPERATION b 20-x e (134,15)

(0,4.6) , , ,

0 0 100 200 300 Rod Index (% Withdrawn)

GR 5 '

0 75 100 GR 6 0 25 75 100 GR 7 0 25 100 l

8-15

Figure 8-7. Regulating Group Position Limits, O to 25+10/-0 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-3a) 100 -

$g 80 -

(229,77)

(275,77) 0

' (300,77)

(270,69.5) f ce SHUTDOWN MARGIN (250,60.5) h 60 -

LIMIT S PERATION UNACCEPTABLE y

• OPERATION 40 -

(1G9,38) (225,38) 5 E

E T 20 -

ACCEPTABLE y OPERATION o

" (86,11.75)

(0,6.2) , i 1 0 200 300 9 100 Rod Index (% Withdrawn)

GR 5 75 100 0

t t I I GR 6 25 75 100 0

t i I GR 7 25 100 0

8-16

l I

Figure 8-8. Regulating Group Position Limits, 25+10/-0 to 200 10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-3b) 100 -

E W 80 ~ (275,77) g (229,77) a (300,77)

E IEl0sC9.5) h

~ - SHUTDOWN 60 MARGIN (225'50*S) 8 LIMIT Q

" yyACCEPTABLE OPERATION d

u 40 -

(159,38) $ (200,38)

E f

20 ~ ACCEPTABLE b TION

~ --

'(86,11,73)

I ' _I O 200 E00 300 Rod Index (g y; drawn) n I 1 GR S O 75 100 GR 6 *' 3b 7; 100

, 1 GR 7 ' 25 100 8-17

Figure 8-9. Regulating Group Position Limits, 200 t10 to 330 10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-3c) l 100 -

S 3 80 - (275,77)

' (267,77)

(300,77)

N OPERATION

$ RESTRICTED (270,69.5)

E -

0 l

8 (239,60.5) i e 5

l SHUTDOWN i

% MARGIN a 40 - LIMIT b UNACCEPTABLE

, (200'38) l b OPERATION i b 20 -

ACCEPTABLE

j. OPERATION (126,11.75)

(0,4.7) , , ,

l 0

0 100 200 300

' ' Rod Index (% Withdrawn) l GR 5 I

0 75 100

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

l i

r i

l i

l 8-18 i

L

Figure 8-10. Regulating Group Position Limits, 330 110 to 390 10 EFPD, Three RC Pumps, APSRs Withdrawn -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-3d) 100 -

2 s

g 80 -

(271,77) o a (300,77) 5 5 60 - SHUTDOWN o MARGIN N LIMIT N UNACCEPTABLE g OPERATION u 40 g (206,38)

E b 20 ACCEPTABLE OPERATION

[o

• (134,11.75)

(0,4.0 t i I 0 200 300 0 100 Rod Index (% Withdrawn)

GR 5 ' ' '

0 75 100 0 25 75 100 GR 7 100 0 25 8-19

Figure 8-11. APSR Position Limits, O to 25+10/-0 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-Sa)

(9,102) (38,102) 100 -

'(9,92)

- (38,92)' RESTRICTED p REGION u.

g 80<

(0,80) (50,80) 5 5 60 -

8 s -

N (100,50) o -

PERMISSIBLE 40 y OPERATING REGION 8 -

b S.

t 20 E

a- _

0 ' I I I I I I I I I 0 10 20 30 40 50 60 70 80 90 100 APSR Position (% Withdrawn) 8-20

Figure 8-12. APSR Position Limits, 25+10/-0 to 200 10 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-5b)

(9,102) (42,102)

C ~'

100 -

'(9,92) RESTRICTED (42'92) p REGION w

@ 8 04 0,80)

I (50,80)

• 60 -

8 4

(100,50)

% PERMISSIBLE a 40 -

OPERATING REGION 5

E _

E b 20 -

E

' ' ' ' ' ' ' ' ' J 0 -

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

\ 8-21

\\

Figure 8-13. APSR Position Limits, 200 10 to 330 110 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-5c)

(9,102) (42,102) 100 -

~

p(9,92) RESTRICTED (42,92) REGION g

w

$ 80<

(0,80) (50,80)

[

I a:

(100,70)

