ML20069M126
| ML20069M126 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/31/1982 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20069M125 | List: |
| References | |
| BAW-1747, NUDOCS 8211230148 | |
| Download: ML20069M126 (65) | |
Text
._
BAW-1747 g
July 1982 I
i ARKANSAS NUCLEAR ONE, UNIT 1
- Cycle 6, Reload Report -
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July 1982 l
I A7dOLNSAS NUCLEAR ONE, UNIT 1
- Cycle 6 Reload Report -
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l BABCOCK & WILCOX Nuclear Power Group Utility Power Generation Division P. O. Box 1260 i
Lynchburg, Virginia 24505 i
Babcock & Wilcox
l CCNTENTS Page i
1-1 1.
INTRODUCTION AND
SUMMARY
2-1 2.
OPERATING HISTORY 3-1 3.
GENERAL DESCRIPTION 4-1 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.
Thermal Design.......................
4-3 4.4.
Material Design 4-3 4.5.
Operating Experience.
5-1 5.
NUCLEAR DESIGN..........................
5-1 5.1.
Physics Characteristics 5-1 5.2.
Analytical Input......................
5-2 5.3.
Changes in Nuclear Design 6-1 6.
THERMAL-HYDRAULIC DESIGN.....................
7-1 7.
ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.
General Safety Analysis 7-1 7.2.
Accident Evaluation 8-1 8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS........
9-1 9.
STARTUP PROGRAM - PHYSICS TESTING 9-1 9.1.
Precritical Tests 9-1 9.1.1.
Control Rod Trip Test 9-2 9.2.
Zero Power Physics Tests..................
9-2 9.2.1.
Critical Boron Concentration..
9-2 9.2.2.
Temperature Reactivity Coefficient.........
9.2.3.
Control Rod Group Reactivity Worth.........
9-2 9-3 9.2.4.
Ejected Control Rod Reactivity Worth.
....... Babcock &Wilcox
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COFTENTS (Cont'd)
Page 9-3 9.3.
Power Escalation Tests.
9.3.1.
Core Power Distribution Verification at %40, 75, and 100% FP With Nominal Control Rod 5
Position......................
9-3 9.3.2.
Incore Vs Excore Detector Imbalance
[
Correlation Verification at s40% FP 9-5 9.3.3.
Temperature Reactivity Coefficient at 4
s100% FP.
9-5 9.3.4 Power Doppler Reactivity Coefficient at s100% FP.
9-5 9.4.
Procedure for Use if Acceptance Criteria Not Met 9-5 REFERENCES.
A-1 List of Tables Table 4-1.
Fuel Design Parameters and Dimensions 4-4 4-2.
Fuel Thermal Analysis Parameters..
4-5 5-1.
Physics Parameters for ANO-1, Cycles 5 and 6..........
5-2 5-2.
Shutdown Margin Calculations for ANO-1, Cycle 6 5-4 6-1.
Maximum Design Conditions, Cycles 5 and 6 6-2 6-2.
DNBR Rod Bow Penalty, ANO-1 Cycle 6 6-3 7-1.
Comparison of FSAR and Cycle 6 Doses.
7-4 7-2.
Comparison of Key Parameters for Accident Analysis..
7-5 f-7-3.
Bounding Values for Allowable LOCA Peak Linear Heat Rates 7-5 F
t t
List of Figures 4
Figure 3-1.
Core Loading Diagram for ANO-1 Cycle 6 3-3 l[e 3-2.
Enrichment and Burnup Distribution, AND-1 Cycle 6 Off l
455 EFPD Cycle 5 3-4 3-3.
Control Locations and Croup Designations for ANO-1 Cycle 6 3-5 3-4.
LBP Enrichment and Distribution, ANO-1 Cycle 6 3-6 5-1.
ANO-1 Cycle 6, BOC (4 EFPD) Two-Dimensional Relative Power l
Distribution -- Full Power Equilibrium Xenon, Normal Rod Positions...
5-5 8-1.
Core Protection Safety Limits.
8-2 8-2.
Protective System Maximum Allowable Setpoints.
8-3 8-3.
Boric Acid Addition Tank Volume and Concentration Requirements l
Vs RCS Average Temperaure.
8-4
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Babcock 4.Wilcox
Figures (Cont'd)
Figure Page 8-4.
Rod Position Limits for Four-Pump Operation From 0 to 60 EFPD - ANO-1, Cycle 6......................
8-5 8-5.
Rod Position Limits for Four-Pump Operation From 50 to 200 1 10 EFPD -- ANO-1, Cycle 6 8-6 8-6.
Rod Position Limits for Four-Pump Operation From
...........- 8-7 200 t 10 to 350 1 10 EFPD -- AND-1, Cycle 6 8-7.
Rod Position Limits for Four-Pump Operation Af ter 350 10 EFPD -- ANO-1, Cy cl e 6......................
8-8 8-8.
Rod Position Limits for Three-Pump Operation From 0 to 60 EFPD -- ANO-1, Cycl e 6......................
8-9 8-9.
Rod Position Limits for Three-Pump Operation From 50 to 200 1 10 EFPD -- ANO-1, Cycle 6 8-10 8-10.
Rod Position Limits for Three-Punp Operation From 200 10 to 350 2 10 EFPD -- ANO-1, Cycle 6................
8-11 8-11.
Rod Position Limits for Three-Pump Operation Af ter 350 10 EFPD -- ANO-1, Cy cl e 6......................
8-12 8-12.
Rod Position Limits for Two-Pump Operation From 0 to 60 EFPD -- AN O-1, Cycle 6...................... 8-13 8-13.
Rod Position Limits for Two-Pump Operation From 50 to 200 1 10 EFPD -- ANO-1. Cycle 6 8-14 8-14.
Rod Position Limits for Two-Pump Operation From 200 10 to 350 1 10 EFPD -- ANO-1, Cycle 6................
8-15 8-15.
Rod Position Limits for Two-Pump Operation After 350 ! 10 EFPD -- AN O-1, Cycle 6...................... 8-16 8-16.
Operational Power Imbalance Envelope for Operation From 0 to 60 EFPD -- ANO-1, Cycle 6................... 8-17 8-17.
Operational Power Imbalance Envelope for Operation From 50 to 200 1 10 EFPD -- ANO-1, Cycle 6................ 8-18 8-18.
Operational Power Imbalance Envelope for Operation From 200 10 to 350 1 10 EFPD -- AND-1, Cycle 6 8-19 8-19.
Operational Power Imbalance Envelope for Operation After 350 10 EFPD -- ANO-1, Cycle 6 '.................
8-20 8-20.
APSR Position Limits for Operation From 0 to 60 EFPD -
8-21 ANO-1, Cycle 6 8-21.
APSR Position Limits for Operation From 50 to 200 10 EFPD -- ANO-1, Cycle 6......................
8-22 8-22.
APSR Position Limits for Operation From 200 10 to 350 1 10 EFFD - ANO-1, Cycle 6 8-23 8-23.
APSR Position Limits for Operation Af ter 350 10 EFPD --
8-24 ANO-1, Cycle 6 Babcock & Wi!cox
- iv -
1.
INTRODUCTION AND
SUMMARY
This report justifies the operation of the sixth cycle of Arkansas Nuclear One, Unit 1 (ANO-1) at the rated core power of 2568 MWt.
Included are the required analyses as outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," June 1975.
To support cycle 6. operation of ANO-1, this report employs analytical tech-niques and design bases established in reports that have been submitted to and accepted by the USNRC and its predecessor, the USAEC (see references).
The cycle Sjand 6 reactor parameters relsted to power capability are summarized briefly in section 5 of this report. All of the accidents analyzed in the FSAR have been reviewed for cycle 6' operation. 'In those cases where cycle 6' I
characteristics were conservative compared to those analyzed for previous cy-cles, no new accident analyses were performed.
The Technical Specifications have been reviewed, and the modifications required for cycle 6 operation are justified in this report.
Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cooling Systems, it has been concluded that ANO-1 can be operated safely for cycle 6! at a rated power level of 2568 MWt.
}
The cycle 6 core for ANO-1 will contain four once-burned lead test assemblies {
(LTAs). These assemblies,are part of a Department of Energy Extended Burnup f
l Test program. The LTA design is described in reference 2.
t Babcock s.Wilcox 1-1
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OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Arkansas Nuclear One, Unit 1 is the currently operating cycle 5.
