ML20028G782
| ML20028G782 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/30/1990 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20028G780 | List: |
| References | |
| BAW-2114, NUDOCS 9009040126 | |
| Download: ML20028G782 (58) | |
Text
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ARKANSAS NUCLEAR ONE, UNIT 1
- Cycle 10 Reload Report -
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B&W Fuel Company j88"28ssen8886 PDC
i ARKANSAS NUCLEAR ONE, UNIT 1
- Cycle 10 Reload Report -
f B&W Fuel Company 9009040126 900808
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PDR ADOCK 05000313 P-PDC
'BAW-2114
- June ' 1990 =-
- ,4 P
??.UWSAS NUCIZAR ONE, UNIT 1
- Cycle 10 Reload Report -
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' 19 B&W Fuel Caupany P. O. Box 10935 Lynchburg, Virginia 24506-0935 l
B&W FuelCompany m
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COEDES Wge 1.
INIR3DUCTIQ4 AND SDtORY...........
14 2-1 2.
OPEPATDC HISKEY
.+........,.........
3-1 3.
GD4ERAL IESCRIPTIQ1 4-1 4.
FUEL SYSTD4 DESIG4..........
4-1 4.1.
fuel Assembly Mechanical Desigt 4-1 4.2.
Ibel Rod Dasip 4-1 4.2.1.
Clas_:urg Collapse 4-2 4.2.2.
Claddits Stress 4-2 4.2.3.
Claddirg Strain 4.3.
'1hermal Design.......................
4-3 4-3 4.4. ' Material Design 4.5.
Operating Experience....................
4-3 5.
NUCI. EAR IESIG4..........................
5-1 5-1 5.1.
Physics Characteristics 5-1 5.2.
Analytical Input.
5.3.
Changes In Nuclear Design.................
5-1 6-1 6.
'INERMAleHYtEAULIC DESIGi.
7*1 7.
ACCIDENT AND 'IRANSIDC ANAL'iS16.
7-1 7.1.
General Safety Analysis 7-2 7.2.
Accident Evaluation 8.
PIOPOSED )CDIFICATIONS 'IO TECHTICAL SPECIFICATIQ1S.
8-1 9-1
-9.
STARIUP PROGRAM - THYSICS TESTDG 9-1 9.1.
Precritical 'Ibcts 9-1 9.1.1.
Control Red Trip 'Ibst 9-1 9.1.2.
RC Flcw......................
9.2.
Zero Ptwer Physics Tests.................,.
9-1 9.2.1.
Critical Boron Co:rentration............ 9-1 9.2.2.
Te:perature Reactivity Coefficient.
9-2 9.2.3.
Canttel Rod Group / Boron Reactivity Worth......
9-2 9.3.
Ptwer Escalation 'Ibsts...................
9-3 9-3 9.3.1.
Core Symmetry Test..
111 B&W FuelCompany
Mah (Cbnt'd)
Page 9.3.2.
Cbre Power Distribution Verification at Intermediate Pcuar Inval (IPL) and 100% FP With Ncrninal Control Rod Position..........
9-3 9.3.3.
Incore vs b: core Detector Imbalance Correlation Verification at the IPL.........
9-4 9.3.4.
'huperature Reactivity Coefficient at -100% FP.
9-4 9.3.5.
Power Doppler Reactivity Coefficient at ~100% FP..
9-5 9.4.
Procedure for Use if Acceptance Critaria Not Met......
9-5
- 10. REFERDiCES.
10-1 List of ah Table 4-1.
Nel Design ParNneters and Dimensions 4-4 4-2.
Fuel 'Iharmal Analysis Paraneters................
4-5 4-3.
Operati.rg Experience......................
4-6 5-1.
Physics Parameters for NO-1, Cycles 9 and 10 5-3 5-2.
Shutdown Margin Calculations for NO-1, Cycle 10..
5-5 6-1.
MsMe Design Conditions, Cycles 9 ard 10.
6-2 7-1.
Ccmparison of Cycle 9 and Cycle 10 Accident h 7-4 7-2.
Ctmparison of Key Parameters for Accident Analysis.......
7-5 7-3.
Ecurding Values for Allowable IDCA Peak Linear Heat Rates 7-5 Iist of Fiaurus Figure 3-1.
Core Loading Diagram for N O-1, Cycle 10...
3-2 l
3-2.
Enrichment and Eurnup Distribution, NO-1 Cycle 10 off 420 EFPD Cycle 9........................
3-3 3-3.
Control Rod IIx:ations ard Group Designations for 3-4 N O-1, Cycle 10 3-4.
IBP Enrichment and Distribution, N O-1, Cycle 10........
3-5 4-1.
Mark BB Fuel Rod Design Features................
4-7 5-1.
NO-1, Cycle 10, IOC (4 ETPD) ho-Dimensional Relative Power Distribution - Nll Power Equilibrium Xenon, Normal Rod Positions 5-6 6-1.
Core Protection Safety Limits - NO-1, (%ch Spec Figure 2.1-2).........................
8-4 8-2.
Protective System Maxinun Allcwable Setpoints
- NO-1, (Tech Spec Figure 2.3-2) 8-5 8-3.
Rod Position Setpoints for Four-Pump Operation fran 0 to 30 +10/-0 EFPD - NO-1, Cycle 10
(%ch Spa Figure 3.5.2-1A).....
8-8 8-4.
Rod Position Setpoints for Four-Punp Operation fran 30 +10/-0 to 3351 10 EFPD - NO-1, Cycle 10 (%Ch Spec Figure 3.5.2-1B 8-9 iv B&W FuelCompany i
YAat of Ficrurts (Cont'd)
Figure 8-5.
Rod Position Setpoints for Four-Purp Operation after 335 2 10 EFPD - NO-1, Cycle 10
(%ch Spec Figure 3. 5. 2-1C)................... 8-10 8-6.
Rod Position Setpoints for 'Ihree-Punp Opera *lon Tzun 0 to 30 +10/-0 EFPD - NO-1, Cycle 10
(%ch Spec Figure 3. 5.2-2A)................... 8-11 8-7.
Rod Position Setpoints for 'Ihree-Punp Operation Fran 30 +10/-0 to 335 10 EFPD - NO-1, Cycle 10 (%ch Spec Figurs 3.5.2-2B).............. 8-12 8-8.
Rod Position Setpoints for Three-Punp Operation After 335 t 10 EFPD - NO-1, Cycle 10 (Totri Spec Figure 3. 5. 2-2C)................... 8-13 8-9.
Rod Position Setpointo for 'No-Punp Operation Fran 0 to 30 +10/-0 EFPD - NO-1, Cycle 10
(% ch Spec Figure 3.5.2-3A)................... 0-14 8-10. Rod Position Setpoints for 'No-Punp Operation Fran 30 +10/-0 to 335 10 EFPD - NO-1, cycle 10 (%ch Spec Figure 3.5.2-3B).............. 8-15 8-11. Rod Position Setpoints for Two-Punp Operation After 335 i 10 EFPD - NO-1, Cycle 10 8-16
(%ch Spec Figure 3.5.2-3C).
8-12. Operational Power Imbalance Setn ints for Operation Fran 0 to 30 +10/-0 EFPD - ?.Pel, Cycle 10
(% ch Spec Figure 3.5.2-4A)...................
8-17 8-13. Operational Power Inbalance Setpoints for Operation Frun 30 +10/-0 to 335 10.TPD - NC-1, Cycle 10 (%ch Spec Figure *.5.2-4B)..............
8-18 8-14. Operational Power Imbalance.'etpoints for Operation After 335 10 EFPD - NO-1, Cycle 10 (Tech Spec Figure 3. 5. 2-4C)...................
8-19 8-15. LOCA Limited Maxintun A11cuable Linear Heat Rate - NO-1 Cycle 10 (Tech Spec Figure 3. 5. 2 -D )............................
8-20 V
B&W FuelCompany
1.
IITITOCUCTIOtt NtD SUl4%RY
'Ihis report justifies the operation of the tenth cycle of Arkansas Ituelear One, Unit 1 (No-1) at the rated core power of 2568 }Wt -
Include:1 are the required analyses as outlined in the US!mC hment, " Guidance for Pr W License Aiim.b ats Relating to Refueling," June 1975.
'Ib reupport cycle 10 operation of NK>-1, this report er. ploys analytical techniques and design bases established in reports that have been subnitted to and accepted by the US!mC and its predecessor, the US/IC (see references).
'Ibe cycle 9 ard 10 reactor parameters related to power capability are sumarized briefly in section 5 of this report.
All of the accidents analyzed in the PSAR1 have been reviewed for cycle 10 operation.
In those cases where cycle 10 characteristics were conservative compared to those analyzed for previous cycles, new accident analyses were not performed.
'Ibe 'Iuchnical Specifications have been reviewed, and the modifications required for cycle 10 operation are justified in this report.
Based on the analyses perfonned, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Dnertjency Core Cooling Systems, it has been concitded that NO-1 can be operated safely for cycle 10 at a rated power level of 2568 MWt.
1-1 B&W FuelCompany i
2.