M

" 60 -

O PERMISSIBLE h -

OPERATING REGION

$ 40 -

E E -

E g 20 -

5

' ' ' ' ' ' ' ' ' 8 0 90 0 10 20 30 40 50 60 70 80 100 APSR Position (% Withdrawn) 8-22

Figure 8-14. APSR Position Limits, 330 110 to 390 10 EFPD, Three or Four RC Pumps, APSRs Withdrawn --

Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-5d) 100 -

2 -

W 2

_j 80 -

C '

1. f 60 APSR INSERTION NOT ALLOWED EE IN THIS TIME INTERVAL o

a 5 40 E

ss -

u y 20 -

c2 0

' ' ' i i i -i i 0 10 20 30 40 50 60 70 80 90 1.00 APSR Position (% Withdrawn) l l

[

8-23

Figure 8-15. APSR Position Limits, 0 to 25+10/-0 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-Se)

_ 100 -

a:

Y 2

l 80 - (9,77) (38_,77 )

g a RESTRICTED E ,(9 *69.5 )

e (38,69.5) REGICN Y 60 E (0,60.5) (50,60.5) o a

$ 40 -

( 00,38)

PERMISSIBLE u OPERATING REGION

$ 20 -

o.

0 0 10 20 30 40 50 50 70 80 90 100 APSR Position (% Withdrawn) 8-24

l Figure 8-16. APSR Position Limits, 25+10/-0 to 200 t10 EFPD, Three RC Pumps -- Davis-Bess? 1, Cycle 5 (Tech. Spec. Figure 3.1-5f)

-, 100 85 3

m.

!! 80 - (9,77).

(42,77)

^

OC RESTRICTED EE

'(9,69.5) (42,69.5) gj 60 (0,60.5) (50,60.5) 15 Y

8 40 -

u (100,38)

JE PERMISSIBLE b

20 -

OPERATING REGION B

o.

0 0 10 20 30 40 50 60 70 80 90 100 APSR Position (% Withdrawn) i 8-25

Figure 8-17. APSR Position Limits, 200 10 to 330 t10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.1-59)

- 100 -

5 5

c.

a j 80 - (9,77) (42,77)

~

w RESTRICTED 5 '(9,69*5) (42,69. 5 )'

REGION g 60 (0,60.5) (50,60.5)

D (100,5Y)

E y 40 -

$ PERMISSIBLE OPERATING REGION u

as o 20 -

0 0 10 20 30 40 50 60 70 80 90 100 APSR Position (% Withdrawn) e 8-26

Figure 8-18. Axial Power Imbalance Limits, O to 25+10/-0 EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.2-la)

(-23,102) -

,02)

-100 l

(-25,92) (

m

-90 5

(-30,80)r

{--80 a

(30,80) h--70 E

r 00 g

e

= -

-50 RESTRICTED PERMISSIBLE REGION OPERATING o REGION g -

-40 8

5 -

-30 3

5 -

-20 5

a.

-10 I f f I f f

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance (%)

l l

l 8-27 Babcock &Wilcox a McDermott company

Figure 8-19. Axial Power Imbalance Limits, 25+10/-0 to 200 10 EFPD, Four RC Pump -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.2-lb)

(-23,102)~ ' (23,102) -

- 100

(-30,92)e ' (30,92) c2 -- 90 O

- 80 A

r 5-- 70 N

8-- 60 -

e

- 50 RESTRICTED PERMISSIBLE o REGION OPERATING 5-- 40 REGION

{

cb -- 30 5

g-- 20 c.

- 10 I f I f f I

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance (%)

8-28

1 Figure 8-20. Axial Power Imbalance Limits, 200 10 to 330 i10 l EFPD, Four RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.2-1c)

(-23,102) = ^(

--100 '

(-30,92)' E - -90 i(30,92)

-80 I

-70 E

S -

-60

!E e

g -

-50 RESTRICTED PERMISSIBLE E REGION OPERATING 8

-40 REGION 5 6 -

-30 b

g -

-20 c.

