ML19323F282

Revision to Davis-Besse Nuclear Power Station Unit 1, Cycle 2 - Reload Rept.

ML19323F282
Person / Time
Site:	Davis Besse
Issue date:	05/19/1980
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19323F273	List: ML19323F272 ML19323F278 ML19323F282
References
BAW-1598, NUDOCS 8005280725
Download: ML19323F282 (20)

Text

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. Toledo Edison Company

) g Revision dated May 19, 1980 i W to BN4-1598

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, DAVIS-BESSE NUCLEAR POWER STATION i

i UNIT 1, CTCLE 2 - RELOAD REPORT  ;

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B&W Revisica 1 (5/13/80)

3. GENERAL DESCRIPTION The Davis Besse Unit I reactor core is described in detail in chapter 4 of the Final Safety Analysis Report 2 for the unit. The cycle 2 core consists of 177 FAs, each of which is a 15 by 15 array containing 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. All FAs in batch 4

.' have a constant nominal fuel loading of 468.25 kg of uranium. Batches IB, 2, and 3 have a fuel load'.ng of 472.24 kg of uranium. The fuel consists of dished-end cylindrical pellr.cs of uranium dioxide clad in cold-worked Zircaloy-4. The i undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuci 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 1 cycle 2. The initial enrichment of batch 1B is 1.98 wt % uranium-235. Batches 2, 3, and 4 have a 2.63, 2.96, and 3.04 uranium-235 enrichment, respectively. There will be 44 1

batch 1 assemblies discharged at the end of cycle 1; the remaining 12 batch 1 and all batch 2 and 3 assemblies will be shuffled to new locations. The batch 4 as'semblies will occupy the periphery of the core. Note that the designation -

1B is used to identify the batch 1 assemblies being rcused for cycle 2. Batch 1A is now the remainder of batch 1 assemblies which have not been scheduled

for reinsertion. Figure 3-2 is an eight.h-cpre map showing each assembly's burnup at the beginning-of-cycle (BOC) 2 and its initial enrichment.

Cycle 2 will be operated in a feed-and-bleed mode. The core reactivity con-

trol will be supplied mainly by soluble boron and supplemented by 53 full length Ag-In-Cd control rod assemblies (CRAs). In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for addi-tional control of the axial power distribution. The cycle 2 locations of the 61 control rods and the group designations are indicated in Figure 3-3. Al-though the rod group designations differ, the core locations of the 61 control l -

rods for cycle 2 are identical to those of the reference cycle 1.

i j 3-1 Babcock & Wilcox 4

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B&W Rsvision 1 (5/13/80)

Figure 3-1. Davis Besse 1, Cycle 2 Shuffle rust Tannstta cAnn. ,

x 1

A F F F F F

, F F 3 2 3 2 3 F F M2 F7 N2 F9 M14

• 3 F 2 2 3 2 3 2 2 F 3 N14 E6 C6 L1 C8 L15 C10 E10 P4 o F F IB 3 2 2 3 2 2 3 1B F F

, C5 L2 N11 D7 PS~ D9 N5 Ll4 E13 t F 2 3 2 3 3 2 3 3 2 3 2 F F5 B10 D4 N3 Kl' ES K15 N13 D12 B6 Fil F 3 2 2 3 IB 2 3 2 1B

' 3 2 2 3 F

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Bil F3 M12 C12 ' E3 B7 R8 B9 Cll C4 M4 F13 B5 F 2 3 2 3 2 3 IB 3 3

' ~2 2 3 2 F C6 A10 G4 A9 G2 013 D10 03 G14 A7 G12 F1 G10

., F 3 2 3 2 3 1B 2 1B 3 2 3 2 3 F _,

P12 H3 H14 HS H15 F4 H8 L12 H1 Hll H2 H13 B4 g F 2 3 2 3 2 3 IB 3 1 2 3 2 3 2 F K6 R10 K4 R9 K2 Cl3 N6 C3 K14 R7 K12 R6 K10 F 3 2 2 3 IB 2 3 2 1B t 3 2 2 3 F Pll L3 E12 012 05 P7 A8 P9 M13 04 E4 L13 PS

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F 2 3 2 3 3 2

, 3 3 2 3 2' F L5 P10 N4 D3 G1 M8 G15 D13 N12 P6 Lil n F F 1B 3 2 2 3 2 2 3 1B F F M3 F2 Dll N7 B8 N9 D5 F14 011

, 3 F 2 2 3 2 3 2 2 F 3 B12 M6 06 A6 08 FIS 010 M10 D2 e F F 3 2 3 2 3 F F E2 L7 D14 L9 E14 a F F F F F I

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

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XX Batch F Fresh Fuel YYY Previous Cycle Core Location u