This cycle 6 design is based on a nominal cycle 5 length of 455 effective full power days.,(EFPD).
/
s No anomalies occurred during cycle 5 that would adversely affect fuaJ,pprfor-mance during cycle 6.
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2-1 Babcock &Wilcox
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CENERAL DESCRIPTION The ANO-1 reactor core is described in detail in section 3 of the Arkansas Nuclear Station, Unit 1 Final Safety Analysis Report (FSAR).
The cycle 6 core contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore in-strument, guide tube. The fuel comprises dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4. The fuel assemblies in all batches have an average nominal fuel loading of 463.6 kg of uranium, with the exception of four batch 7 lead test assemblies (LTAs), which have a nominal loading of 440.'O kg uranium. The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters are given in Tables 4-1 and 4-2 for all fuel assemb'. tes except the LTAs; the corresponding parameters for the LTAs are included in reference 2.
Figure 3-1 is the fuel shuffle diagram for ANO-1, cycle 6.
The initial enrich-2:s, respectively.
ments of batches 6C, 7, and 8 are 3.19, 2.95, and 3.21 wt %
U All the batch SB assemblies and 23 of the twice-burned batch 6 assemblies will be discharged at the end of cycle 5.
The remaining 37 twice-burned batch 6 assembifes (designated batch 6C) will be shuffled to the core interior. Most of the once-burned batch 7 assemblies will be shuffled to locations on or near the core periphery. The 72 fresh batch 8 assemblies will be loaded in a symmetric checkerboard pattern throughout the core. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 6.
j Re'activitf is controlled by 61 full-length Ag-In-Cd control rods, 64 hPRAs, k-l and soluble boron shim.
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 6 locations of the 69 control rods and the group designations are indicated in Figure 3-3.
The core locations of the total pattern (69 control rods) for cycle 6 are identical to those of the reference cycle indicated in the reload report for ANO-1, cycle 5.s The group 3-1 Babcock & Wilcox
.g E
0 4
designations, however, differ between cycle 6 and the reference cycle in order to minimize power peaking. The cycle 6 locations and enrichments of the BPRAs are shown in Figure 3-4.
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f 3-2 Babcock & Wilcox
i Figure 3-1.
Core Loading Diagram for ANO-1 Cycle 6 x
I 7
7 7
7 7
^
K6 K2 08 K14 K10 7
8 7
8 7
8 7
8 7
8 L3 F
M3 F
MB r
413 F
111 7
8 7
8 7
8 7
8 7
8 7
C H15 F
L5 F
K4 F
K12 F
L11 F
R8 7
6 6C 8
6C 8
6C 8
6C 8
6C 8
7 D
rio r
F1 r
se F
v7 r
nio r
ain F
rs 8
7 8
6C 8
7 8
7 8
6C 8
7 8
E (P)
(P)
F E10 F
A6 F
N11 F
N5 F
FIS F
E6 F
7 7
8 6C 8
6C 8
6C 8
6C 8
6C 8
7 7
=
F F9 C12 F
F2 F
A7 F
P5 F
G15 F
F14 r
C4 F7 7
8 7
8 7
8 6C 6C 6C 8
7 8
7 8
7 G
(LTA)
(LT4) 89 F
09 F
M12 r
013 t12 c1 r
m F
57 r
n?
7 7
8 6C 8
6C 6C 6C 6C 6C 8
6C 8
7 7
y 16=H w11 uit r
P11 r
wie 02 N2 M14 F2 r
95 F
WS W1 7
8 7
8 7
8 6C 6C 6C 8
7 8
7 8
7 i
K (LTA)
(LTA) l P9 F
N9 F
E12 F
C13 P12 C3 F
E4 F
N7 F
P7 7
7 8
6C 8
6C 8
6C 8
6C 8
6C 8
7 7
L L9 012 F
L2 F
K1 F
B11 F
R9 F
L14 F
04 L7 8
7 8
6C 8
7 8
7 8
6C 8
7 8
M (P)
(P)
F W10 F
t1 F
D11 F
05 F
R10 F
u6 F
7 8
6C 8
6C 8
6C 8
6C 8
6C 8
7 M
010 F
P6 F
P6 F
Els F
P10 F
L15 F
06 7
8 7
8 7
8 7
8 7
8 7
0 A8 F
F5 F
G4 F
G12 F
F11 F
M1 7
8 7
8 7
8 7
8 7
p F3 F
D3 F
E8 F
D13 F
F13 7
7 7
7 7
R G6 G2 C8 G14 G10 I
Z 1
2 3
4 5
6 7
8 9
10 11 12 13 W
15 LEGEND
~
Batch Cycle 6 Location F = Fresh Fuel Assembly LTA = lead Test Assemblies i
P = Precharacterized Standard i
Mark B Assembly 3-3 Babcock s.Wilcox
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Figure 3-2.
Enrichment and Burnup Distribution, ANO-1 Cycle 6 Off 455 EFPD Cycle 5 8
9 10 11 12 13 14 15 3.19 3.19 3.19 3.21 3.19 3.21 2.95 2.95 H
20051 20070 22678 0
22664 0
18373 18118 3.19 3.21 2.95 3.21 2.95 3.21 2.95 K
21951 0
17715 0
18302 0
15576 3.19 3.21 3.19 3.21 2.95 2.95 20941 0
23462 0
13108 17818 3.19 3.21 2.95 3.21 19880 0
17831 0
l 3.19 3.21 2.95 N
16854 19952 0
t I
2.95 l
0 10521 P
R 1
i Initial Enrichment.
X*XX wt%235g l
BOC Burnup, l
XXXXX mwd /mtU 3-4 Babcock & Wilcox
a-
- ...-.s......
Figure 3-3.
Control Locations and Group Designations for ANO-1 Cycle 6 X
f A
B 4
7 4
C 2
6 6
2 D
7 8
5 8
7 E
2 5
1 1
5 2
f F
4 8
3 7
3 8
4 g
6 1
4 4
1 6
H W
7 5
7 2
7 5
7 y
K 6
1 4
4 1
6 4
8 3
7 3
8 4
[
M l
2 5
1 1
5 2
N 7
8 5
8 7
0 2
6 6
2 l
I 4
7 4
P a
lI II l
1.
(
l 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 t
x Group Number Group No. of Rods Function _
j 1
8 Safety l
2 9
Safety 3
4 Safety 4
12 Safety 5
8 Control 6
8 Control 7
12 Control 8
8 APSRs Total 69 i
i f
Babcock s.Wilcox 3-5
Figure 3-4.
LBP Enrichraent and Distribution, ANO-1 Cycle 6 8
9 10 11 12 13 14 15 i
H 1.4 1.1 1
4 K
1.4 1.4 0.2 l
L 1.4 1.4 O.5 l
M 1.4 1.4 l 1.1 N
1.4 1.1 0.2 0
1.1 0.5 0.2 l
i P
o.2
/
R l
X,X LBP Concentration, wt % B4C in A1203 l
i l
3-6 Babcock s.Wilcox
4.
FUEL SYSTEM DESIGN 4.1.
Fuel Assembly Mechanical Design The type of fuel assemblies and pertinent fuel design parameters for ANO-1 cy-cle 6 are listed in Table 4-1.
All fuel assemblies listed are identical in concept and are mechanically interchangeable. All results, references, and identified conservatisms presented in section 4.1 of the cycle 5 reload report are applicable to Mark B4 assemblies.
In addition to the assemblies listed, four lead test assemblies (LTAs) are in their second cycle of irradiation in cycle 6.
One standard Mark B fuel assembly. contains annealed guide tubes to compare with the LTAs. The analysis and justification for the LTAs and an-nealed guide tubes are reported in reference 2.
Retainer assemblies will be used on the fuel assemblies that contain BPRAs to provide positive retention during reactor operation. This will be the third cycle of operation for the retainer assemblies. The justification for the de-sign and use of the retainers is described in reference 4, and is applicable to ANO-1, cycle 6.
Similar retainer assemblies will be used on the two fuel assemblies containing the regenerative neutron sources.
l 4.2.
Fuel Rod Design l
The mechanical evaluation of the fuel rod is discussed below.
4.2.1.
Cladding Collapse The batch 6 fuel is more limiting than batches 7 and 8 because of its previous incore exposure time. The batch 6 assembly power histories were analyzed to determine the most limiting three-cycle power history for creep collapse.