OPEPATI1G HIS'IORY 2e refererce cycle for the nuclear and therral-hydraulic analyses of Arkansas Nuclear One, Unit 1 is the currently operating cycle 9.
Cycle 9 began in Wr, 1988 and operated at 100% power for approximately 30 effective full power days (EITO).
At that time, backflow through the HPI line led to a shutdown of approximately two months after which the unit was limited to 50% until analyses were available to justify a power increase.
Except for one minor shutdown, the unit cperated for about 25 EFPD at 50%
power arxi followirg a (two week) outage, restarted to 80% TP.
Operation at 80% lasted about 35 EFPD after which power was limited to 75% due to operation with only three reactor coolant punps.
mree punp operation limited power to 75% for approximately 90 EFPD when a brief outage was used to repair the inoperable RC purp.
SeI
. of the cycle was to be operated at 80% Full Power.
We cycle 10 design is based on a design cycle 9 lerrJth of 420 EFPD.
No anomalies occurred during cycle 9 that would adversely affect fuel performance during cycle 10.
2-1 B&W FuelCompany
t 3.
GDEPAL DESCRIPTION he MD-1 reactor core is described in detail in section 3 of the Arkansas Nuclear One, Unit 1, Final Safety Analysis Report (ISAR).1 he cycle 10 core contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods,16 m Lul rod guide tubes, and one iricore instrument guide tube.
The fuel is ocmprised of dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4.
We fuel assemblie in all batches have an average nominal fuel loadig of 463.6 kg of uranium.
We undensified ncaninal active fuel legths, theore-tical densities, fuel and fuel rod dimensions, and other related fuel para-matars are given in Table 4-1 for all fuel assenblies.
Figure 3-1 is the fuel shuffle diagram for NO-1, cycle 10.
We initial enrichments of batches 6E,10B,11 and 12 are 3.19, 3.35, 3.45 ard 3.49 wt %
U-235, respectively.
Se batch 6D assembly, all of batch 9B, and 4 of the twice-burned batch 10 assenblies will be discharged at the erd of cycle 9.
%e centar location will contain a batch 6 assenbly discharged at the end of cycle 5 (designated 6E).
%e romaining 60 twice-burned batch 10 assenblies (designated 10B) will be shuffled to new locations, with 12 on the core perighery.
We 60 once-burned batch 11 assemblies will be shuffled to new locations, and the 56 fresh batch 12 assemblies will be loaded in a symetric checkerboard pattern throughout the core.
Figure 3-2 is an eighth-core map showing the assenbly burnup and enrichment distribution at the beginnig of cycle 10.
Reactivity is controlled by 60 full-length Iq-In-Od control rods, 48 burnable poison red assemblies (BPPAs), and soluble boron shim.
In addition to the full-length control rods, eight Inconel axial power shapig rods (gray APSPs) are provided for additional control of the axial power distribution.
We cycle 10 locations of the 68 control reds and the group designations are indicated in Figure 3-3.
We core lccations and the group designations of the 68 oontrol rods for cycle 10 are the same as those of the reference cycle. We locations and enrichments of the BPRAs are shcun in Figure 3-4.
3-1 B&W Fuel C=aany
Figure 3-1.
Core loading Diagram for ANO-1 Cycle 10 x
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10 11 12 13 14 15
(
l XXX Batch mte s r Denotas Tresh T.aal XXK Prev 2cus Cbru Izx:sticn r
3-2 B&W FuelCompany t
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Tigure 3 2.
Enrichment and Burnup Distributien.
ANO 1 Cycle 10 off 420 ETPD Cycle 9 8
9 10 11 12 13 14 15 3.19 3.45 3.35 3.45 3.35 3.45 3.45 3.45 H
20,770 17,838 32,917 13,461 32,925 13,460 17,280 17,806 3.35 3.49 3.35 3.49 3.35 3.49 3.45 K
22,295 0
26,775 0
22,788 0
17,146 3.35 3.49 3.35 3.49 3.45 3.45 L
34,293 0
25,520 0
14,967 17,274 3.35 3.49 3.45 3.45 M
33,075 0
17,588 17,564 3.35 3.49 3.35 N
33,795 0
31,648 3.35 0
22,288 P
R X.XX Initial Enrichment, vt t U 235 XX XXX BOC Burnup.
MVd/mtU B&W FuelCompany 3,3
~ _ _..
i Figure 3-3.
Control Rod Locations ar.d Group Designations for ANO-1 Cycle 10 l
Fuel Transfer Canal A
B 1
6 1
I C
3 5
5 3
D 7
8 7
8 7
E 3
5 4
4 5
3 F
1 8
6 2
6 8
1 J
G S
4 2
2 4
5 i
H W-6 7
2 2
7 6
-Y K
5 4
2 2
4 5
1 L
1 8
6 2
6 8
1 M'
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4 4
5 3
N l
7 8
7 8
7 o
l 3
5 5
3 P
l l
1 6
1 R
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Z 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 J
X Group Number Group No. of Rods Function i
1 8
Safety 2
8 Safety 3
8 Safety 4
8 Safety 5
12 Control 6
8 Control 7
8 Control 8
8 APSRs 3-'
B&W FuelCompany
I Pigure 3 4 LBP Enrichment and Distribution, ANO 1 Cycle 10 8
9 10 11 12 13 14 15 H
K 0.80 1.10 0.50 L
0.80 1.10 1.10 M
1.10 1.10 N
1.10 1.10 0
1.10 P
0.50 R
X.XX LBP Concentration, vt t B C in Al 0 )
4 23 B&W FuelCompany 35
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4.
IUEL SYSTD4 DESIGi
' 4.1. Ibel Assembiv Mechanical Desian he types of i4 assenblies ard pertinent fuel design parameters for AIC-1, g
cycle 10 are lirled in Table 4-1.
Bat & 6E and 10B are the Mark-B4 design.
Bate 11 is the Mark-B6 design, ard Batch 12 is the Mark-B8 design.
i Mark-B8 Ibel Assembly Descriptien:
l.
2e Mark-B8 fuel assembly is an imprwed Mark-B6 fuel assembly with design
. features to allcw for high burnup, and to prwide protection against debris fretting damage to the fuel rods. % prwide for high burnups, the lower and l
fitting has been shortened by approximately 0.7 inches.
me guide and instrument tubes were largthened by the same amount, and the fuel rod l
lengthened by.4 in 6 es. mis prwides additional gas volume ard grwth room i
L for the fuel red.
2 protm?. against debris irduced fretting failure of the fuel rod the followirg design changes were made.
Se lower and plug solid portion was extended in length.
Se icwer spacer grid location was &M so that the solid and plug extends through the lcwor spacer grid.
We intention of the j
design is to trap any debris capable of fuel rod frutting belcw the bottom
{
spacer grid where the solid 1cuer and plug will prevent failure.
A canpariscn of the design features of the various Mark-B type fuel rods is shcun in Figure 4-1, 1
Forty Eight BIRAs will be used with the 56 batch 12 fuel assemblies.
4.2. Fuel Rod Desian me fuel rod design and mechanical evaluation are di='Je=ai here.
4.2.1. Claddina Collaase l
'L' e operating power history for the most limiting fuel assenbly was ckY' ermined for each of the three previously burned fuel batches. Se history I
for each batch was acrpared to that used in the generic creep collapse analysis.
Batches 10B ard 11 slightly eW_ the generic envelope. A new 4-1 B&W FuelCompany
envelope was formlated and new creep collapse analyses run.
Both the new specific analyses arx1 the generic analyses are MW on the nethod of referen:e 2.
%c analysis predicted a creep collapse life longer than 35,000 DTH.
mis is lorger than the mximm batch residence in cycle 10 ktdch is 29,800 DMI for batch 10B.
For Batch 12 the creep collapse analysis followed the method frun reference 3.
We operational conditions and mechanical ctaracteristics of the batch 12 fuel assemblies was cutparmd to an envelope formlated by &TC (reference 3) ard approved by the NRC (reference 4).
All values of the Mark-BB fuel assemblies ars WW by the con +-ding parameters of the limiting envelope.
Sczne as-built data for the Mark-BB assemblies (presently unavailable) were approximated frun as-ballt values of past %7C fuel and then cartpared to the limiting envelope. his is reasonable as the tolerances affecting these as-built values have not changed frun past BWFC fuel designs.
We creep collapse life of the batch 12 fuel rtds baM on reference 4 is 55,000 mwd /mtU.
This is greater than the maximum projected end of cycle burnup of 16,061 mwd /mtU for batch 12.
4.2.2. ClarMim Stage Se stress parameters for the fuel rod designs are enveloped by conseIvative fuel rod stress analyses.
We same method was used for analysis of cycle 10 that had been used on the previous cycle.
%e stress margins are in excess of 11.2%.
4.2.3. ClaMim Strain he fuel rod design criteria specify a limit of 1% on claddirg plastic tensile circumferential strain.
We fuel pellet is designed to ensure that plastic cladding strain is less than 1% at design local pellet burnup and heat generation rate.
Se design values are higher than the worst-case values A!O-1 cycle 10 fuel is expected to experience. W e strain analysis is conservatively based on enveloping the upper tolerance values for the fuel pellet diameter and density ard the loser tolerance limit for cladding inside diameter.