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B&k' Ravision 1 (5/13/80)

Figure 3-2. Enrichment and Burnup Distribution for Davis Besse 1, Cycle 2 8 9 10 11 12 13 14 15 2.63 1.98 2.96 2.63 2.96 2.63 2.96 3.04 H

14,621 11,188 9,588 14,583 13,918 14,437 6,750 0 2.96 2.63 2.96 2.63 2.96 2.63 3.04 K

7,720 12,991 9,139 14,109 7,083 14,262 0

_ 1.98 2.96 2.63 2.63 2.96 3.04 L

11,298 11,622 13,618 13,370 9,922 0 2.63 2.96 2.63 3.04 M

14,181 12,663 13,882 0 1

1.98 3.04 3.04 N

11,297 0 0 2.96 0

6,750 P

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x.xx Initial Enrichment xx,xxx BOC Burnup, MWt 3-3 Babcock & Wilcox

B&W Revision 1 (5/13/80)

Figure 3-3. Control Rod Locations for Davis Besse 1, Cycle 2 X

I A

3 4 7 4 C 1 5 5 1 D 7 8 2 8 7 E

1 6 6 1 F

4 8 7 6 7 8 4 c 5 3 3 5 g

W- 7 2 6 3 2 7 -I 6

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

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

! 5 8 Control

' 5 8 Control 7 12 control 8 8 APSRs Total # 61 3-4 Babcock & Wilcox

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Table 4-1. Fuel Design Parameters and Dimensions Batch IB Batch 2 Batch 3 Batch 4 Fuel assembly type Itark B-4A Mark B-4A Mark B-4A Mark B-4A No. of assemblies 12 61 60 44 l1 Fuel rod OD, in. 0.430 0.430 0.430 0.430 Fuel rod ID, In. 0.377 0.377 0.377 0.377 Flexible spacers, type Spring Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 Undensified active fuel length, in. 143.5 143.5 143.5 143.44 Fuel pellet OD (mean specifled), in. 0.3675 0.3675 0.3675 0.3697 Mean specified fuel pellet initial density, % TD 96.0 96.0 96.0 94.0

[ Initial fuel enrichment..wt % 235 U. 1.98 2.63 2.96 3.04 Estimated residence (max), EFPil 15,168 15,168 21,864 19,848 l1 Cladding collapse time, EFril >30,000 >30,000 >30,000 >30,000 n

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B&W Revisien 1 (5/13/80)

Table 4-2. Fuel Thermal Analysis Parameters Batch IB Batch 2 Batch 3 Batch 4 No. of assemblies 12 61 60 44 l1 Nominal pellet density, % ID 96 96 96 94 Pellet diameter, in. 0.3675 0.3675 0.3675 0.3697 Stack height, in. 143.0 143.5 143.5 143.44 Densified Fuel Parameters ("}

Pellet diameter, in. 0.3651 0.3651 0.3651 0.3648 Fuel stack height, in. 143.14 143.14 143.14 141.65 Average LHR @ 2772 MWt, kW/ft 6.14 6.14 6.14 6.21

, Fuel T at nominal LHR, 'T 1340 1340 1340 1355 LHR to [ fuel melt, kW/ft 20.4 (b) 20.4 20.4(b) 20.4 Note: Core average densified LHR at 2772 MWt is 6.16 kW/ft.

(*)Densification to 96.5% TD is assumed.

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Two batch 1B and five batch 3 FAs have LHR to centerline melt limits of 20.17 and 20.35 kW/ft, respectively. All FAs in batches 13, and 3 are limited to less than 20.17 kW/ft by RPS setpoints.

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B&k' R: vision 1 (5/13/80)

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of cycles 1 and 2. These val-ues were generated using PDQ075 for both cycles. Since the core has not yet reached an equilibrium cy. le, differences in core physics parameters are to be expected between the cycles. Figure 5+1 illustrates a representative rel-ative power distribution f or BOC 2 at full power (FP) with equilibrium xenon and group 8 inserted.

The critical boron concentrations for cycle 2 differ from those of the ref- l1 erence cycle 1 due to the difference in design cycle lengths. The hot full power (EFP) control rod worths are different because in cycle 2 the bank loca-tions and designations have changed from those of the reference cycle. Control rod worths are sufficient to maintain the required shutdown margin as indicated 1 in Table 5-2. The ejected rod worths in Table 5-1 are the maximum calculated values. It is difficult to compare maximum ejected rod worths between cycles since neither the rod patterns from which the rod is assumed ejected nor the

, isotopic distributions are identical. Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits presented in section 8.