This worst-case power history was then compared against a generic analysis to l
ensure that creep-ovalization will not affect fuel performance during ANO-1 cycle 6.
The generic analysis was performed based on reference 5 and is ap-plicable for the batch 6 fuel design.
I l
4-1 Babcock &)Milcox
ne creep collapse analysis predicts a collapse time greater than 35,000 ef-r fective full-power hours (EFPH), which is longer than the maximum expected residence time of 28,992 EFPH (Table 4-1).
4.2.2.
Cladding Stress The ANO-1 stress parameters are enveloped by a conservative fuel rod stress analysis. For design evaluation, the primary membrane stress must be less than two-thirds of the minimum specified unirradiated yield strength, and all stresses must be less than the minimum specified unirradiated yield strength.
In all cases, the margin is greater than 30%. The following conservatisms with respect to the ANO-1 fuel were used in the analysis:
1.
Low post-densification internal pressure.
2.
Low initial pellet density.
3.
High system pressure.
4.
High thermal gradient across the cladding.
4.2.3.
Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding plastic tensile circumferential strain. The pellet is designed to ensure that cladding plas-tic strain is less than 1% at design local pellet burnup and heat generation rate. The cesign burnup and heat generation rate are higher than the worst-case vdlues that ANO-1 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 value for the cladding ID.
4.3.
Thermal Design All fuel in the cycle 6 core is thermally similar. The design of the four batch 7 lead test assemblies is such that the thermal performance of this fuel is equivalent to or slightly better than the standard Mark B design used l
I in the remainder of the fuel.
The thermal design analysis of the LTAs using
(
the TACO-2 code is described in referen<,e 2.
8 The results of the thermal design evaluation of the cycle 6 core are summarized in Table 4-2.
Cycle 6 core protection limits were based on a linear heat rate (LHR) to centerline fuel melt of 20.15 kW/f t as determined by the TAFY-3 code 7, with no credit taken for the increased LHR capability of the LTA fuel. The maximin fuel rod burnup at EOC 6 is predicted to be less than 42,000 mwd /mtU.
4-2 Babc0Ck 8,Wilcox
s Fuel rod internal pressure has been evaluated with TAFY-3 for the highest burn-up fuel rod and is predicted to be less than the nominal reactor coolant system pressure of 2200 psia.
4.4.
Material Design The chemical compatibility of all possible fuel-cladding-coolant-assembly in-teractions for the batch 8 fuel assemblies is identical to that of the present l fuel.
4.5.
Operating Experience Babcock & Wilcox operating experience with the Mark B,15 x 15 fuel assem 31y has verified the adequacy of its design. As of May 31, 1982, the following experience has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel asrembly:
Max F mp Cumulative net Current electrical output,( )
Reactor cycle Incore Discharged MWh j
Oconee 1 7
41,600 40,000,
37,367,569 l
Oconee 2 6
20,565 36,800 34,229,828 f
Oconec 3 7
19,650 35,072 35,984,594 I
TMI-1 5
25,000 32,400 23,840,053 ANO-1 5
i 32,484 33,220 31,843,210 Rancho Seco 5
33,280 37,730 27,752,205 Crystal River 3 4
20,700 29,900 18,359,090 3
Davis-Besse 1 3
}
24,617 25,326 11,473,226.
\\
As of May 31, 1982.
k (b)As of January 31,1982.k
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4-3 Babcock & Wilcox L
s Table 4-1.
Fuel Design Parameters and Dimensions Batch 6 Batch 7 Batch 8 Fuel assembly type Mark B4 Mark B4, Mark B4 Nbrk BEB
\\
64 Mark B, 72 No. of assemblies 37 4 Mark BE3 Fuel rod CD (nom), in.
0.430 0.430 0.430 Fuel rod ID (nom), in.
0.377 0.377 0.377 6
i i
i Flexible spacers Spring Spring Spring l
l
, Rigid spacers, type Zr-4 Zr-4 Zr-4
'Undensified active fuel
' 142.25 141.8 141.8 l
length (nom), in.
l
- Fuel pellet OD (mean 0.3695 0.3686 0.3686 specified), in.
I E Fuel pellet initial 94.0 95.0 95.0 l
density (nca), % TD
- Initial fuel enrichment, 3.19 2.95 3.21 23s I
,'wt %
U l
' Average burnup, BOC, 21,630 16,554 0
mwd /mtU
-Cladding collapse
>35,000
>35,000
>35,000 time, EFPH
, Estimated residence 28,344 28,992 26,616 time EFPH (max) l l
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~4-4 Babcock & Wilcox
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Table 4-2.
Fuel Thermal Analysis Parameters Batch 6C Batch 7 Batch 8 64(a) 72 i
No. of assemblies 37 initial density, % TD 94.0 95.0' 95.0 Pellet diemeter, in.
0.3695 0.3686 0.3686 Stack height, in.
142.'25 141.80 141.80
' ansified Fuel Parameters Pellet diameter, in.
0.3646 0.3649 0.3649 Fuel stack height, in.
140.5 140.74 140.74 Nominal linear heat rate 5.80 5.79 5.79 at 2568 MWt, kW/ft Avg fuel temperature at 1320 1310 1310 l
nominal LHR, F LHR capability, kW/f t( )
20.15 20.15 20.15 Nominal core avg LHR = 5.79 kW/f t at 2568 MWt.
- Four LTAs were also analyzed; the results are reported in reference 2.
( } Centerline fuel melt based on fuel specification values.
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4-5
I 5.
NUCLEAR DESIGN w
5.1.
Physics Characteristics The Table 5-1 lists the core physics parameters of design cycles 5 and 6.
b values for both cycles were generated using PDQ07. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are 1
to be expected between cycles. Figure 5-1 illustrates a representative rela-tive power distribution for the beginning of cycle 6 at full power with equi-librium xenon and nominal rod positions.
Operational changes as well as differences in cycle length, feed enrichment.
BPRA loading, andishuffle pattern make it difficult to compare the physics parameters of-cycles 5 and 6.
Calculated ejected rod worths and their ad-herence.to criteriatare 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 stock red worth for cycle 6 is less than that for the design cy-l l
cle 5 at BOC and EOC. All safety criteria associated with these worths are l
l The adequacy of the shutdown margin with cycle 6 stuck rod worths is f
met.
demonstrated in Table 5-2.
The following conservatisms were applied for the i
shutdown calenlations:
1.
Poison material depletion allowance.
2.
10% uncertainty on net rod worth.
3.
Flux redistribution penalty.
Flux redistribution was accounted for since the shutdown analysis was calcu-l {*
1ated using a two-dimensional model. The reference fuel cycle shutdown mar-5 gin is presented in the ANO-1 cycle 5, reload report.
l 5.2.
Analytical Input 1
The cycle 6 incore measurement calculation constants to be used for computing core Wwer distributions were prepared in the same manner as those for the reference cycle.
Babcock &Wilcox 5-1
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5.2.
Changes in Nuclear Design There are no significant core design changes between the reference and reload cycles. The calculational methods and design information used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle. There is one significant operational change from the reference cycle: the withdrawal of the APSRs during the last 37 EFPD of cycle 6.
The calculated stability index without APSRs is -0.031 h-1, which demon-strates the axial stability of the core. The operating lim'ts (Technical Specification changes) for the reload cycle are given in section 8.
Physics Parameters for ANO-1, Cycles 5 and 6(a)
Table 5-1.
Cycle 5(b)
Cycle 6(c)
Cycle length. EFPD 455 387 Cycle burnup, mwd /mtU 14,259 12,128 Avg corn burnup, EOC, mwd /mtU 23,18; 23,009 Initial core loading, atU 82.0 82.0 Critical boron -- BOC, ppm (no Xe)
HZP(d), group 8 ins 1538 1463 HFP, group 8 ins 1370 1273 Control rod worths -- HFP, BOC, % ok/k Group 6 1.26 1.13 Croup 7 1.47 1.36 Group 8 0.46 0.42 Control rod worths -- HFP, EOC, % ok/k Group 7 1.59 1.40 Max ejected rod worth -- HZP, % ak/k(*}
BOC (N-12), group 8 ins 0.53 0.53 0.46 350 EFPD (N-12), group 8 ins EOC (M-11), group 8 out 0.58 0.47 Max stuck rod worth -- HZP, % ok/k BOC (N-12) 1.57 1.50 350 EFPD (N-12), group 8 ins I 1.63 EOC (N-12), group 8 out 1.74 1.43 Critical boron -- EOC, ppm HZP, group 8 out, no Xe 487 704 HFP, group 8 out, eq Xe 17 95 Power deficit, HZP to HFP, % Ak/k BOC 1.33 1.68 EOC 2.39 2.38 5-2 Babcock & Wilcox
Table 5-1.