4-2 B&W FuelCompany
7 l
I 4.3.
'thar=1 Desian All fuel assemblies in the cycle 10 core are themally similar. 'Ibe design cf the batch 12 Mark BB assemblies is such that the thermal performance of this fuel is ecpivalent to the fuel design used in the remairder of the core.
me analysis for all fuel was perfomed with the TACD2 ocde as described in reference 5.
Nominal urdensifia:1 input parameters used in the themal analysis are presented in Table 4-2.
Densification effects were accountad for in 'IAcc2.
Se results of the thermal design evduation of the cycle 10 core are sumarized in Table 4-2.
Cycle 10 ooru protection limits are haami on t.
linear heat rate (IHR) to centarline fuel melt limit of 20.5 ):W/ft as deter-mined by the TACD2 code.
me =vt== fuel assembly burnup at EOC 10 is predicted to be less than 47,000 )Nd/mtU (batch 10B).
De fuel rod internal pressures have been evaluated with 'IACD2 for the highest burnup fuel rods and are predicted to be less than the naminal reactor coolant pressure of 2200 psia.
4.4. Material Desian me chemical carpatibility of all possible fuel-cladding-coolant-assembly interactions for batch 12 fuel assemblies is identical to those for previous fuel assemblies because no new materials were introduced in the batch 12 fuel assemblies.
4.5. Ocaratinct Exoerience naWk & Wilcox operating experience with the Mark B 15x15 fuel a-mbly has verified-the adequacy of its design, me acx:um11ated operatirg experience for eight B&W 177 fuel assembly plants with Mark B fuel is shown in Table 4-3, 4-3 B&W FuelCompany
i i
Table' 4-1. - Fuel Desian Par==atars ard Dinensions
[
Batcti 6E 10B 11 12 i
Fuel assembly type MK-B4 MK-B4 MK-B6 MK-B8 Number of assemblius 1
60 60 56 Ibel rod OD ncrtinal, 0.430 0.430 0.430 0.430 in.
Tbel Rod ID nominal 0.377 0.377 0.377 0.377 in.
Undensified active 142.25 141.8 141.8 141.8 fuel length, in.
Fuel pellet OD,
.3695
.3686
.3686
.3686 (mean), in.
Ibel pellet initial 94.0 95.0 95.0 95.0 density, (Nam), % TD Initial fuel enrictnant 3.19 3.35 3.45 3.49 wt.% U235 Average burnup, BOC 20,770 28,337 16,604 0
Mid/mtU.
Mav4== assembly 32,209 46,301 30,735 16,061 burnup, B3C Mid/mtU.
Exposure time, IDC 28,700 29,800 19,200 9,100 EITH.
Claddirg collapse
>35,000
>35,000
>35,000 NA l
time, EFTH.
Cladding Collapse NA NA NA 55,000 R1rnup, Mid/IrJJ.
i l
4-4 B&W FuelCompany
r-i-
Table 4-2.
n=1 h Analysis Par== tars Batch 6E Batch 10B Batch 11 Batch 12 f
No. of assemblies 1
60 60 56 Initial density, % TD 94 95 95 95 Initial pellet OD, in 0.3695 0.3686 0.3686 0.3686 Initial stack height, in 142.25 141.80 141.80 141.80 l
Enrichment, wt % U-235 3.19 3.35 3.45 3.49 Naminal linear heat rate 5.73 5.74 5.74 5.74 at 2568 mt, W/ft(a) i tam *> ha=ad Prwi4ctions Average fuel tanparature i
at ncuninal IRR, F (ICL) 1406 1400 1400 1400 MininIm DR to malt, W/ft 20.5 20.5 20.5 20.5 Core average DR = 5.74 W/ft (a) Based on a ncminal stack height i
f I
l 4-5 B&W FuelCompey 1
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i Table 4-3.
Ocaratirn Emerience j
i Cunulative Current Max TA burnuo,14*i/mtU(a) electric
-f Reactor cvele_
Ircore Discharged output.MHb(b) j Coonee 1 12 40,595 58,310 84,346,419 Occnee 2 11 34,646 42,820 78,744,917 Ooonee 3 12 35,594 42,740 78,231,767 mtwe Mile Island 7
33,966 33,863 47,186,490 Arkansas Nuclear One,-
9 34,972 57,318 63,712,046 Unit 1 Rancho Seco 7
34,123 38,268 43,215,399 Crystal River 3 7
38,793 35,350 50,831,632 Davis-Besse 6
33,690 40,300 38,787,158 (a) As of %2 31,1989.
l (b)As of M r 31, 1989.
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5.
!UCIIAR DESIGN I
5.1. mvsics daractaristics I
Table 5-1 lists the core physics parameters of design cycles 9 and 10.
Se j
values for both cycles were calculated with the N00 DIE code.
Figure 5-1 6
I
?
illustrates a *r_:itative relative pcuer distribution for the beginning of cycle 10 at full pcuer with equilibrium xenon and ncuninal red positions.
Se differences in feed enrichment, BPPA loading, and shuffle pattern caused little charge in the physics parameters between cycles 9 and 10. Calculated i
ojected rod worths and their adherence to critaria are considered at all times in life and at all power levels in the developnent of the rod position
-l limits presented in sectior. 6.
Se==v4=nn stuck rod worths for cycle 10 are i
l less than for cycle 9 before APSR pull and greater at and of cycle.
All safety critaria associated with these worths arn met.
Se adequacy of the shutdown margin with cycle lo stuck tod worths is denonstrated in Table 5-2.
l Se followirg conservatisms were applied for the shutdown calculations I
1.
Poison material depletion allowance.
2.
10% uncertainty on net red worth.
3.
Flux rodistribution penalty.
Flux redistribution was accountad for since the shutdown analysis was l
l calculatM using a two-dimensional model.
Se reference fuel cycle shutdown l
margin is presented the Alc-1 cycle 9 reload report.7 5.2. Analvtical Inoat I
2e cycle 10 incore measurement calculation constants to be used for conpu-tirg core power distributions were prepared in the same manner as those for the reference cycle.
5.3. Garnes in M.1 clear Desian Se only design charge for cycle 10 is the shorter design cycle lergth.
l Se gray APSRs will be withdrawn frun the core near the end of cycle 10 (335 EFPD) where the stability and control of the core in the feed-and-bleed node with APSRs removed has been analyzed.
Se calculated stability irdax at 339 5-1 B&W FuelCompany
b~
I EFPD without APSRs is -0.066 h-1 which demonstratas the axial stability of the oors.
'Ihe calculatioral methods used to obtain tlw ligertant rax: lear design parameters for this cycle were the same as those used for the reference cycle. 'Ibe operatirg limits ('nu:hnical Specifications charges) for the reload cycle are given in section 8.
5-2 B&W FuelCompany
in 1
l i
i l
't.
c l[
Table 5-1.
mvslas Par =*m for AND-1. Cvelas 9 ard 10(a) cvele 9(b) cyei io(c)
. q-Cycle length, DTD 420 380
, m Cycle burnup, 1H:l/mtU 13,143 11,892 Average core burnup - EOC, PSt1/mtU 27,271 27,244 Initial core loadirg, atU 82.1 82.1 Critical boron - ICC, ppn (no Xe)
}EP(d), group 8 inserted 1552 1548
}FP, group 8 inserted 1379 1373 Critical boren - IDC, ppu (eq Xe) i HZP, groqp 8 out 539(e) 274 IFP, group 8 out O(f) 7 l
i control rod worths - IFP, BOC, % AP/k Group 6 1.11 1.08 Grcup 7 0.98 0.88
^
Group 8 (may4==)
0.19 0.20 control rod worths - }FP,' IDC, k AP/k
)
Group 7 1.05 0.96 Max ejected rod worth (Ir10) - }CP, % AP/k l
BOC, group)s 5-8 ins 0.35 0.29 335 EFPD(v, groups 5-8 ins 0.41 0.29 EOC, groups 5-7 ins 0.41 0.33 Max stuck rod worth (M-13) (h)' - FCP, % AP/k 100, gr q(v),1-8 ins 1.49 1.30 335 EFPD groups 1-8 ins 1.47 1.40 EOC, groups 1-7 ins 1.42 1.45 Power deficit, IEP to IFP, % AP/k BOC 1.60 1.61 EOC 2.35 2.31 Doppler coeff - IFP, 10-5 (3pjyjoy)
BOC (no xe)
-1.59
-1.59 EOC (eg Xe)
-1.86
-1.80 5-3 B&W FuelCompany
Table 5-1.
(Cont'd)(a)
Cvele 9(D)
Cvele 1Q(C) l Moderator coeff - HFP,10-4 (LP/P/ F) 0 B3C (no Xe, crit ppm, group 8 ins)
-0.58
-0.60 D3C (eq Xe, O ppu, group 8 out)
-2.82
-2.81 Boron worth - HFP, ppt /% AP/k B3C 130 129 EOC 111 111 Xenon worth - HFP, % AP/k BOC (4 EFPD) 2.56 2.55 IDC (equilibrium) 2.71 2.69 Effective delayed neutron fraction - HFP BOC 0.0062 0.0061 D3C 0.0052 0.0053 (a) Cycle 10 data are for the conditions stated in this report. 'Ihe cycle 9 core conditions are identified in refere:x:e 7.