The maximum stuck rod worths for cycle 2 are also different than those for t

cycle 1. The adequacy of the shutdown margin with cycle 2 stuck rod worths is shown in Table 5-2. The following conservatisms were applied for the shut-down calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth, t.

3. Flux redistribution penalty.

Flux redistribution was taken into account since the shutdown analysis was cal-culated using a two-dimensional model. The cycle 2 power deficit from HZP to HFP 1

at BOC is more negative than that for cycle 1 due to the more negative moderator 5-1 Babcock s.Wilcox

B&W R:visien 1 (5/13/S0) coefficient in cycle 2. The EOC power deficits are similar. The differential l1 boron and xenon worths are di'ferent at BOC because the reference (first) cycle had no plutonium present. In cycle 2, the presence of plutonium causes a spec-trum shift shich results in differences in these parameters. At EOC the dif-ferential boron and xenon worths are similar in both cycles. The effective delayed neutron fractions for cycle 2 show a decrease with burnup (also report-ed in the reference cycle 1).

5.2. Changes in Nuclear Design

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There are no significant core design changes between the reference and the

cyc?e 2 designs. The same calculational methods and design information were s

used to obtain the important nuclear design parameters. No significant opera-tional or procedural changes exist with regard to axial or radial power shape, xenon, or tilt control. The operational and RPS limits (Technical Specifica-tion changes) for cycle 2 are presented in section 8.

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B&W R:vicien 1 (5/13/80)

Table 5-1. Davis Besse 1 Cycle 2 Physics Parameters (*}

Cycle 1 Cycle 2 Cycle length, EFPD 433 248 Cycle burnup, mwd /mtU 14,360 8,240 Average core burnup -- EOC, mwd /mtU 14,360 17,069 initial core loading, atU 83,6 83.4 Critical horon -- BOC, ppm (no Xe)

KZP(b), group 8 (37.5% vd) 1,520 1,379 KFP, group 8 inserted 1,408 1,197 Critical boron -- EOC, ppm (eq Xe)

HZPl 456 422 HFPJgrup8(37.5%wd,eqXe) 129 149 Control rod worths -- HFP, BOC, % ak/k Group 6 2.00 0.98 Group 7 1.50 1.47 Group 8 (37.5% vd) 0.46 0.34 Control rod worths - HFP, EOC, I ak/k Group 7 1.12 1.56 Croup 8 (37.5% wd) 0.30 0.41 Max ejected rod worth -- RZP, % ak/k("}

! BOC 0.85 0.77 EOC 0.63 0.82 1 I

Max stuck rod worth - KZP, : Ak/k BOC 2.98 1.09 EOC 1.25 1.20 Power deficit, RZP to HFP, % ak/k BOC -1.00 -1.51 EOC -2.24 -2.24 Doppler coeff -- BOC 10-5 (ak/k/*F) 100% power (no Xe) BOC -1.25 -1.44 100% power (eq Xe) EOC -1.45 -1.53

!' Moderator coeff -- HFP,10-4 (ak/k/*F)

BOC (no Xe,1064 ppa, group 8 in) -0.06 -0.56

, EOC (eq Xe,17 ppa, group 8 in) -2.60 -2.73

, Boron worth -- EFP, pps/% ak/k BOC (1150 ppm) 99 111 4

r EOC (17 ppa) 102 98

t. Xenon worth -- EFP, % ak/k BOC (4 EFPD) 2.73 2.67 EOC (equilibrium) 2.71 2.78 s

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B&k' R:vicion 1 (5/13/80)

Table 5-1. (Cont'd)

Cycle 1 Cycle 2 Effective delayed neutron fraction -- HFP BOC 0.00690 0.00594 1 EOC 0.00514 0.00534 ,

(* Table 5-1 contains the cycle 1 values that are for the original 433 EFPD cycle 1 design.2 The modified cycle

. I had a design cycle length of 485 EFPD, however, it is now planned to refuel af ter 360 EFFD. The 360 EFPD cycle l1 1 length has been input to the cycle 2 calculations as presented in this report.

HZP denotes hoe zero power (S32F T**E); HFP denotes hot full power (584F core T""8).

(c) Ejected rod worth for groups 5 through 8 inserted.