(Cont'd)
Cycle 5 Cycle 6 Doppler coeff - HFP, 10-s (ak/k/F)
BOC (no Xe),
-1.52
-1.54 EOC (eg Xe)
-1.82
-1.82 Moderator coeff - HFP, 10-" (ak/k/F)
BOC, (no Xe, crit ppm, group 8 ins)
-0.49
-0.84 EOC, (eq Xe, O ppm, group 8 out)
-3.00
-2.89 Boron worth - HFP, ppm /% Ak/k BOC 122 123 EOC 103 109 Xenon worth - HFP, % ak/k BOC (4 EFPD) 2.58 2.57 EOC (equilibrium) 2.70 2.69 Effective delayed neutron fraction - HFP BOC 0.0063 0.0063 EOC 0.0052 0.0053 I") Cycle 6 data are for the conditions stated in this report.
The cycle 5 core conditions are identified in reference 3.
(b) Based on 329 EFPD at 2568 MWt, cycle 4.
(*) Based on 455 EFPD at 2568 MWt, cycle 5.
(d)HZP denotes hot zero power (532F T
), HFP denotes hot full power (579 Tavg)-
(*) Ejected rod worth for groups 5 through 7 inserted, group 8 as stated.
l 5-3 Babcock & Wilcox
Table 5-2.
Shutdown Margin Calculations for ANO-1, Cycle 6
% Ak/k
% Ak/k
% Ak/k Available Rod Worth Total rod worth, HZP 8.80 9.27 9.01 Worth reduction due to
-0.42
-0.42
-0.42 poison material burnup Maximum stuck rod, HZP
-1.50
-1.63
-1.43 Net Worth 6.88 7.22 7.16 Less 10% uncertainty
-0.69
-0.72
-0.72 Total available worth 6.19 6.50 6.44 Required Rod Worth Power deficit, HFP to HZP 1'.97 2.35 2.38 Allowable inserted rod worth 0.41 0.50 0.55 Flux redistribution 0.48 0.93 0.95 Total required worth 2.86 3.78 3.88 Shutdown margin (total 3.33 2.72 2.56 available worth minus total required worth).
Note: The required shutdown margin is 1.00% Ak/k.
I 5-4 Babcock s,Wilcox '
.g.
Figure 5-1.
ANO-1 Cycle 6. BOC (4 EFPD) Two-Dimensional Relative Power Distribution - Full Power Equilibrium Xenon, Normal Rod Positions 8
9 10 11 12 13 14 15 1.28 lI I
0.96 0.49 H
0.90 0.96 1.04 1.27 1.14 K
0.98 1.21!
1.16 1.25 1.16i
- 1. lie 0.50 i
L 1.12,'
1.24 0.98 1.28 0.9d 0.39 1
M 1.15 1.24 1.06, 0.93 i
I N
1.08 1.04!
0.47 I
0 0.58
~
P R
d N. X.
Inserted Rod Group No.
XXX Relative Power Density i
Babcock & Wilcox 5-5
f 6.
THERMAL-HYDRAULIC DESIGN The fresh batch 8 fuel is hydraulically and geometrically similar to the pre-viously irradiated batch 6C and 7 fuel. The four batch 7 (LTAs) have been
{
analyzed to ensure that they are never the limiting assemblies during cycle 6 operation. The results of the thermal-hydraulic analysis for the LTAs are pro-vided in reference 2.
The thermal-hydraulic design evaluation supporting cycle 6 operation is based on methods and models described in references 1, 3, and 4.
The cycle 6 thermal-hydraulic design is identical to cycle 5.
The thermal hydraulic design condi-tions for cycles 5 and 6 are summarized in Table 6-1.
The rod bow penalty for cycle 6 is based on a procedure approved by reference 10.
A rod bow penalty was calculated for each fuel batch based on the highest assembly burnup in that batch. The penalty was then applied to the minimum DNBR determined for each batch. Results are summarized in Table 6-2.
The fresh batch 8 fuel is the most restrictive both before and after the rod bow l
penalties are applied. Therefore, the cycle 6 rod bow penalty is based on the highest predicted batch 8 assembly burnup of 16,883 mwd /mtU. The resulting rod bow penalty is 0.1% when 1% credit is taken for use of the flow area re-duction hot channel factor in DNBR calculations.
l 6-1 Babcock & Wilcox
,.u..
,-.>.....i 2
i Table 6-1.
Maximum Design Conditions, Cycles 5 and 6 i
Cycle 5 Cycle 6.
Design power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 Vessel inlet / outlet coolant temp 555.6/602.4 555.6/602.4 at 100% power, F Reference design radial-local 1.71 1.71 power peaking factor Reference design axial flux 1.5 cosine 1.5 cosine shape Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flou area 0.98 0.98 Active fuel length, in.
140.2 140.2
'~
Avgheatfluxa 100% power, 175 175 10 Btu /h-ft2(a Max heat flux at 100% power, 449 L 449 8
I 10 Btu /h-ft:(b) 3 CHF correlation BAW-2 BAW-2 Minist;m DNBR At 112% power 2.05 l 2.05 At 108% power 2.18 2.18 At 100% power 2.39,
2.39 I") Heat flux was based on densified length (in the hottest core location).
Based on average heat flux with reference peaking.
I 1
i Babcock & Wilcox 6-2
.. c Table 6-2.
DNBR Red Bow Penalty, ANO-1 Cycle 6 Rod bow DNBR less Max BU,
- penalty, Max pin DNBR rod bow Batch mwd /mtU peak BAW-2 penalty (a) 6C 36,818 5.3 1.27 3.38 3.23 7
33,181 4.3 1.25 3.45 3.34 8
16,883 1.1 1.46 2.77 2.77
(* Includes 1% credit for flow area reduction hot channel factor.
i l
l l
l l
l l
l 6-3 Babcock & Wilcox
i 7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cycle 6 parameters to determine the effect of the cycle 6 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 eval-uated and are reported in reference 8.
Since batch 8 reload fuel assemblies contain fuel rods whose theoretical density is higher than those considered in the reference 8 report, the conclusions in that reference are still valid.
The radiological dose consequences of the accidents presented in Chapter 14 of the FSAR were re-evaluated for this reload report. The reason for the re-eval-uation is that, even though the FSAR dose analyses used a conservative basis for the amount of plutonium fissioning in the core, improvements in fuel man-agement techniques have increased the amount of energy produced by fissioning ass ass, the mixture of plutonium. Since Pu has different fission yields than U
2s 2ss fission product nuclides in the core changes slightly as the Pu to U
fission ratio changes; i.e., plutonium fissions produce more of some nuclides and less of others. Since the radiological doses associated with each accident are impacted to a different extent by each nuclide and by various mitigating factors and plant design features, the radiological consequences of the FSAR accidents were recalculated using the specific parameters applicable to cycle 6.
The bases used in the dose calculation are identical to those presented in the FSAR except for the following two differences.
1.
The fission yields and half-lives used in the new calculations are based on more current data.
2.
The steam generator tube rupture accident evaluation considers the in-creased amount of steam released to the environment via the main steam re-lief and atmospheric dump valves because of the slower depressurization due to the reduced heat transfer rate caused by tripping of the reactor 7-1 Babcock &)Milcox L
coolant pumps upon actuation of the high-pressure injection (a post-TMI-2 modification).
A comparison of the radiological doses presented in the FSAR to those calcu-lated specifically for cycle 6 (Table 7-1) show that some doses are slightly higher and some are slightly lower than the FSAR values. However, except for MHA, all doses are either bounded by the values presented in the FSAR or are a small fraction of the 10 CFR 100 limits; i.e., below 30 Rem to the thyroid or 2.5 Rem to the whole body. For the MHA,,the 2-hour EAB, 30-day LPZ thyroid doses are 157.3 and 72.9 Rem, respectively (52 and 24% of 10 CFR 100 limits, respectively), while the 2-hour EAB, 30-day LPZ whole body doses are 7.1 and 2.4 Rem, respectively (28 and 10% of 10 CFR 100 limits, respectively).