-10 I I f f I I

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance (%)

8-29

Figure 8-21. Axial Power Imbalance Limits, 330 i10 to 390 10 EFPD, Four RC Pumps, APSRs Withdrawn --

Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.2-1d)

(-23,102)- (23,102)

- 100

(-30,92)" g-- 90 '(30,92)

Y E--

a 80 f

$-- 70 S-- 60 ti

[-- 50 RESTRICTED PERMISSIBLE REGION OPERATING REGION t--

40 s.

y-- 30 t

j a.

- 20

- 10 1 I I f I t

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance (%)

8-30

Figure 8-22. Axial Power Imbalance Limits, 0 to 25+10/-0 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Fiqure 3.2-2a)

- 100 2

so

- 80

(-17.25,77)- (17.25,77)

(-18.75,69.5) @ (18.75,69.5)

(-22.5,60.5) E- - 60 (22.5,60.5)

S tice z -

- 40 S e RESTRICTED $$ $

REGION L 2

~

$5 -

- 20 53 b .

Em '

B 8

i I , ,1 I I l. , t I

-40 -30 -20 -10 0 10 20' 30 40 Axial Power Imbalance (%)

8-31

Figure 8-23. Axial Power Imbalance Limits, 25+10/-0 to 200 10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.2-2b) 100 2

Y o --

-- 80

(-17.25,77 - -

- 17.25,77) a

(-22.5,69.5)i $ 1(22.5,69.5)

F- --

- 60 8

ti a:

z o --

- 40 S a RESTRICTED bj$ 5 REGION $ a: E

$1E SE E p: 20

&& 5 a-r o

3 n-t t y

I . t I; e- 1 i

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance (%)

8-32

Figure 8-24. Axial Power Imbalance Limits, 200 10 to 330 t10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 5 (Tech. Spec. Figure 3.2-2c)

- 100 E

s

(-17.25,77 _

- 80 , 17.25,77)

(-22.5,69.5)i E g (22.5,69.5) 5 a

- 60 Y

E 5

+

wE -

-40 RESTRICTED SE U REGION @g $

EC 2 EE ~

-20 "Mo b L

5 a t t t t Isk i 1

-40 -30 -20 -10 0 10 20' 30 40 Axial Power Imbalance (%)

8-33

F .

i g

u r

(

e R 2 8 E 2 -

RS 2 ET 5 ( 5 GR , -

- II 6 1 3 A OC 9 7 0 NT DEA E 5 2 aFx D ) 5 vPi s ,

2

- u 7 s ,D a i

l I

7 -

A 0 E{ 0mgL a ) B T .'

x l J , eho i

a 85{E gb5 srw l -

see 1 t eer P 0 1RI o Cm w a85^

Q

. 8b ~ ,58C Q "eS HE e b 8 r CPa

- yul 3 I 0

- - . - - cma 4 m - - . - - lp n b

a 2 4 6 8 1 e ,s ec l

0 0 0 0 0 5 a 0 AL n 1 (Pi c 0 I

TSm e eRi cst

(

h. W ,s

) 2 a i 0 i 2

6

(

\(1 2

7 2

St3 p h3 ed0 c.

r at 3 .

2 5 w1 0 i Fn0 5 7 i

, 7 g - t 6 ) u- o 9 r 5

e 3 4 9 t ) 3 0 0

2 t

- 1 2 0 d

)

l l ll

Figure 8-26. Control Rod Core Locations and Group Assignments -- Davis-Besse 1, Cycle 5 E (Tech. Spec. Figure 3.1-4)

L m

X

=

A r

R B 3 7 3 F

C 2 6 6 2 1

D 7 8 5 8 7 m,

F E 2 5 5 2 F 3 8 1 7 1 8 3 E- G 6 4 4 6 w- -y 7

g 7 5 7 1 7 5 7 b

K 6 4 4 6 L 3 8 1 7 -1 8 3 M 2 5 5 2 E

I n 7 8 5 8 7 b_ 0 2 6 6 2 w

e P 3 7 3 R

(

7 ,

Z'

_ 1 2 3 4 5 6 7 8 9 3 11 12 13 14 15

=

w ,

I. .

Group

~

Functions L 1 ~6' Safety 2 8 Safety g 'X7 Group Number 3 8 Safety

i 4 4 Safety 5 8 Control 6 8 Control

=r 7 12 Control Elk <

8435 8 8 APSRs Total # 61 EM: M b

9. STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is out-lined bel ow. These tests verify that core performance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of the unit.