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B&W Revision 1 (5/13/80)

Table 5-2. Shutdown Margin Calculation for Davis Besse 1 Cvele 2 BOC, EOC,

% ak/k  % ak/k Available Rod Worth Total rod worth, HZPI ") 7.09 7.47 Worth reduction due to burnup of poison material -0.05 -0.10 Maximum stuck rod, HZP -1.09 -1.20

"_ Net worth 5.95 6.17 Less 10% uncertainty -0.60 -0.62 Total available worth 5.35 5.55 Required Rod '4 orth y Power ceficit, HFP to HZP 1.51 2.24 Max allowable inserted rod worth 0.25 0.40 Flux redistribution 0.60 1.10 I

Total required worth 2.36 3.74

. Shutdown Margin Total available minus total required 2.99 1.81 Note: Required shutdown margin is 1.00% ak/k -

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(* HZP denotes hot zero power (532F T"#8); HFP denotes hot full power (584F core T, )..

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3&W Ravision 1 (5/13/80)

Figure 5-1. BOC (4 EFFD), Cycle 2 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, APSRs Inserted (a) 8 9 10 11 12 13 14 15 H 1.05 1.03 1.25 , 1.09 1.13 1.08 1.20 0.96 K 1.26 1.12 1.20 1.03 1.19 0.99 0.88 L 0.95 1.10 0.87 0.97 0.96 0.65 1

M 0.97 1.01 0.96 0.95 t

, N 0. 9.0 1.16 0.71 I

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R X Inserted Rod Group Number

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X.XX Relative Power Density

(*) Calculated results from two-dimensional pin-by-pin PDQ07.5 5-6 Babcock a Wilcox 1 -

B&W Ravicion 1 (5/13/80)

6. THERMAL-HYDRAULIC DESIGN The incoming batch 4 fuel is hydraulically and geometrically similar to the fuel remaining in the core from cycle 1. The thermal-hydraulic design evalua-tion supporting cycle 2 operations used the models and methods described in references 1, 6, and 7.

The flux / flow trip setpoint of 1.07 has been established for cycle 2 operation.

This setpoint and other plant operating limits based on DNBR criteria contain margin to the design minimum DNBR of 1.30 B&W-2 to account for the DNB rod bow penalty.

A rod bow penalty has been calculated according to the procedure approved in reference 8. The burnup used is based on the maximum FA burnup of the batch that contains the FA with the maximum radial x local peak. For cycle 2, this burnup is 23,020 mwd /mtU in a batch 3 assembly. The resulting net rod bow penalty is 1.0% after the 1% flow area reduction factor credit is included.

The rod bow penalty is accounted for by including DNER margin in trip set-points and operating limits.

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6-1 IBabcockuk VVilcox

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B&W R:visicn 1 (5/13/80)

Table 7-1. Comparison of Key Parameters for Accident Analysis FSAR, Predicted densif cycle 2 Parameter value value BOL Doppler coeff, 10-5, Ak/k/*F -1.28 -1.44 EOL Doppler coeff, 10-5, ak/k/*F -1.45I ") -1.53 BOL moderator coeff, 10-", ak/k/*F +0.13 -0.56 EOL moderator coeff, 10-4, ak/k/*F -3.0 -2.73

. All rod bank worth (HZP), % ak/k 10.0 7.09 1 Baron reactivity worth (HFP), pps/1% ak/k 100 111 Max ejected rod worth (HTP), Ak/k 0.65 0.39 Max dropped rod worth (HFP), I ak/k 0.65 0.20 Initial boron cone (HFP), ppm 1407 1197

(*) 1.77 x 10-5 ak/k/*F was used for steamline failure analysis.

Table 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates Allowable Core peak linear elevation, heat rate, i ft kW/ft 2 16.5 4 17.2 6 18.4  !

8 17.5 )

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Raviecd by TECo May 19, 1980

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 2 operation to account for changes in power peaking and control rod worths. In addition, changes were the result of the following:

1. The reduction of the fuel rod bow DNB penalty3 has permitted the relaxa-tion of certain operating limits.

2. The required quantity and concentration of boric acid necessary to reach a cold shutdown condition have been reviewed and found to be within pre-vious Technical Specifications Ibnits. Page B 3/4 1-2 of the Technical Specifications has been changed so that the basis is consistent with cycle 2 requirements.

3. All references to two-pump operation have been removed to avoid confusion since two-pu=p operation is not allowed by the operating license. Figures 3.1-c, 3.1-3d, 3.2-3a, and 3.2-3b have been removed.

4. The DNB limits in the bases have been changed from 1.32 to 1.30 to be consistent with the approvod limit of the B&W-2 correlation.