The small increases in some doses are essentially offset by reductions in other doses. Thus, the radiological impact of accidents during cycle 6 are not significantly different than those described in Chapter 14 of the FSAR.
7.2.
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:
core thermal parameters, thermal-hydraulic parameters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.
Core thermal properties used in the FSAR accident analysis were design oper-ating values based on calculational values plus uncertainties. The thermal parameters for Batches 66, 7, and 8 are given in Table 4-2.
The cycle 6 thermal-hydraulic maximum design conditions are compared to the previous cycle 5 values in Table 6-1.
These parameters are common to all the accidents considered in this report. The key kinetics parameters from the FSAR and cycle 6 are compared in Table 7-2.
A generic LOCA analysis for a B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103).11 This analysis is generic since the limiting values of key parameters for all plants in this category were used. Furthermore, the combination of average fuel temperatures as a function of LHR and lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the analysis and the LOCA limits 1
7-2
reported in BAW-10103 and substantiated by reference 12 provide conservative results for the operation of the reload cycle. Table 7-3 shows the bounding values for allowable LOCA peak LHRs for ANO-1 cycle 6 fuel. The basis for two sets of LOCA limits is provided in reference 13.
It is concluded from the examination of cycle 6 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ANO-1 plant's ability to operate safely during cycle 6.
Considering the previously accepted design basis used in the FSAR i
and subsequent cycles, the transient evaluation of cycle 6 is considered to be bounded by previously accepted analyses. The initial conditions for the tran-sients in cycle 6 are bounded by the FSAR, the fuel densification report, and/
or subsequent cycle analyses.
l 7-3 Babcock a,Wilcox
~_
Table 7-1.
Comparison of FSAR and Cycle 6 Doses Accidents FSAR dose, Rem Cycle 6 dose, Rem Fuel Handling 2-hour thyroid (EAB) 0.920 1.060 2-hour whole body (EAB) 0.540 0.5888 30-day thyroid (LPZ)
(a) 0.1582 30-day whole body (LPZ)
(a) 0.08787 1
Rod Ejection i
2-hour thyroid (EAB) 11.40 12.13 2-hour whole body (EAB) 0.014 0.0132 30-day thyroid (LPZ) 8.3 9.037 30-day whole body (LPZ) 0.0099 0.009724 Steam Line Break 2-hour thyroid (EAB) 1.6 1.712 2-hour whole body (EAB)
(a) 0.01328 l
30-day thyroid (LPZ)
(a) 0.2555 30-day whole body (LPZ)
(a) 0.001981 Steam Generator Tube Failure 2-hour thyroid (EAB) 0.0087 6.139 2-hour whole body (EAB) 0.16 1.13 30-day thyroid (LFZ)
(a) 0.9162 30-day whole body (LPZ)
(a) 0.1686 Waste Gas Tank Release i
l 2-hour thyroid (EAB) 0.22 0.05409 2-hour whole body (EAB)
(a) 3.366 30-day thyroid (LPZ) 4.5 9.008072 30-day thyroid (LPZ)
(a) 0.5023 LOCA 6
2-hour thyroid (EAB) 3.6 3.992 4
2-hour whole body (EAB) 0.057 0.05873 l
30-day thyroid (LPZ) 1.66 2.038 30-day whole body (LPZ) 0.043 0.0446 Maximum Hypothetical Accident 2-hour thyroid (EAB) 153.0 157.3 2-hour whole body (EAB) 10.0 7.099 30-day thyroid (LPZ) 64.1 72.91 30-day whole body (LPZ) 3.4 2.439
(*}Not calculated in FSAR.
Babcock a,Wilcox 7-4
-_m r
. c..
Table 7-2.
Comparison of Key Parameters for Accident Analysis I
FSAR and densification ANO-1 Parameter report value cycle 6 Doppler coeff (BOC), 10-5 Ak/k/*F
-1.17
-1.54 Doppler coeff (EOC), 10-5 Ak/k/*F
-1.30
-1.82 Moderator coeff (BOC), 10-" Ak/k/*F 0.0(a)
-0.84 Moderator coeff (EOC), 10-" Ak/k/*F
-4. 0(b )
-2.89 All-rod group worth (HZP), % Ak/k 12.9 8.80 Initial boron concentration, ppm 1150 1273 Boron reactivity worth (HFP),
100 123 ppm /% Ak/k Max ejected rod worth (HFP), % Ak/k 0.65 0.27 I
Dropped rod worth (HFP), % Ak/k 0.65 0.20
(* +0.5 x 10-" Ak/k/*F was used for the moderator dilution analysis.
I
-3.0 x 10-" Ak/k/*F was used for the steam line failure analysis.,
i Table 7-3.
Bounding Values for Allowable LOCA Peak Linear Heat Rates I
l Allowable Allowable Core peak LHR, peak LHR, elevation, first 50 EFPD, balance of cycle, ft kW/ft kW/ft 2
14.5 15.5 4
16.1 16.6 6
17.5 18.0 8
17.0 17.0 10 16.0 16.0 7-5 Babcock & )Milcox i
8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 6 operation to ac-count for changes in power peaking and control rod worths inherent with the transition to 18-month LBP fuel cycles.
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. The following pages con-tain the revisions to previous Technical Specifications.
l l
f I
i l.
I t
l 8-1 Babcock & VVilcox l
. ~....
. ~.....
Figure 8-1.
Core Protection Safety Limits (Tech Spec Figure 2.1-2)
Tnermal Power Level, 5 FP 120
(-42.56,112)
(24.64,112) 1125 FP Il = 2.90 l ACCEPTABLE l
82 = -1.26 i
4 PUMP I OPERATION l
- 100 1
l l
(38.08,95.1)
(.48.39,95.1) 88.74% FP I
g I
I ACCEPTABLE l
l l 4 AND 3 PUMP 80 l
l l
OPERATION
( 48.39,71.84)
(38.08,71.84) 1 I
l 82.115 FP l
UNACCEPTABLE
- 60 l UNACCEPTABLE OPERATION l
ACCEPTABLE 4,3 AND 2 PUMP l OPERATION l
l l
OPERATION l
l
( 48.39,45.21) l l
(38.08,45.21) l
- 40 l
l l
l l
l l
l M
S!
20 E
i lpl ill 5l i
l n
n is 1
5lE El EI l
t I
i I
i l
ll I
l 60
-40
-20 0
20 40 60 Reactor Power imoalance, 5 l
l l
l l
l 8-2 Babcock a.Wilcox 1
l 1t
'I lll 1l l
l i,
l I
U ON 6
PA i
0 EC RC AE TP
(
I T OA E
3
(
NB
(
l L
,4 E
3
=
F 3
4 3
4 i
8 1
g 1
5 u
5 2
r 4
i 4
8 g ;:
)
7
(
e 0
)
3 5
1 8
,8 I
l I
I l lI I l I l
l
(
(
2 1
1
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5 R
5 I
P e
1 4
r
)
o g;:
a 8
c 2 t 0
)
}
t i
i
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Tc l
I l I l
g l l
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et
<8E35
<yt=$
yE=e r
ci P
m hv
- w. "
, 2t n a, Et Ek a
e o
S w
l p S e
oE52=
ogOez yO 2,.
P ey r
o cs 8
I w
t e
Fe m
4 3
D 0 r
im a
g 1
1 L
uM l
- s ;
2 4
6 0
2 e
ra a
i 0
0 0
0 0
v ex nc e
i 1
l l I
l l l
l I i l l
l e
l 2m
(
u 9
5 3m 5
5
(
F 2A 1
,9 P
) l 2
8 1
l 0 [ng
)
0 o
5 wa I
l j l
l l l l
l l
4 b
(
(
E
)
l
(
2 2
2 e
2
,5
,5
,5
=
S 3
5 U
e ON 8
t 8
PA 5
1 1
p B
4 4
3 EC
)
o i
RC 2
i a
0
)
)
b AE 7
n c
TP 5
t o
I T c
OA s
k NB L
s.
E W
i lc 6
e 0
i x
[4 i!
3
s t
l Figure 8-3.