9.1. Precritical Tests 9.1.1. Control Rod Trip Test Precritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptance criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions stated above.

It should be noted that safety analysis calculations are based on a rod drop time of 1.40 seconds from fully withdrawn to two-thirds inserted.

Since the most accurate position indication is obtained from the zone reference switch at the 75% inserted position, this position is used for data gathering instead of the two-thirds inserted position. The acceptance criterion of 1.40 seconds corrected to a 75% inserted position (by rod insertion versus time correlation) is 1.66 seconds.

9.1.2. Reactor Coolant Flow Reactor coolant (RC) flow with four reactor coolant pumps (RCPs) running will be measured at HZP steady-state conditions. Acceptance criteria require that the measured flow be within allowable limits.

9-1 Babcock &Wilcom a McDermott company

9.2. Zero Power Physics Tests 9.2.1. Critical Boron Concentration Criticality is obtained by deboration at a constant dilution rate. Once criticality is achieved, equilibrium boron is obtained and the critical boron concentration determined. The critical boron concentration is cal-culated by correcting for any rod withdrawal required in achieving equilib-rium boron. The acceptance criterion placed on critical boron concentra-tion is that the actual boron concentration must be within 100 ppm boron of the predicted value.

9.2.2. Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the HZP rod insertion limit. The average coolant temperature is varied by 5'F. During the change in temperature, =

reactivity feedback is compensated by a discrete change in rod motion; the change is then calculated by the summation of reactivity (obtained from a reactivity calculation on a strip chart recorder) associated with the tem-perature change. Acceptance criteria state that the measured value shall not differ from the predicted value by more than 0.4 x 10-2 (% Ak/k)/ F (predicted value obtained from Physics Test Manual curves).

The moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of the fuel Doppler coefficient of reactivi ty is added to obtain the moderator coefficient. This value must not be in excess of the acceptance criteria limit of +0.9 r. 10-2

(% ak/k)/*F.

9.2.3. Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6, and 7) are measured at HZP conditions using the boron / rod swap method. The boron / rod swap method consists of establishing a deboration rate in the RC system and compensat-ing for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 in incremental steps. The reactivity changes that occur during these measurements are cal culated based on reactimeter data, and dif ferential rod w3rths are obtained from the measured reactivity worth 9-2 Babcock & Wilcox a McDermott company

versus the change in rod group position. The differential rod worths of each of the controlling groups are then summed to obtain integral rod group worths. The acceptance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

predicted value - measured value x 100 < 15.

measured value --

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

predicted value - measured value x 100 < 10.

measured value --

9.2.4. Ejected Control Rod Reactivity Worth After CRA groups 7, 6, and 5 have been positioned near the minimum rod insertion limit, the ejected rod is borated to 100% withdrawn and the worth obtained by adding the incremental changes in reactivity by boration.

After the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is then swapped with the controlling rod group and the worth determined by the change in the previously calibrated controlling rod group position. The boron and rod swap values are averaged and error-adjusted to determine ejected rod worth. Acceptance criteria for the ejected rod worth test are as follows:

1. predicted value - measured value x100 < 20.

measured value --

2. Measured value (error adjusted) j[ 1.0% ak/k.

The predicted ejected rod worth is given in the Physics Test Manual.

9.3. Power Escalation Tests 9.3.1. Core Power Distribution Verification

! at $40, s75, and s100% FP With Nominal Control Rod Position Core power distribution tests are performed at $40, s75, and s100 FP. The test at 40% FP is essentially a check on power distribution in the core to identify any abnormalities before escalating to the 75% FP plateau, Rod in-dex is established at a nominal FP rod configuration at which the core power distribution was calculated. APSR position is established to provide 9-3 Babcock & WHcom a McDermott company

a core power imbalance corresponding to the imbalance at which the core power distribution calculations were performed.

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

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

2. The minimum DNBR must be greater than 1.30.

3. The value obtained from the extrapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30 or the extrapolated value of imbalance must fall outside the reactor protector system (RPS) power / imbalance / flow trip envelope.