(b)Bamai on 440 EFPD at 2568 }Wt, cycle 8.
(c) Based on 420 EFPD at 2568 PWt, cycle 9.
(d)HZP denotes hot zero power (532F Tavg); HFP denotes hot full power (581F Tavg)-
(*) Calculated with no xenon for cycle 9.
(f)At HFP conditions, O ppm occurs at 411 EFPD.
(9) Calculated at 360 EFPD for cycle 9.
(h)'Ihe maximm worth stuck ro:1 was in location N-12 for cycle 9.
5-4 B&W FuelCompany
I r
h hie 5-2.
Shttdmn Marcin Niculatien for AND-1. Cvele 10
- B3C, 335 LTPD, 380 EFPD,
% Ak/k
% Ak/k
% Ak/k Available Pa$ Worth
'!btal rod worth, HZP 8.351 8.910 8.855 Worth reduction due to poison natarial burnup
-0.100
-0.100
-0.100 Haximum stuck rod, HZP
-1.303
-1.398
-1.447 Net Worth 6.948 7.412 7.308 Imss 10% uncertainty
-0.695
-0.741
-0.731 Total available worth 6.253 6.671 6.577 w h wi Red Worth Power deficit, HFP to HZP 1.605 2.239 2.308 Allowable insertad rod worth 0.246 0.370 0.365 Flux redistribution 0.381 0.60f M il
'Ibtal required worth 2.232 3.214 3.234 Shutdown margi.n (total available worth minus total required worth) 4.021 3.457 3.343 HGE: 'Ibe required shutdown margin is 1.00% ak/k.
5-5 1
B&W FuelCompug
i i
Figure 5 1.
ANO.1 Cycle 10, BOC (4 EFFD) Two Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions 8
9 10 11 12 13 14 15 H
0.99 1.12 0.95 1.19 0.98 1.25 1.11 0.61 K
1.07 1.29 1.06 1.30 1.14 1.20 0.59 8
L 0.99 1,29 1.09 1,29 0.96 0.44 M
0.99 1.28 1.11 0.67 N
0.92 1.11 0.35 0
0.50 P-R l
8 Inserted Rod l
group No.
X.XX Relative Power Density B&W FuelCompany 56
l l
l 6.
'IHER!%Ir-NYIPAULTC DESIGi 1
'Ihe thermal-hydraulic design evaluation supportirg cycle 10 operation utilized the methods 'ard models described in references 7,
8 and 9 as supplemented by reference 10, which inplements the BWC (refemnoe 11) OF oorrelatico for analysis of the Zircaloy grid fuel assembly.
'Ibe inocning j
batch 12 fuel is hydraulically and W trically similar to the fuel remaining in the core frum previous cycles.
Cycle 10 is the secord cycle in the transition frce the Mark-B Inconal-grid fuel design to the Mark-BZ, Zircaloy grid fuel design.
'Ihe cycle 10 core is ocrrprised of 116 Zircaloy grid fuel assemblies, 48 of which contain BIHAs, and 61 Mark B fuel assemblies.
'Ihirteen (13) of the Mark-B fuel assemblies and 48 of the Zircaloy grid fuel assemblies have open, or unplugged, control rod guide tubes.
'Ihe oors bypass flow, which is depardent on the number of open control rod guide tubes, is 8.5% for thia configuration.
'Ihe Zircaloy grid fuel assemblies exhibit a slightly higher pressure drop than the Mark-B assemblies.
'Ihis tends to cause some coolant flew diversion frcn the Zircaloy grid assemblies to the Mark-B fuel and creates the need to consider o " transition core penalty."
'Ibe referunoe analysis for cycle 10 thermal-hydraulic design is the same as that used for cycle 9 and censiders a full core of Zircaloy grid fuel assablies with a bypass flew of 8.8%.
A cycle-specific analysis, which j
l l
modeled the actual cycle 10 core configuration ard bypass flow value, has
]
been performed to demonstrate that the reference analysis remains applicable l
ard a transition core pena 1T is not r===an.
Table 6-1 provides a sumary l
-of the DiB analysis parametent for cycles 9 and 10.
1 No rod bow penalty was considend in the cycle 10 analysis as justified by i
reference 12.
i l
l l
6-1 B&W FuelCompsy j
Table 6-1.
MavMnn Desian Corditions. Cveles 9 ard 10 Ovele 9 Cycle 10 Design power level, MRt 2568 2568 System pressure, psia 2200 2200 Reactor coolant f1w, gpre 374880 374880 Core bypass fl w, % (a) 8.8 8.8 INBR modeling Crossf1 w Crossf1w Reference design radial-local pwer peaking factor 1.71 1.71 Referenm design axial fim shape 1.55 cosine 1.65 cosine Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 F1w area 0.97 0.97 Active fuel length, in. (b) 141.8 141.8 Avg heat flux at 100% power, 103 2
Btu /h-ft 174 174 Max heat flux at 100% power, 103 2
Btu /h-ft 492 492 OF correlation IHC IMC OF correlation INB limit 1.18 1.18 Mininum INBR at 112% power 1.77 1.77 at 102% power 2.01 2.01 (a)Used in the analysis.
(b) Cold nominal stack height.
(c)'Ihis represents initial condition ENBR for accident analyses.
6-2 B&W Fuelcampany
l 7.
ACCIDDE AND 'IRANSIDE ANALYSIS
,".1.
General Safety Analysis Each ISAR accident analysis has been examined with respect to changes in cycle 10 parameters to determine the effect of the cycle 10 reload and to ensure that thermal performance during hypothetical transients is not degraded.
We effect of fuel densification on the FSAR accident results has been evaluated anf. are reported in reference 13.
Since batch 12 rel,oad fuel assemblie.4 contain fuel rods whose theoretical density is higher than those considered in the reference report, the canclusions in that reference are still valid.
2e radiological dose consequences of the accidents presented in Chapter 14 of the updated ISAR were re-evaluated for this reload report except for the waste gas tank rupture. We waste gas tank rupture was not reevaluated since
%chnical Specification 3.25.2.5 controls the maximum tank inventory on the basis of Xe-133 equivalent curie content such that the analysis of the event is not cycle depandent.
me evaluation of the remaining events was made in order to incorporate more current plant data as well as the information in the updated ISAR.
All of the cycle 10 accident doses are haaM on radionuclide sources calculated for the actual cycle 10 core design and irradiation history.
Table 7-1 shows a canparison between cycle 9 and cycle 10 doses for the Chapter 14 accidents that result in significant offsite doses.
We difference between cycle 9 and cycle 10 LOCA and MHA doses resulted frcu the adoption of updated iodine species fractions per Reg Guide 1.4 Rev. 2 and inclusion of effects of throttled reactor building spray.
2 e radiological doses frcm all of the accidents evaluated with the specific nuclide inventory frca cycle 10 are lower than the NRC acceptance criteria of NURD:i-0800, and thus are within acceptable limits.
7-1 B&W FuelCompany
l 7 2.
Accident Evaluaticm r
The key parameten that have the greatest effect on determining the autocane of a transient can typloally be classified in three major areast oors thermal parameters, themal-hydraulic parameters, and kinetics paratr stars, including the reactivity faa**ck coefficients and control rod worths.
The core themal properties uord in the FSAR accident analysis were design operating values based on calculational values plus uncertainties.
Themal parameters for fuel batches 6E,10B,11, and 12 are given in Table 4-2.
The cycle 10 thermal-hydraulic nav4=nu design conditions are ocupared wita the previous cycle 9 values in Table 6-1.
These parameters are oczmon to all the accidents ocmsidered in this report.
The key kinetics parameters frun the PSAR and cycle 10 are ocmpared in Table 7-2.
A generic lossef-coolant accident (IOCA) analysis has been performed for the NW 177-FA lowered loop nuclear steam supply (NSS) system using the Final Acceptance criteria Emergency core cooling System (EoCS) Evaluation Model (reported in BAW-10103A, Rev.314), updated with an upgraded fuel perfomance model (reported in BAW-177515) and the B&W modified version of FIZCSEP 16 and BAW-10104PA, Rev. 517).
These analyses are f
(reported in BAW-1915PA generic, since the limiting values of key parameters for all the B&W plants in this category were used.
Ruthermore, the ocanbination of averags fuel tenperatures as a function of linear heat rate and lifetime pin pressure data l
ased in the generic IDCA linear heat rate limits analysis is conservative ocmpared to those calculated for this reload.
Table 7-2 shows the boundirg l
v3.iues for maximIm allowable LOCA linear heat rate limits for Arkansas Nuclear One - Unit 1 (No-1) cycle 10 fuel as a function of burnup. 7he LOCA linear heat rate limits for beginning et cycle operation include the ocznbina.1 effects of the NURD3-0630 cladding swell aM rupture model, use of the BWC OIF oorzelation, ra%*4 fuel rtd prepressure, and inplementation of the B&W nodified version of PLECSET.