-Boric Acid Addition Tank Volume and Co.icentration
, Requirements Vs RCS Average Temperature (Tech Spec Figure 3.2-1) t' l
'0PERAft0N AROVE AND TO THE LEFT OF THE CURVES IS ACCEPTABLE
}
8700 PPM 6000 - *..
9500 PPM s
10,000 PPM 2
g' 5000 j
2 y
e g
s 12,000 PPM '
g 4000 oa a
u i
3000 E
a I
2000 TEMP. F REQUIRE 0 VOLUME, GAL.,
875) PPM 9500 PPM 10,000PPN 12,000PM 579 6115 5568 5274 4354 532 5184 4720
- 4470, 3689,
500 4375 3984' 3773 -
- 3117, 1000 400 2366 2156 2042, '
1688 300 981 834 847 7El 200 0
0 0
0 C
0 200 300 400 500 600 700 800 RCS Average; Temperature, F 8-4 Babcock s Wiicox
_._s_.
i Figure 8-4.
Rod Position Limits for Four-Pump Operation i
From 0 to 60 EFPD - ANO-1, Cycle 6 l
(Tech Spec Figure 3.5.2-1A) 110 (I'
}
j 00,102)
(176,102)
/
(270,92) 90 SHUTDOWN MARGIN Lluli POIER LEVEL CUTOFF 80 (258,80) t 70 OPERATION RESTRICTED E
OPERAT;0H IN THIS
= 60 REGION 13 NOT
~
ALLOWE0 i t $g (118,50)
(175,50)
PERMISSIBLE t 40 OPERATING j
REGION 30 20 (57,15) lo 0
O 20 40 60 80 100 120 140 160 180 200 220 240 260 240 300 i
p zo 4p 6,0 8,0 1,00 i
GROUP 7 0
29 (0
6,0 8p 10,0 i
I GROUP 6 O
2,0
~4p 6,0 8,0 100 Roo Inner, 5 30 GROUP 5 L
8-5 Babcock s.Wilcox
/
'^
s
.u _
._*_.-c 1
Figure 8-5.
Rod Position Limits for Four-Pump Operation From 50 to 200 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-1B) 110 (270,102)
,(300,102)
(176,102)
-(270,92)
SHUTOOWN BARGIN LIMIT I
90 POWER LEVEL CUT 0FF (258,80) to E
OPERATION IN THIS R
0 REGION IS NOT
- 60 ALLOWE0
- 50 (118,50)
(175,50)
I D
40 r
1 30 PERNISSIBLE OPERATING l
20 REGION (57,15) 10, (0.8) 0 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 g
20 4,0 6,0 8,0 1,00 GROUP 7 i
g 2,0 4,0 6,0 8,0 100 GROUP 6 g
2,0 4,0 60 8,0 1,00 GROUP 5 Roa index, 5 30 l,
l l
l 8-6 Babcock 3 Wilcox N-.
........... ~.
I l
Figure 8-6.
Rod Position Limits for Four-Pump Operation From 200 1 10 to 350 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-1C) 110 (219,102)
(270,102)
(300.102) i 100 90 pagER LEVEL t>270,92)
CUTOFF (258,80) 80 RESTRICTED SHUT 00EN MARGIN LIMli REGION
=
10 E
OPERATION IN THIS 60 3
REGION IS NOT ALLOWED 50 (157,50)
(gy$,$$)
. g 40 PERMISSIBLE OPERATING E
REGION
^
30 20 10 (78,15)
(0.7) o e
i 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 4
,0 20 40 60 80, 10,0 i
f 2,0 4,0 60 80 100 4
f 2,0 4,0 6,0 80 100 GROUP 5 Roa inaex, 5 50 8-7 Babcock & Wilcox
I Figure 8-7.
Rod Position Limits for Four-Pump Operation After 350 2 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-1D) 110
, 300,102)
(
(265,102) 10 0 100 7
POWER 90 LEVEL (260,92)
SHUT 00EN MARGIN LIMIT CUTOFF 80 (225,80)
OPERATION IN THIS 4
E 60 REGION IS NOT ALLOgE0
=
l22 50 (156,50)
(175,50)
- 40
'I 30 PERNISSIBLE OPERATING 20 sg (80,15)
(0.6l 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
2,0 40 6,0 8,0 lpo 0
20 40 60 80 1,00 i
0 2,0 40 6,0
.,0 lp0 8
l GROUP 5 Roa Inaex, s 30 I
l l
i l
l l
l 8-8 Babcock a.Wilcox
, -.. _ _ = = -
m.. m Figure 8-8.
Rod Position Limits for Threa-Pump Operation From 0 to 60 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-24) 110 100 90 80 (176,77)
(250,77) 70 SHUTDOWN FARGIN LIMIT OPERATION E
RESTRICTED g 60 0
OPERATION IN THIS REGION IS NOT PERMISSIBLE E 50 ALL0tED GPERATING REGION (118,38) 0 40 E
(134,38) 30 20 l
IU (57,11)
(0,6) 0 e
i e
i e
i e
e i
e i
i e
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80 1,00 t
t t
t t
GROUP 7 0
20 40 60 80 100 L
f f
f f
p GROUP 6 1
0 20 40 60 80 100 Roa Index, 5 NO
'I I
e t
i
+
GROUP 5 1
8-9 Babcc 71; & Wilcox
^
Figure 8-9.
Rod Position Limits for Three-Pump Operation I
From 50 to 200 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-2B) 110 100 90 (250,77) l 80 (178,77) 70 SHUT 00tN NARGIN LIMIT OPERATION y
E 60 RESTRICTE E
OPERATION IS THIS
,"_ M REGION IS NOT ALLORED Q
m 40 (118,38) m (134,38)
E 30 PERNISSIBLE OPERATING REGION 20 l
10 (57,11)
(0,6,)
o 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 1
I GROUP 7 0
20 40 60 80 100 t
I t
i t
t
.G OW 6 0
20 40 60 80 100 Roa inaex, 5 10 GROUP 5 8-10 Babcock & Wilcox
.....=.... ---...,.. 4*.~'J I
Figure 8-10.
Rod Position Limits for Three-Pump Operation From 200 1 10 to 350 ! 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-2C) 110
{
100 90 (219,77)
(250,77) 80 SHUTDOWN 70 MARGIN LIMIT
~
=
OPERAil0N IN THIS REGION
$g+
g 60 15 NOT ALLOWE0 n
% 50 I
e d 48 g
(151,38)
PERiflSSIBLE
{
^
OPERATING 30 l
REGION t
20 10 (yg,it) 0 8
O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80 100 i
e t
I t
a f
l GROUP 7 l
0 2,0 4,0 6,0 8,0 100 GP.00P 6 0
2,0 40 6,0 go 100 Roa Index, 5 10 GROUP 5 6
3 I
l 8-11 Babcock & Wilcox
Figure 8-11.
Rod Position Limits for Three-Pump Operation After 350 10 EFPD - ANO-1 Cycle 6 (Tech Spec Figure 3.5.2-2D) 110 100 90 80 (227.77)
SHUTDOWN 70 MARGIN 3
OPERAil0N IN THIS REGION LIMIT IS NOT ALLOWED
==
60
'E a
50 a
f 40 30 (156,38)
PERillSSIBLE OPERATING REGION 20 10 (0.5 (80,81) 4 0
8 i
i i
i e
i e
i a
e i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 303 i
0 20-40 60 80 100 i
I i
a i
I I
ERCUP 7 0
20 40 60 80 100 i
a e
i 8
I i
GROUP 6 0
20 40 60 80 100 e
i e
I i
i Roo index, s 30 GROUP 5 l
l 1
8-12 Babcock s.Wilcox
Figure 8 -
J.od Position Limits for Two-Pump Operation From 0 to 60 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-2E)
Il0 100 90 I
80 70 l
I OPERATION RESTRICTED 60 I 1,52)
OPERATION IN THIS (176.52) 0 REGION IS NOT SHUTDOWN ALLOWED HARGIN 40 LIMIT I
j PERillSSIBLE 30 OPERATING REGION (1l8,26) 20 10 -(0,5) 1 (57,8) 0 i
e i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 2p j0 6p 8,0 10,0 0
20
'40 60 80 100 8
I I
I i
g Group 6
?
.{0 j0 6p qu 100 Roo incer, $ 30 GROUP 5 l
l i
Babcock s.Wilcox 8-13
Figure 8-13.
Rod Positico Limits for Two-Pump Operation From 50 to 200 ! 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-2F) 110 100 90 80 10 OPERATION RESTRICTE0 E
OPERATION IN THIS
", 60 REGION IS NOT
,q (176,52) -
(181,52)
ALLOWE0 50 SHUTDOWN E
MARGIN LIMIT 40 N
E PERillSSIBLE O
OPERATING REGION (118,26) 20 10 (51,8)
(0,5 h 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80 100 I_
l t
I i
1 GROUP 7 0
20 40 60 80 100 t
i I
I t
1 GROUP 6 0
20 40 60 80 100 t
t t
t f
f Roo index, 5 30 GROUP 5 8-14 Babcock & Wilcox
Figt.re 8-14.