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

5. The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.

6. The highest measured and predicted radial peaks shall be within the fol-lowing limits:

predicted value - measured value x 100 < 8.

measured value

7. The highest measured and predicted total peaks shall be within the following limits:

predicted value - measured value x 100 < 12.

measured value Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and thermal calculational models, thereby verifying the acceptability of data from these models for input to safety evaluations.

Items 3 and 4 establish the criteria whereby escalation to the next power plateau may be accomplished without exceeding the safety limits specified by the' safety analysis with regard to DNBR and LHR.

The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test except that core equilibrium xenon is established before 9-4 NM EMN a McDermott company

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

1. The highest measured and predicted radial peaks shall be within the fol-lowing limits:

predicted value - measured value x 100 < 5.

measured value

2. The highest measured and predicted total peaks shall be within the fol-lowing limits:

predicted value - measured value x 100 < 7.5.

measured value -

9.3.2. Incore Versus Excore Detector Imbalance Correlation Verification ats40% FP Imbalances are set up in the core by control rod positioning. Various imbalances are read simultaneously on the incore detectors and excore power range detectors. The excore versus incore detector cffset slopes must be W at least 1.15. If the excore versus incore detector offset slope criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.

9.3.3. Temperature Reactivity Coefficient at s100% FP

~

The average RC temperature is decreased and then increased by about SF at constant reac' tor power. The 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.

The temperature reactivity coefficient is calculated from the measured changes in reactivity and temperature.

Acceptance criteria state that the moderator temperature coefficient shall i not be positive above 95% FP.

9.3.4. Power Doppler Reactivity Coefficient l

at s 100% FP Reactor power is decreased and then increased by about 5% FP. The reactivity change is obtained from the change in controlling rod group 9-5 Babcock &WHcom a McDermott company

position. Control rod group worth is measured using the fast insert /with-draw method. Reactivity corrections are made for changes in xenon and RC temperature that occur during the measurement. The power Doppler reactivity coefficient is calculated from the measured reactivi ty change, which is adjusted as stated above, and the measured power change.

The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual . Acceptance criteria state that the measured value shall be more negative than -0.55 x lv-2 (% Ak/k)/% FP.

9.4. Procedure for Use When Acceptance Criteria Are Not Met If acceptance criteria for any test are not met, an evaluation is performed with participation by B&W technical personnel as required. Further speci-fic actions depend on the evaluation results. These actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for anomalies, or design personnel performing detailed analyses of potential safety problems because of parameter deviation. Power is not escalated until the evaluation shows that plant safety will not be compromised by such escalation.

9-6 Babcock &WHcom a McDermott company

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

2 BPRA Retainer Design Report, BAW-1496, Babcock & Wile ox, Lyachburg, Virginia, May 1978.

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

4 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084PA, Rev 2, Babcock & Wilcox, Lynchburg, Virginia, December 1978.

5 TACO-2 Fuel Pin Performance Analysis, BAW-10141P, Babcock & Wilcox, Lynchburg, Virginia, January 1979.

6 J. H. Taylor (B&W) to J. S. Berggren (NRC), Letter, "B&W's Responses to TAC 02 Questions," April 8,1982.

7 TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1972.

8 B&W Version of PDQ07 Code, BAW-10117A, Babcock & Wilcox, Lynchburg, Virginia, January 1977.

9 Core Calculational Techniques and Procedures, BAW-10118A, Babcock &

Wilcox, Lynchburg, Virginia, December 1979.

10 Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Babcock &

pilcox, Lynchburg, Virginia, May 1977.

11 Davis-Besso Unit 1 Fuel Densification Report, BAW-1401, Babcock &

Wilcox, Lynchburg, Virginia, April 1975.

12 Attachment 1 to Application to Amend Operating License for Removal of Burnable Poison Rod and Orifice Rod Assemblies, BAW-1489, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, May 1378.

A-1 Babcock & Wilcox a McDermott company

13 Fuel . Rod Bowing in Babcock & Wilcox Fuel Designs, BAW-10147P, Babcock &

Wilcox, Lynchburg, Virginia, April 1981.

14 ECCS Evaluation of B&W's 177-FA Raised-Loop NSS, BAW-10105, Rev. 1, Babcock 8 Wilcox, Lynchburg, Virginia, July 1975.

A-2 Babcoc8:4WIfcom a McDermott comparty

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