In order to inprove the calculated peak clad 0
terperature margin to the 10CFR 50.46 limit of 2200 F, at the six foot core elevation, the IDCA linear heat rate limit was reduced to 16.1 kW/ft at the beginning of cycle.
The end of cycle loCA linear heat rate limit was also I
reduced to 16.1 kW/ft.
This change was hnami on the peak clad tanperature l
behavior as a function of burnup for the ruptured and un-ruptured nodes as l
l shcun in BAW-1775.
7-2 B&W FuelCompany
b It is cx:ricluded fran the examination of cycle 10 core thennal and kinetics i
prtperties, with respect to acceptable previous cycle values, that this core reload will not achersely affect the MD-1 plant's ability to operate safely during cycle 10.
0:rsidering the previously acompta$ design basis used in the ISAR and subsequent cycles, the transient evaluation of cycle 10 is considera$ to be bounded 'by previously accepted analyses.
The initial conditions for the transients in cycle 10 are boundvl by the FSAR, the fuel densification report, ancyor subsequent cycle analyses.
,.a'
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7-3 L
B&W FuelCompany
kIi Table 7-1.
rWravisen of Cvele 9 and Cvele 10 Accident twome Cycle 9 doses, Cycle 10 doses, h
Rarn Eggl Hardlira Accident Syroid dose at EAB (2 h) 1.12 1.13 Whole body dose at EAB (2 h) 0.22 0.22 steam 11ne armak Ryroid dose at EAB (2 h) 1.82 1.80 Whole body dose at EA3 (2 h) 0.01 0.01 Steam Generator 'nibe Failq@
myroid dose at EAB (2 h) 6.53
~6.45 Nhole body dose at EAB (2 h) 0.56 0.55 Control Rod 1Mection Accident Syroid dose at EAB (2 h) 7.02 7.02 Whole body dose at EAB (2 h) 0.006 0.006-Syroid dose at LPZ (30 d) 5.64 5.64 Whole body dose at LPZ (30 d) 0.005 0.005 Loss of Coolira Accident (a)
@yroid doec.t "AB (2 h) 4.22 2.32 Whole body d,.a at EAB (2 h) 0.03 0.03
@yroid dose at'IPZ (30 d) 2.47 1.09 Whole body dose at LPZ (30 d)
-0.02 0.02 Maximwn Hvoothetical Accident (a)
Syroid dose at EAB (2 h) 165.1 91.5 Whole body dose at EAR (2 h) 5.03 4.85 Ryroid dos,e at LPZ (30 d) 87.8 39.4 Whole body dose at LPZ (30 d) 1.78 1.70 (a) me cycle 10 doses were calculated using reviced reactor building spray data.
7-4 B&W FuelCompany C
Table 7-2.
Cornparison of Fay Parameters 1
for Accident Analysis
- and Densification 90-1 Parameter Report Value AM Doppler coeff (BOC),10-4 Ak/k/T
-0.117
-0.159 Doppler coeff (EOC),10~4 ok/k/ F
-0.130
-0.180 O
Moderator coeff (SOC),10-4 ok/k/ F 0.0(a) (b)
-0.60 O
Moderator coeff (EDC),10~4 Ak/k/ F
-4.0(c)
-2.81 O
All-red group worth (HZP), % ak/k 12.90 B.38
=B Initial boron concentration (HFP), ppm 1150 1373 Boron reactivity worth (HFP),
100 129 pFIV% ak/k Max. ejected rod worth (HFP), % ok/k 0.65 0.20 Dsvpped red worth (HFP), % ak/k O.65
$0.20 (a)
+0.5 x 10-4 ak/k/ F was used for the moderator dilution analysis.
O (b) Transient results have been shown to be acceptable with a value of +0.9 x 10~4 AW F.
O (c)
_3,o x 10-4 3 p oF was used for the steam line failure analysis.
Table 7-3.
Bounding Values for Allevable IDCA Peak Linear Heat Rates Allowable Allowable Core peak IER, peak IHR, elevation, 0-1000 mwd /mtU, after 1000 mwd /mtU, ft kW/ft kW/ft 2
14.5 15.5 4
16.1 16.6 6
16.1 16.1 6
17.0 17.0 10 16 0 16.0 7-5 B&W FuelCompany
~
8.
PROPOSED }ODIFICATIONS TO TEONICAL SPECIFICATIOtB 2he Technical Specifications have been revised for cycle 10 operation for changes in core reactivity, pwer peaking and control rod worths.
We cycle 10 design umlysis basis includes the inpact of extended lw-power operation Ct 8V4 of rated power for cycle 9.
The cycle 10 basis also includes a very lw leakage fuel cycle design, a mixed Mark B4/ Mark B6/ Mark B8 fuel assembly core, gray APSRs, and crossflw analysis.
Se LOCA linear heat rate-limits used to develop the Technical Specification Limiting Conditions for Operation include the combined effects of the NUREG-0630 cladding swell and rupture i
nodel, use of the BWC QiF correlation, reduce.d fuel rod pre-pressure, and inplementation of the B&W moiitled version of FLECSET.14,15,16,17 A
cycle 10 specific analysis was conducted to generate Technical Specification Limiting Conditions for Operation (rod index, axial power imbalance, and quadrant tilt), Mwi on the methodology described in reference 18.
The effects of gray APSR repositioning were included in the analysis.
The burnup-dependent allowable LOCA linear heat rate limits used in the analysis are provided in Figure 8-15.
The analysis also determined that the cycle 10 'nx:hnical Specifications provide protection for the overpower coniition that could mmm during an overcooling transient because of nuclear instrumentation errors, and verified removal of the power level' l.
cutoff hold requirement.
'DK:hnical Specification section 3.5.2.4 was revised to am - -Yate a change in the quadrant tilt setpoint, hawl on incere detector sensitivity depletion.19 The full incere quadrant tilt setpoint reported in section 3.5.2.4 is the bounding value, derived from the detector sensitivity j'
. depletion at end-of-cycle 10, of the full incere quadrant tilt setpoint P
values detemined for cycle 10.
'n= Leasur ament system-independent rod position an:1 axial pwer imbalance limits deterrined by the cycle 10 analysis were enor-adjusted to generate alam setpoints for pwer operation and are reflected in a Technical Specification reviLion to sections 3.5.2.5 ar'd 3.5.2.6.
The alarm setpoints are provided in Figures 8-3 through 8-14.
8-1 B&W FuelCompany
7 Based on the analyses and Technical Specification revisicm described in this report, the Final Acceptance Criteria ECG limits will not be aM, ter will' the thermal de'.41gn ' criteria be violated.
The followirg pages contain
-the revisions to the Technical Specifications.
- \\'
8-2 4
B&W FuelCompany
i g
13.1.7I Mmtor Tamrmnture Coefficient of Raactivity Specification 3.1.7.1
. Ihe moderator tatperature coefficient (MIC) shall. be irsr-positive whenever themal power is 295% of nted thermal power and shall be Ak/k/ F whenever thermal power is l
less positive than 0.9 x 10'4 0
<95% of rated thermal power and the reactor is not shutdown.
3.1.7.2-The MIC shall be determined to be within its limit by confirmatery measw.
hnts prior to initial operation above 5% of rated thermal power after each fuel loadirg.
MIC measured values shall be extrapolated and/or % sated to permit direct cxatparison with the limits in 3.1.7.1 above.
3.1.7.3 With the MIC outside any one of the above ' limits, be in at least Hot Standby within 6 hoars.
Enses A non-positive moderator mefficient at power levels above 95% of rated power is specified such that the -4==
clad tenperatures will-not (W the
-Final Acceptance criteria based on IDCA analyses.
Below 95% of rated power,-
-the Final Acceptance Criteria will not be ex
- with a positive moderator Ak/k/ F-w rrected to 95% of rated l
temperature coefficient of
+0.9 x 10~4 O
power.
All other accident a:.alyses as reported in the IEAR have been performed'for a rarge of moderator temperature coefficients including +0.9 x 10~4 Ak/k/ F.
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8-3
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B&W FuelCompany
,+
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Figure 8-1.
Core Protection Safety Limits -- ANO-1 (Tech Spec Figure 2.1-2)
Thermal Power Level
% FP
. 120
_ (-33.5.112.0)
_ (33.5.112)
ACCEPTABLE I 4 PUMP l
i(42.6,103.6)
-100
(-49.7,96.9) l OPERATION g
1(-33.5.90.75)
(33.5,90.75) l l
I 8
l ACCEPTABLE 3&4 PUMP g
- 80
.(42.6,82. 35 ) -
(-49.7,75.65)y I OPERATION I
l-1 I
g l
l (-33.5,64.08)
(33.5,64.08)I l
l'
. 60 l
l ACCEPTABLE 2,3&4 PUMP
>(42.6,55.68)
(-49.7,48.98)
OPERATION g
I l-t.0 -
3 I
I l
l-I i
I g
i i
l l
- 20 l
l l
l l
1 I
i l
i I
I I
I If I
i
-60
--40
-20 0
20 40 60 80 Reactor Power Imbalance, %
8-4 B&W FuelCompag
l:
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. Figure 8-2. -Protective System Maximum Allowable Setpoints -- ANO-1
'(Tech Spec Figure 2.3-2)
. Thermal Power Level,
% FP 120
(-18.0.107 ),
_ (18.0,107) 100 ACCEPTABLE-1 4 PUMP l
(31.8,93.2)
(-38.5,86.5)
OPERATION l
(18.0,79.9)
(-18.0,79.9) l~
I l
l l
l ACCEPTABLE I
(31.8,66.1) 384 PUMP
(-38.5,59.4) g OPE TION 60 I
l
(-18.0,52.6) I
^
l (1 8.0,52.6) l' l
40 ACCEPTABLE -
> (31.8,38.8)
.(-38.5,32.1)l 2,384 PUMP l
OPERATION g
l l
1 g
l l
20-g 1
I i
3 I
I l
L 1
l l
I i
il ll' 11 l
'l 1
i
,.-40
-20 0
20 40-60 80
?