Rod Position Limits for Two-Pump Operation From 200 1 10 to 350 ! 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-2G) 110 100 I
~
90 80 70 E=
60 (219,52)
SHUTDOWN
^
50 3
MARGIN e
OPERATION IN THIS REGION LIMIT
~
f NOT ALLOWE0 E
PERMISSIBLE 30 a-OPERATING (157,26)
RANGE 20 10 (0,4 47g,g) 0 8
2 8
8 I
I I
O 20 40 60 80 100 120 140 160 180 200.220 240 260 280 300 l
1 l
0 20 40 60 80 100 a
a 1
i t
l GROUP 7 0
20 40 60 80 100 I
i a
i t
l l
i GROUP 6
{
2,0 4,0 j0 8,0 100 GROUP 5 Roo inoex, 5 50 l
1 8-15 Babcock a Wilcox
l
...._v._m.__.
Figure 8-15.
Rod Position Limits for Two-Pump Operation After 350 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-2H) l 110 100 90 80 70 60 E
(227.52)
SHUTDOWN
~
OPERATION IN THIS REGION IS NOT ALLONEO 40 PERillSSIBLE OPERATING E
30 E
(156,26)
REGION 20 10
'(80,5)
, p 0
i e
i e
a e
i i
i e
i i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 p
2,0 4,0 6,0 80 100 GROUP 7 0
20 40 60 80 100
(
n i
i e
i e
i GROUP S 0
20.40 60 80 100 i
i a
a i
i Roo inou, 5 50 GROUP 5 8-16 Babcock 8.Wilcax
. _._ _. i Figure 8-17.
Operational Power Imbalance Envelope for Operation From 50 to 200 2 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-3B)
(-23,102) c-n(20,102)
(-23,92)
(20,92) 90 PERMISSIBLE OPERATING,
REGION
(-34,80)
.. '8 0 (32,80)
- 70
((jjgCTED l
fi O-E 3-60 N
- 50 i
5
=
I 40 m
30 I
i 1
20 10 t
f I
I I
I l
3 I
-50
-40
-30
-20
-10
.0 10 20 33 40 50 AxialPowerImbalance,%f
(
1 i
l i
8-18 Babcock s Wilcox
i l
Figure 8-16.
Operational Fower Imbalance Envelope for Operation l
From 0 to 60 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-3A)
.,110
(-17.5,102)C o(20.102) i' I
- 100 i
(- 17. 5,92)
(20,92) j 90 PERMIS$1BLE OPERATING l
j k (32,80)
REGI N
(-26,80);
80
.70 RESTRICTED:
RESTRICTED'
^
REGION I
E REGION
=.
60 i
E M
t 50 o
l 5
40 l
E I
,. 3 0 i
20 I
10 i
I l
i I
I I
I I
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l
-50
-40
-30
-20
-10 0
10 20 30 40
.50 l
l Axial Power Imbalance, %
8-17 Babcock 8 Wilcox
.- n
~
~
Figure 8-18.
Operational Power Imbalance Envelope for Operation From 200 10 to 350 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-3C) r l
l
\\
(-22,102)
(20,102) j 100 i.
(-22, 92 )
(20,92) j 90 a -
PERMISSIBLE OPERATING I
REGION
(-32,80)
- 80 (32,80) 70 RESTRICTED
. RESTRICTED lI I REGION REGION
=
60 l
EN 50 i
o D
40 E2 30 20 i
- 10 i
l I
I I
I I
I I
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i
-50
-40
-30
-20
-10 0
10 20 30 40 50 i
Axial Power Imbalance, % (
8-19 Babcock & Wilcox
Figure 8-19.
Operational Power Imbalance Envelope for Operation After 350 2 10 EFPD - ANO-1. Cycle 6 (Tech Spec Figure 3.5.2-3D) 9
(-24,102)
(18,102)
O 7
100 a
(-26,92) 90 PERMIS$1BLE!I OPERATING
(-30,80)
"'S* " 8 0 RESTRICTED O
RESTRICTED \\
, 70 REGION
\\
REGION g.
l E
60 w.
i
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(-32,50) 6 5.
50 i
=2.
, 40
}
I 30 t
1 20 10 t
I i
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-50
-40
-30
-20
-10 0
10 20 30 40 50 AxialPowerImbalance,%{
8-20 BabC0Ck & Wilcox t
.-....:.a Figure 8-20.
APSR Position Limits for Operation From 0 to 60 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-4A) 110 (8,102)
(31,102) 100 RESTRICTED' REGION j
(8,92) t (31,92) l (41,80) 80 t (0,80) l D
'70 E
l2 60 PERMISSIBLE OPERATING
'E REGION 50 j *,
(100,50)
!:ja 40 ta.
30 1
20 l
10 0
O 10 20 30 40 50 60 70 80 90 100 i
% Witnarawn i
8-21 Babcock & Wilcox
- - - - - ~
Figure 8-21.
APSR Position Limits for Operation From 50 to 200 t 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-4B) i
(
110 (8 102)
(38,,102)
RESTRICTED 100 I
REGION I
90
> (8,92) y,(38,92)
(0,80)
(41 80) l 00 <
70 l4 m
PERMISSIBLE; OPERATING i REGION i
=
60 n
(100,50) 50 40 i
1 l
l 30 20 10 n
i e
a a
a a
a i
y O
10 20 30 40 50 60 70 80 90 100
(
5.Witnarawn Babcock 3.Wilcox 8-22
_ _ _ __..a..
Figure 8-22.
APSR Position Limits for Operation From 200 2 10 to 350 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-4C) 110 (s.102)
(38,102) 100 RESTRICTEDi REGION i
4 I(38,92) 90 (8,92) 80 (
(0,80) 41,80) 70 PERMISSIBLE; g
OPERATING !
!==
60 l
REGION i
g E 50 (100,50)
\\
f
., 4 0 30 3
20 10 O
t i
a i
0 10 20 30
.40 50 SO 70 30 90 100 5.fitnarawn l
l l
l 8-23 Babcock & Wilcox
~ ^ ^ ~ =
Figure 8-23.
APSR Position Limits for Operation Af ter 350 10 EFPD - ANO-1, Cycle 6 (Tech Spec Figure 3.5.2-4D) 110 100 90 80 l
70 APSR INSERTION NOT ALLOWED E
IN THIS TIME INTERVAL 60 l
8 N
50 ow l
f 40 E
m 30 20 10 t
0 i
0 10 20 30 40 50 60 70 80 90 100 i
l APSR, % Witndrawn l
l l
l l
l 8-24 Babcock & Wilcox l
Using a local quality limit of 22% at the point of minimum DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the j quality at the exit is higher than the quality at the point of minimum DNBR.
The DNBR, as calculated by the BAW-2 correlation, continually increases from point of minimum DNBR so that the exit DNBR is always higher and is a function of the pressure.
The magnitude of the rod bow penalty applied to each fuel cycle is equal to or greater than the necessary burnup-dependent DNBR rod bow penalty for the applicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis. Pressure-temperature operating limits are cur-j.
rently based on an original method of calculating rod bowing penalties that j
3 are more conservative than those that would be obtained with new approved pro-cedures". For the current cycle of operation,'this subrogation results in a DNBR margin of 10%, which is partially used to offset the reduction in DNBR due to fuel rod bowing. The flux-flow limit is calculated with the DNBR, mar-gin necessary to offset the rod bow penalty calculated with the new pro-cedures.
The maximum thermal power for three-pump operation is 88.74% due to a power level trip produced by the flux-flow ratio (74.7% flow x 1.054 = 78.73% power) plus the maximum calibration and instrumentation error. The maximum thermal power for other reactor coolant pump conditions is produced in a similar man-ner.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22% for that particular reactor coolant pump situation. Curves 1 and 2 of Figure 2.1-3 are the most restrictive be-cause any pressure-temperature point above and to the left of this curve will be above and to the left of the other curve.
l References 1 Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, Babcock & Wilcox, Lynchburg, Virginia, May 1976.