5. Sections 3.1.3.9 and 4.1.3.9 have been added to provide axial power shaping rod group limits for physical insertion. Special test exceptions in 1

Sections 3.10.1 and 3.10.2 have been modified in regard to these limits.

6. The high pressare limits appearing in the Technical Specifications, Table 2.2-1, have been previously submitted, but not acted on, in a letter from Toledo Edison Company to NRC, Serial Number 527, July 13, 1979, see attach-ment 1 to this report.

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7. Changes were made in the Quadrant Power Tilt Limits, Table 3.2-2, to account l 1

for the increased uncertainty in the detectors because of aging.

- Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated.

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B&W Revision 1 (5/13/80)

Table 3.2-2. Quadrant Power Tilt Limits Steady state Transient Maximum limit limit limit Measurement Independent QUADRANT POWER TILT 4.92 11.07 20.0 QUADRANT POWER TILT as measured by:

Synsneerical Incore Detector System 3.21 8.71 20.0 1 Power Range Channels 1.96 6.96 20.0 Minimum Incore Detector System 1.90 4.40 20.0 C'

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DAVIS-BESSE, UNIT 1 3/4 2-12 8-10a 4

e B&W Rrvision 1 (5/13/80)

Revised by TEco (5/19/80) .

l 3/4.10. SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.1. The group height, insertion and power distribution limits of Speci-fications 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.1.3.9, 3.2.1 and l1 r.

3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

., a. The THERMAL POWER is maintained s 85% of RATED THERMAL POWER,

b. The High Flux Trip Setpoint is 510% of RATED THERMAL POWER higher than

-[ ' the THERMAL POWER at which the test is performed, with a maximum setting i of 90% of RATED THERMAL POWER, and,

c. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in 4.10.1.2 below. I APPLICABILITY: MODE 1.

ACTION:

r .. With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7. 3.1.3.9, 3.2.1 or 3.2.4 are suspended, either:

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a. Reduce THERMAL POWER sufficiently to satisfy the ACTION requirements

, of Specifications 3.2.2 and 3.2.3, or

b. 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 />.

l SURVEILLANCE REQUIREMENTS

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4.10.1.1. The High Flux Trip Setpoint shall be determined to be set within the limits specified within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initiation of and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during PHYSICS TESTS.

, 4.10.1.2. The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 i .,

shall be performed at least once per two hours during PHYSICS TESTS.

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SPECIAL TEST EXCEPTIONS PHYSICS TEST LIMITING CONDITION FOR OPERATION 3.10.2. The limitations of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, and 3.1.3.9 may be suspended during the performance of l1 PHYSICS TESTS provided:

, a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and

b. The reactor trip setpoints on the OPERABLE High Flux Channels are set at 5 25% of RATED THERMAL POWER.

c. The nuclear instrumentation Source Range and Intermediate Range high startup rate control rod withdrawal inhibit are OPERABLE.

_ APPLICABILITY: MODE 2.

ACTION:

i With the THERMAL POWER > 5% of RATED THERMAL POWER, innediately open the con-trol rod drive trip breakers.

i SURVEILLANCE REQUIREMENTS

4.10.2.1. The THERMAL POWER shall be determined to be s 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2. Each Source and Intermediate Range and High Flux Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior t.o initiating PHYSICS TESTS.

me a 96 6* e DAVIS-BESSE UNIT 1 3/4 10-2 8-13b

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_e B&W Rsvision 1 (5/13/80)

Figure 8-18. Control Rod Core Locations and Group i Assignments - Davis-Besse 1, Cycle 2 f

A 4

l 8 4 7

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C 1 5 5 1 7 8 2 8 7 O

p...

6 1 E 1 6 4 7 6 7 8 4

? F 8 G 5 3 3 5 3l

~

H 7 2 6 6 2 7 I~

K 5 3 3 5 4 7 6 7 8 4 L 8 11 1 6 6 1 N 7 8 2l 8 7 1

0 1 5 5 1 P 4 7 4 R

I I 2 3 4 5 6 7 8 9 10 11 12 13 14 15 No. of Grouc Rods Purpose 1 8 Safety 2 4 Safety 3 5 Safety

-. 4 8 Safety 5 8 Control 6 8 Control 7 12 Contral L- 8 8 APSRs Total

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[. TECHNICAL SPECIFICATION FIGURE 3.1-4 Control Rod Core Loca-tions and Group Assign-ments - Davis-Besse 1, DAVIS-BESSE,I; NIT 1 3/4 1 31 8-36 Babcock & Wilcox