Reactor Power Imbalance, %
l~
t h
p 8-5 B&W FuelCompany
as.
-) k N u[ g
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r 6.
If a u.mLv11 red in the' regulating or axial power ' shaping -
groups is declared inoperable per Specification 4.7.1.2 operatica abcVe 60 percent of the thermal power allcwable for the reactor coolant punp ombination may continue provided.the :
rods in the group are positioned such that the rod that was declared incperable is contained within allowable group average position limits of Specification 4.7.1.2 - and the withdrawal limits of Specification 3.5.2.5.3.
3.5.2.3
'Ibe worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.
3.5.2.4 Quadrant Tilt:
1.
Except for physics tests, if quadrant tilt exceeds 4.24%,
l redace power so as not.to exceed the allowable power level for the existing reactor coolant punp ombination-less at least 2%
for each 1% tilt in excess of 4.24%.
l.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall.be zwhvwi to less than 4.24% except for physics tests, or the l
following adjustments in setpoints ard limits shall be made:
'Ihe protection system maximum allowable setpoints (Figure a.
2.3-2) shall be reduced 2% in power for each 1% tilt, b.
'Ibe control rod group and APSR withdrawal limits shall'be e v'ad 2% in power for each 1% tilt in excess of 4.24%.
l r
c.
'Ibe _ operational imbalance lindts shall be reduced 2% in power for each 1% tilt in excess of 4.24%.
l 3.
If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quadrant power tilt is permitted provided the thermal power allowable for the reactor coolant pump ccubination is restricted as stated in 3.5.2.4.1 above.
4.
Quadrant tilt shall be monitored on a minimum frequency.of once every two hours during power operation above 15% of rated power.
8-6 B&W FuelCompany
3.
Except for physics tests or exercising c.ui.wl rods, the control red withdrawal limits are specified on Figures 3.5.2-1(A-C), 3.5.2-2(A-C), and 3.5.2-3 (A-C) for 4, 3 and 2 punp operation respectively.
If the applicable omtrol rod position limits are e W -, corrective measures shall be-taken imediately to achieve an acceptable control; rod position.
Acceptable control rod positions shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Except for physics tests or exercisirh axial power shaping 4.
rods (APSRs), the following limits apply to APSR position:
Up to 345 EFPD, the APSRs may be positioned as rmmaaney for l
transient imbalance control, however, the APSRs shall be fully withdrawn by 345 EFPD. After 345 EFPD, the APSRs shall not'be l
reinserted.
With the APSRs inserted after 345 EFPD, corretive measures
'l shall be taken imediately to achieve the full withdrawn position.
Acceptable APSR positions shall be attained-within 4 hcurs.
3.5.2.6 Reactor power Inbalance shall be monitored on - frequency not to exceed' 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power eperation above 40% rated power.
Except for physics tests, imbalance shall be maintained within.the envelope defined by Figure 3.5.2-4(A-C).
If the imbalance is not within the envelope defined by Figure
- 3. 5. 2-4 (A-C), corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be r*Wai until imbalance limits are met.
3.5.2.7
'Ibe control rod drive patch panels shall be locked at all times with limited a m a to be authorized by the Superintendent.
Bases
'Ihe power-imbalance envelope defined in Figure 3.5.2-4(A-C) is b-i on IDCA analyses which have ' defined the maximum linear heat rate.
(see -Figure 3.5.2-5),
such that the maxinum cladding temperature will not exceed the Final Acceptance Criteria.
Corrective measures will be taken imediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundaries.
Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly inprobable because all of the power distribution' parameters- (quadrant tilt, rod position, and imbalance) must be at their limits while 8-7 B&W FuelCompany
l Figure 8-3.
Rod Position Setpoints for Four-Pump Operation from 0 to 30
+10/-0 EFPD -- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-1 A) 110 (200,9,102)
(300,102)
(272.9,102)?
100 SHUTDOWN (270.5,90) 90 MARGIN LIMIT 80 OPERATION IN THIS REGION IS NOT a
(255.5.78)
E ALLOWED OPERATION 70 m
RESTRICTED m
60 g
g 50 (109.5,48)
(227.5,48) k o.
40 p
30 PERMISSIBLE OPERATING 20 10 (73.5.13)
(0,0) 0 I
I I
I I
I I
I I
I I
I I
i 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80 100 1
1 I
t i
i Group 7 0
20 40 60 80 100 i
i l
I i
I Group 6 0
20 40 60 80 100 l
i I
i l
l Group 5 Rod Index, % WD 1
8-8 B&W FuelCompany l
i Figure 8-4 Rod Position Setpoints for Four-Pump Operation From 30 +10/-0 to 335 2.10 EFPD -- ANO-1 Cycle _10 (Tech Spec Figure 3.5.2-1B)_
4 (300,102) 1 (200.9,102)-
(270.1.102) ;
, gg SHUTDOWN (266.5,90) 90 MARGIN LIMIT
.g 80 (251.5,78) co OPERATION IN THIS
' $ 70 REGION IS NOT-
"~
OPERATION ALLOWED RESTRICTED o
60
- g 2
50
-(109.5,48)
(227.5,48)'
40 30 PERMISSIBLE OPERATING:
REGIM 20 10 (73.5,13)
(0,0) O I
I i
I I
i l
I I
I l'
l I
I 0
20 40 60 80 100 120 140 160 180 200 220 240 260'.280.300 0
20 40 60 80 100 l
i i
1 Group 7 J
0 20 40 60 80 100 l
I l
i 1
i Group 6 0
20 40 60 80 100 l
l 1
i i
Rod Index, % WD Group 5 8-9 B&W FuelCompany
Figure 8-5.
Rod Position Setpoints for Four-Pump Operation After 335' 210 EFPD -- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-1C)
~
110=
(215.0,102)
(300,102)
(270.1,102);
100 SHUTDOWN 90-MARGIN (266.5.90)
OPERAT10N IN THIS LIMIT REGION IS NOT N
^
~
(251.5,78) e
~
OPERATION 60 64 s.'
g - 50 (121.5,48)
(227.5,48) o.
- 40 30 PERMISSIBLE 20 OPERATING REGION 1o (79.5.13)
I I
I l'
I I
I I
I I
I I
I (0,0)' O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 O
0 20 40 60 80 100 l
I I
I I-1 Group 7 0
20 40 60 80 100 t
l l
l 1
i Group 6
-0 20 40 60 80 100 I
I I
I I
I Rod Index, % WD Group 5 8-10 B&W FuelCompany
l A
figure 8-6.
Rod Position Setpoints for Three-Pump Operation From 0 to
'30 +10/-0 EFPD -- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-2A)'
110 100 90 (300,77) 80 (202.0,77)
(273.0,77)
E 70-SHUTDOWN e
MARGIN (270.5,67) 8 OPERATION IN THIS LIMIT
-[ 60 REGION IS NOT (255.5,58) o
' ALLOWED
" 50 J
OPERATION
-[ 40 RESTRICTED (109.5,36)
(227.5,35.5) 30 PERMISSIBLE
- 20 OPERATING 10 REGION
('0,0) O' I
I I
I I
I I
I I
I 1
0 20 40 60 80 100 120 140 160 180 200 220 240 260- 280 300 0
20 40 60 80 100 1
1 1
1 I
I Group 7 0
20 40 60 80 100 1
I I
I I
I Group 6 0
20 40 60 80 100 t
i i
I I
i Rod Index, % WD Group 5 8-B&W FuelCompany
I 1
?
i Figure 8-7.
Rod Position Setpoints for Three-Pump Operation From 30
+10/-0 to 335 10 EFPD -- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-2B) 110 100 l
i 90 (202.0,77)
(300,77):
.[
go (270.3,77)
,{
SHUTDOWN i
70 e
MARGIN-(266.5,67)
OPERATION IN THIS LIMIT
[
60 REGION IS NOT (251.5,58) o ALLOWED e
~
i OPERATION.-
l 40 RESTRICTED a -
(109.5,36)
(227.5,35.5)
I 30 l
1:
~
PERMISSIBLE b
OPERATING i
L 10 (73.5,9.8)
REGION L
- (0,0)'O -
1 I
l I
I I
I I
I I
I I
I I'
'6~
20 40 60 80 100 120 140 160. 180 200 220 '240 260 280 300 e
-0 20 40 60
,80't : 100 I
i l
1 i
I Group 7 0
20 40 60 80 100 l
l l
l I
I
- j L-
-Group 6 0'
20
'40 60 80 100 i
i i
i i
i i
Rod Index, % WD, Group 5 l
8-12 L
B&W FuelCompany i
j L
' E 1
' Figure'8-8._ Rod Position Setpoints for Three-Pump Operation After.
335 10 EFPD - ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-2C)
- 110 100 90 (300,77) 80 (270.3,77)
.[
(216.1,77) l F
70 OPERATION IN THIS SHUTDOWN (266.5,67)
N REGION IS NOT MARGIN g 60 =--
ALLOWED LIMIT (251.5,58) i
)
\\
c 50 OPERATION g
RESTRICTED n.140L (121.5,36)
(227.5,35.5)-
30 20 PERMISSIBLE l
10 OPERATING.
(79.5,9.8)
REGION
_(0,g) i g
i i
0:
20 - 40 60 80 100 120 140 160 180 200 220 240 260 280 300 l
0 20 40 60 80 100 t
i i
i i
i I
Group.7 0
20 40 60 80 100 i
i I
i i
I Group 6 0-20 40 50 80 100 t
i I
i 1
1 Rod Index, % WD Group 5 i
B-13 B&W FuelCompany
Figure 8-9.
Rod Position Setpoints for Two-Pump Operation From 0 to
.30 +10/-0 EFPD-- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-3A) 110 100 90 g 80 70 t
60 (204.3,52)
(300,52)
(273.2,52) 5-l 50 OPERATION IN THIS SHUTDOWN a-REGION IS NOT MARGIN (270.5,44) i ALLOWED-LIMIT-40 (255.5,38)
OPERATION-30 RESTRICTED (109.5,24)
(227.5.23)'
PERMISSIBLE.
10 OPERATING
~
.(73.5,6.5)
REGION-
'd
.20 40 60-
- 80. 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80' 100 i
l-i i
i 1
Group 7 0
20' 40 60 80 100 I
I I
I I
I Group 6 0
20 40 60 80 100 I
I I
I I
I Rod Index, % WD Group 5 8-14 B&W FuelCompany
Figure 8-10.
Rod Position Setpoints for Two-Pump Operation From 30 +10/-0 to 335 210 EFPD - ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-3B).
110' 100 90 j 80 e
[ 70 t
60 w
(300,52) c (204.3.52)
(270.5.52) g~50 OPERATION >lN THIS a.
. REGION IS NOT SHUTDOWN (266.5,44)
ALLOWED MARGIN 40 LIMIT (251.5,38) 30 OPERATION (109.5.24)_
RESTRICTED.
(227.5,23)-
PERMISSIBLE 0 R 10 (73.5,6.5 (d,0)0 I
I I
I I
I I
I I
I I
I
')
20 40 60-80 100 120 140 160 180 200 220 240 260. 280 300 0
20 40 60 80 100 i
l i
i i
i Group 7 0
20 40 60 80 100 t
I i
1 I
I Group 6 0
20 40 60 80 100 t
i I
I I
I r.od Index, % WO Group 5 8-15 B&W FuelCompany
Figure 8-11 Rod Position Setpoints for Two-Pump 0 m ation After 335 !10 EFPD -- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-3C) 110 100 90
[80 m
%70 o" 60 (300,52) 50 OPERATION IN THIS SHUTDOWN (218.4,52)
(270.5,52)-
REGION IS NOT MARGIN (266*5'44)
ALLOWED LIMIT 40 (251.5,38) 30 (121.5.24)
OPERATION-RESTRICTED (227.5.23) 20 PERMISSIBLE (79.5,6.5 OPERATING 10 REGION
-(0,0) 0, I
I I
I I
I I
I I
I I
1 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80' 100 t
i I
i l
l-Group 7 0
20 40 60 80 100 L_.
l l
l t
i Group 6 0
20 40 60 80 100 I
I I
I-1 I
Rod index, % WD Group 5 8-16 B&W FuelCompany
f Figure-8-12.
Operational Power imbalance Setpoints for Operation From 0 to 30 +10/-0 EFPD -- ANO-1 Cycle 10 (Tech Spec figure 3.5.2-4A)
- 110
(-15.05,102)
- (13.93,102)
(-16.15,92)'
(16.97,92)
- 90
(-22.10,80) 80
>(20.05,80) w w
g
'O g RESTRICTED RESTRICTED
@g
@g REGION g
g REGION j"Eb
-605E S
wtS w%=
n.
- a. o =
(-22.50,50)< >
- 50
<>(20.50,50)
N
.. - 40 30 y1
,- - 20 a..
10 I
I i
i i
t I
1 1
1 40 -30
-20
-10 0
10 20-30 40 50 Axial Power Imbalance, %
1 l
l 8-17 l
B&W FuelCompany i
Figure 8-13.
Operational Power Imbalance Setpoints for' Operation From ;
30 +10/-0 to 335 210 EFPD'- ' ANO-1 CYCLE 10 (Tech Spec Figure'3.5.2-48) i
- - 110
(-19.61,102)
(17.53,102)
^
(-20.72,92)-
- - 90 (20.59,92)
(-23.80,80)I 80 n(23.68,80) w m
a a
52' SE RESTRICTED 70 $ g =
RESTRICTED
$p!
REGION 60' g g S gg REGION gy$
R$$
(-24.26,50) o 50 o(24.18,50) f
. 40 h
30
- - 20 10
- a...
t i
l i
i l
l l
1 I
i
-50 30 -20
-10 0
10 20 30 40 50-
~ Axial Power Imbalance, %
7 lr l
L l
\\
^
8-18 B&W FuelCompany l
1 Figure 8-14 Operational Power. Imbalance Setpoints for Operation After 335 10 EFPD ---ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-4C) i
?
[>
l.
- - 110 1
(-20.47,102)
- (17.58,102)
(-21.59,92) 90
' (21. 59,92 )-
(-23.71,80)<
80
,(23.71,80) i i
4
[
RESTRICTED
$e RESTRICTED 70 e.
REGION g5 g5 REGION
~~
$25
$EE E5G 60 E 5 G E8E E%E
(-24.19,50) o 50
<>-(24.19,50) iE 40 e
g N
30 o
20-I 10 s--
l i
1 1
1 1
I I
I I
vi 40 -30
-20
-10 0'
10 20
.30.
40 50 Axial Power imbalance, %
1-a-19 B&W FuelCompany o-4
Figure 8-15.
LOCA Limited Maximum Allowable Linear Heat Rate -- ANO-1 Cycle 10 (Tech Spec Figure 3.5.2-5) 20 d 18 E
a
.m.
3 m
" 16
/ p- - - -
S
/
^
/
~
2
/
/
8 14
~
B G
'3 12 O to 1000 mwd /mtU x
After 1000 mwd /mtU 10 i
i i
8 O
2 4
6 8
10 12 Axial Location From Bottom of Core, ft 1
l i
8-20 B&W FuelCompany
9.
SIARIUP PBCGRAM - IHYSICS TEtDiG me planned startup test program a=wiated with core perfor.me is outlined-
-below. %ese tests verify that oors performance is within tha===wtions of
> the safety e.aalysis and provide information for continued rafe operation of the unit.
9.1.
Precritical 'hasts 9.1.1.
Control Rod Trio Test Precritical control rod drop times are remrded for all.wuLwl rods at hot full-flot, conditions.before zero power physics testing begins.
Acceptance
-criteria stata that the rod drop time frun fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions ab7ve.
It should be noted that safety analysis calculations are based on a rod drop frm fully withdrawn to two-thirds insertad.
Since the' most accurate position indication is obtained frcan the ::one reference switch at the 75%-inserted position, this position is used instead of the two-thirds inserted position for data gathering.
i 9.1.2.
RC Flow Reactor coolant flow with four RC pumps running will be measured at hot q
shutdown conditions.
Acceptance critaria require that the measured flow be j
within allowable limits.
9.2.
Zero Power Physics Tests 9.2.1.
Critical Boron Concentration i
once initial criticality is achieved, equilibrium boron is obtained and the critical boron concentration determined. We critical boron concentration is calculated by correcting for any rod withdrawal required to achieve equilibrium boron.
We acceptance criterion placed on critical boron concentration is that the actual boron concentration must be within 2100 ppu baron of the predicted value.
8 9-1 l
B&W FuelCompany
h 9.2.2.
Tamrantture Reactivity Coefficient he isothermal HZP tenperature coefficient is measured at approximately the" all-rods-cut configuration.
During M in tenperature, reactivity ersated by control rod movement.
We change in faelhack may be us reactivity is then calculated by the summation of reactivity (obtained from a reactivity calculator strip chart recorder) acwv-iated with the tenperature change.
Acceptance criteria state that the measured value shall not differ 4.' rom the predicted value by more than 0.4 x 10-4 ok/k/ F.
0 The moderator coefficient of reactivity is calculated in conjunction with the tempet are coefficient measuremerru.
After the tenperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain the moderator coefficient. Wis value must not be in excess of the acceptance criteria limit of +0.9 x 10~4 AW F.
O 9.2.3.
Control Rod Grouo/ Boron Reactivity Worth Control rod group reactivity worths (groups 5, 6, an:17) are measured at hot zero power conditions using the boron / rod swap method.
Bis technique consists of establishing a deboration rate in the reactor coolant system and cui,-sating for the reactivity charges from this deboration by inserting control rod groups 7, 6, and 5 in irmatr_ntal steps. W e reactivity changes that e m m during these measurements are ;alculated h==ad on reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the charge in rod group position.
We differential rod worths of each of the controlling groups are then sn-d to obtain integral rod group worths.
We acceptance criteria for the control bank group worths are as follows:
1.
Individual bank 5, 6, 7 worth:
credicted value - measured value x 100 5 15 2.
Sums of groups 5, 6, ard 7:
credicted value - measured valug x 100 < 10 measured value he boron reactivity worth (differential boron worth) is measured by dividing the total inserted rod worth by the boron change rade for the red worth test.
We acceptance criterion for measured differential baron worth is as follows:
9-2 B&W FuelCompany
1.
oradicted value - measurnd value x 100 $ 15 naasured value he predicted rod worths and differential boron worth are taken fzun the PIN.
9.3.
Power Escalation 'IWsts 9.3.1, Core Svmetry 'Ibst
'lhe purpose of this test is to evaluate the symetry of the core at low power during the initial power escalation following a refueling.
Symetry evaluation is based on incore quadrant power tilts during escalation to the inta*4 ate power level.
me core symetry is acceptable if the absolute values of the quadrant power tilts are less than the error adjusted alarm limit.
9.3.2.
Core Power Distribution Verification at Inta==ilate Power Imvel (IPL) and 100% FP With Nominal Control Rod Ibsition Core power ' distribution tests are performed at the IPL and 100% full power (FP).
Equilibrium xenon is established prior to both the IPL ard 100% FP tests.
'Ihe test at the IPL is essentially a check on power distribution in the core to -identify any abnormalities before escalating to the 100% FP plateau.
Peaking factor criteria are applied to the IPL core power distribution results to determine if additional tests or analyses 'are required prior to 100% FP cperation.
% e following acceptance criteria are placed on the IPL and 100% FP tests:
1.
De maxinun IER nust be less than the IDCA limit.
2.
'Ihe minimum INBR must be greater than the initial condition INER limit.
3.
'Ibe value obtained from extrapolation of the mininum DER to the next power plateau overpower trip setpoint must be greater than the initial condition DER limit, or the extrapolated value of imbalance must fall outside the RPS powir/ imbalance / flow trip envelope.
4.
'Ihe value obtained from extrapolation of the worst-case raximum IER to the next power plateau overpower trip setpoint must be less than the fuel melt limit, or the extrapolated value of imbalance n.ast fall outside the RPS power / imbalance / flow trip envelope.
9-3 B&W FuelCompany
5.
Se quadrant pwer tilt shall not EW the limits specified in the Technical Fpecifications.
6.
The highest measured ard predicted radial peaks shall be within the follwing limits:
credicted value - reasured value x 100 nom positive than -5 measured value 7.
%e highest measured and predicted total peaks shall be within the follwing limits:
omdih he - meh nlue x 100 more positive than -7.5 measured value Items 1, 2, and 5 ensure that the safety limits are maintained at the IPL and 100 %FP.
Items 3 and 4 establish the criteria whereby escalation to full power may be accceplished without tan potential for exceeding the safety limits at the overpower trip setpoint with regard to CNBR and linear heat rate.
Items 6 ard 7 are established to determine if measured and predicted power distributions are consistent.
9.3.3.
Incore Vs. Excore Detector Imbalance Cormlation Verification at the IPL Imbalances, wt up in the core by control rod positioning, are read sinultaneously on the incore detectors and excore power rarge detectors. he excore detector offset versus incore detector offset slope must he greater than 0.96.
If this criterion is not met, gain anplifiers on che excore detector signal pmaing equignent are adjusted to prcvide the required gain.
9.3.4.
Ter1oerature Reactivity Coefficient at ~100% FP he average reactor coolant tenparatre is decreamed and then increased by about 5 F at co.ht reactor power.
We reactivity associated with each 0
temperature charge is obtained from the change in the controllirq rod group position.
Controlling rod grwp worth is measured by the fast insert / withdraw method. We tenperature reactivity coefficient is calculated frun the measured charges in reactivity and temperature.
Acceptance criteria state that the moderator temperature coefficient shall be negative.
94 B&W FuelCompany
1 1
9 '. 3. 5.
Power Doooler Reactivity Coefficient at ~ 100% FP
'Ihe power Doppler reactivity coefficient is calculated fmn-data recorded during control rod worth measurements at power.using the fast insert / withdraw l
method.
'Ibe fuel Doppler reactivity coefficient is calculated in conjunction with the jl power Doppler coefficient measurement.
'Ibe power Ecppler coefficient as measured above is us.titiplied by a precalculated conver-4cn factor to obtain
- the. fuel Doppler coefficient. 'Ihis measured fuel Doppler coefficient nust be 4
more negative than the. acceptance critaria limit of -0.90 x 10-5 3g, 9.4.
Procedure for Use if Acceptance Criteria Not Met If acceptance critaria for any test are not met, an evaluation is performed I
before the test program is continued.
Further specific' actions depend on evaluation results.
'Ibese 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 a ealated until evaluation 4 shows that plant safety 'will not be ocmpmnised by such escalation.
9-5 B&W FuelCompany
{'
_(6
[:
I 10.
REFERENCES i
l '.
Arkansas Nuclear one, Unit 1 - Final Safety Analysis Report, Docket 50-313, Arkansas Power & Light.
2.
Pr@ san to Determine In-Restor Performance of B&W Fuels - Claddirg Creep Collapse, BAW-13084P-A, Rev.
2, hhrw X &. Wilcox, Lynchburg, Virginia, October 1978.
3.
Letter, J.H. Taylor (B&W) to C.O. 'Ihcnns (NRC),
Subject:
Creep Collapse Analysis for B&W Fuel, Jifr/86-011A, Dated January 31, 1986.
4.
Intter, Dennis M.
Crutchfield (NRC) to J.H. Taylor (B&W),
Subject:
Acceqt.ance for Referencing of a Special Licensirg Report, Dated Mr 5, 1986.
5.
Tarn 2-Fuel Pin Performance Analysis, BAW-10141P-A. Rev.1, Mhrw'k &
Wilcox, Lynchburg, Virginia, June 1983.
6.
NOODIE - A Multi-Dimensional 'IVo-Group Reactor Simulator, BAW-10152A, hhewk & Wilcox, Lynchburg, Virginia, June 1985.
7.
Arkansas Nuclear One Unit 1, Cycle 9-Reload Report, BAW-2027, hWk &
Wilcox, Lynchburg, Virginia, June 1988.
8.
LYNXT - Core Transient 'Ibermal-Hydraulic Ptwimn, BAW-10156A, hhk &
Wilcox, Lyrchburg, Virginia, Feb'.g 1986.
9.
'Ihermal-Hydraulic Crossflow Aplications, BAW-1829, h W J: & Wilcox, Lynchburg, Virginia, April 1984.
10.
Rancho Seco Cycle 7_ Reload Report - Volume 1 - Mark BZ Pbel Assembly Design Report, BAW-1781P, hWk & Wilcox, Lynchburg, Virginia, April 1983.
11.
BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & Wilcox, Lynchburg, Virginia, April 1985.
l l-l 12.
Fuel Rod Bowirg in kWk & Wi]cox Fuel Designs, BAW-10147P-A, Rev.1, NWk & Wilcox, Lynchburg, Virginia, May 1983.
10-1 B&W FuelCompany L
l
- 13. Arkansas Nuclear One, Unit 1 - Ibel Densification Report, BAW-1391, Babcock & Wilcox, Lyrchburg, Virginia, June 1973.
14.
DCCS Analysis of B&W's 177-FA lowered-Irop NSS, BAW-10103A, Pev.
3,
=
ILW: & Wilcox, Lynchburg, Virginia, July 197..
15.
TACO Loss-of-Cbolant Ao::ident Lif tit Aralysis for 177-FA loav. red Locp Plants, BAW-1775, hWk & Wilcox, Lyrchburg, Virginia, February 1983.
9 16.
Bounding Analytical A== mant of IURD3-0630 Models on LOCA kW/ft Limits With Use of FLECSLT, BAW-1915PA, hWA & Wilococ, Lynchburg, Virginia, November 1988.
17.
B&W ECCS Evaluation Mode.1 Revision 5, BAW-10104PA, Rev.
5, hWk&
Wilcox, Lynchburg, Virginia, November 1938.
p 18.
Normal Operating Controls, BAW-10122A, Rev.
1, Babcock & Wil m,
Lyrchburg, Virginia, May 1984.
19.
J. H. Taylor to J. A. NorbeIg, Letter, bterdad Lifetime Incore Detector hror Allowances, J1fr/88-28, April 21,1988.
m 10-2 B&W FuelCompany ii..
,