2 FSAR, Section 3.2.3.1.1.c.
8-25 Babcock s,Wilcox
8 D. F. Ross and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller (NRC), Memorandum, " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing'on Thermal Margin Calculations for Light Water Reactors,"
December 8, 1976.
L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of In-terim Procedure for Calculating DNBR Reduction Due to Rod Bow," October 18, 1979.
I I
8-26 Babcock a,Wilcox
f I
9.
STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is outlined i
below. 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. Acceptable criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.46 seconds at the conditions above.
it should be noted that safety analysis calculations are based on a rod drop 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 instead of the two-thirds inserted position i
for data gathering.
l j9.2.
Zero Power Physics Tests I
- 9.2.1.
Critical Boron Concentration iCriticality is obtained by deboration at a constant dilution rate. Once f
! criticality is achieved, equilibrium boron is obtained and the critical boron concentration determined. The critical boron concentration is calculated by correcting for any rod withdrawal required to achieve equilibrium boron. The acceptance criterion placed on critical boron concentration is that the ac-i
9.2.2.
Temperature Reactivity Coefficient I
'The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limit. The Babcock & )Milcox 9_1
<s average coolant temperature is varied by first increasing and then decreasing i
temperature by 5'F.
During the change in temperature, reactivity feedback is compensated by discrete changes in rod motion. The change in reactivity is then calculated by the summation of reactivity (obtained from a reactivity Ac-calculator strip chart recorder) associated with the temperature change.
ceptance criteria state that the measured value shall not differ from the predicted value by more than 20.4 x 10~" (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 fuel-Doppler coefficient of reactivity is added to obtain moderator coefficient. This value must not be in excess of the acceptance criteria limit of +0.9 x 10~" (Ak/k)/*F.
9.2.3.
Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6, and 7) are measured at hot This technique con-zero power conditions using the boron / rod swap method.
sists of establishing a deboration rate in the reactor coolant system and l
s compensating for the reactivity changes of this deboration by inserting con-The reactivity changes that trol rod groups 7, 6, and 5 in incremental steps.
occur during these measurements are calculated based 6n Reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each of the The l
controlling groups are then summed to obtain integral rod group worths.
acceptance criteria for the control bank group worths are as follows:
I 1.
Individual bank 5, 6, 7 worth:
predicted value - measured value x 100 5 15 measured value i
2.
Sum of groups 5, 6, and 7:
predicted value - measured valu x 100
$ 10 measured vaine t
i 9-2 Babcock 3.Wilcox
9.2.4.
Ejected Control Rod Reactivity Worth After CRA groups 7, 6, and 5 have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth ob-tained by adding the incremental changes in reactivity by boration.
i After the ejected rod has been borated to 100% withdrawn and equilibrium l
l boron established, the ejected rod is swapped in versus the controlling rod group, and the worth is determined by the change in the control rod group po sition. Acceptance criteri_ for the ejected rod worth test are as follows:
lPredictedvalue-measuredvalu 100l520 1.
x measured value 2.
Measured value (error-adjusted) 5 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 440, 75, and 100% FP With Nominal Control Rod Position Core power distribution tests are performed at 40, 75, and 100% full power (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 index is established at a nominal full-power rod configuration at which the l core power distribution was calculated. APSR position is established to pro-vide a core power imbalance corresponding to the imbalance at which the core l
I power distribution calculations were performed. -
i The following acceptance criteria are placed on the 40% FP test:
1.
The worst-case maximum 1HR mus'. be less than the LOCA limit.
2.
The minimum DNBR must be greater than 1.30.
3.
The value obtained from extrapolation of the minimum DNBR to the acxt power plateau overpower trip setpoint must be greater than 1.30, or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.
i 9-3 Babcock 8.Wilcox
A 4.
The value obtained from 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 /
T l
flow trip envelope.
S.
The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.
6.
The highest measured and predicted radisi peaks shall be within the following limits:
l t
i predicted value - measured val,ug x 100 5 8.
l measured value I
7.
The highest measured and predicted total peaks shall be within the following limits:
j predicted value - measured valu x 100 5 12.
measured value Items 1, 2, 5, 6, and 7 are established to verify core nuclear and thermal calculational models, thet by verifying the acceptability of' data from these 1
models for input to safety evaluaticas.
Items 3 and 4 establish the criteria wbereby escalation to the next/ power plateau may be accomplished without exceeding the safety 1Laits specified by I
the safety analysis with regard to DNBR and linear heat rate.
The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test except thac core equilibrium xenon is established prior to'the 75 l
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 following limits:
U predicted value - measured value x 100 55 measured value I
2.
The highest measured and predicted total peaks shall be within the following limits:
predicted value - measured value x 100 57'.5 measured value 1
9-4 Y Dabcock 8.Wilcox x-
\\
n;3 g
L 3
_a _m....
t-9.3.2.
Incore Vs Excore Detector Imbalance Correlation Verification at 440% FP l
1 Imbalances, set up in the core by control rod positioning, are read simultan-eously on the incore detectors and excore power range detectors. The excore l
detector offset versus incore detector offset slope must be at least 1.15.
If this criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.
i 9.3.3.
Temperature Reactivity Coefficient
{'
at 4100% FP sk The overage reactor coolant temperature is decreased and then increased by i
about 5'F at constant reactor power. The reactivity associated with each temperature change is obtained from the change in the controlling rod group l
f position. Controlling rod group worth is measured by the fast insert / withdraw method. The temperature reactivity coeff1cient is calculated from the measured
. *\\
changes in reactivity and temperature',' Acceptance criteria state that the I
moderator temperature coef
- ~'t shall be negative.
\\
,s, 9.3.4.
Power Doppler Rem
,2ty I
r Coefficient at %100% FP l
Reactor power l's ' decreased and then increased by about 5% FP.
The reactivity
[
change is obtained from the change in controlling rod group position. Con-i x
l~~
trol rod group worth is mea grad using the fast insert / withdraw method.
Re-l
(
)
acti?ity corrections are made for changes in xenon and reactor coolant tem-perdure'thatoccurduringthemeasurement. The power Doppler reactivity con.ficient is calculated from the measured reactivity change, adjusted as j
i i
st.ated above, and the measured power change.
The predicted value of the j
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 10 (ak/k)/% FP.
9.4.
Procedure for Use if Acceptance 7
Criteria Not Met I
If the acceptance criteria for any test are not met, an evaluation is per-l i
formed before the test program, is continued. The results of all tests will be reviewed by the plant's nuclear engineering group.
If the acceptance criteria of the startgp physics tests are not met, an evaluation will be per-I formed by the plant's nuclear engineering group with assistance from general 4[
y 9-5 Babcock & Wilcox 1
Af
,, y
=_ _ __
i The If fice personnel, Middle South Services, and the fuel vendor, as needed.
results of this evaluation will be presen'.ed to the On-site Plant Safety Com-l If a safety Resolution will be required prior to power escalation.
mittee.
question is involved, the Off-site Safety Review Committee would review the situation, and the NRC would be notified if an unreviewed safety question
{
1 exists.
6 i
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4 1
REFERENCES 1 Arkansas Nuclear One, Unit 1 - Final Safety Analysis Report. Docket 50-313, Arkansas Power & Light.
2 T. A. Coleman and J. T. Willse. Extended Burnup Lead Test Assembly Irradi-ation Program, BAW-1626 Babcock & Wilcox, October 1980.
Arkansas Nuclear One, Unit 1 -- Cycle 5. Reload Report, BAW-1658, Rev. 2, 3
Babcock & Wilcox, May 1982.
i
" J. H. Taylor to S. A. Varga, Letter, "BPRA Retainer Reinsertion," January 14, 1980.
Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep 5
Collapse, BAW-10084, Rev. 1 Babcock & Wilcox, November 1976.
- f. H. Hsii, et al., TACO-2 -- Fuel Pin Performance Analysia, BAW-10141P, 8
Babcock & Wilcox, January 1979.
7 C. D. Morgan and H. S. Kao, TAFY -- Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, May 1972.
8 Arkansas Nuclear One Unit 1 -- Fuel Densification Report, BAW-1391, Babcock l
& Wilcox, June 1973.
' D. F. Ross and D. G. Eisenhut (NRC), Memorandum to D. S. Vassallo and K. R.
Goller (NRC) on " Interim Safety Evaluation Report on the Effects of Fuel I
Rod Bowing on Thermal Margin Calculations for Light Water Reactors,"
l December 8, 1976.
10 L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of In-I terim Procedure for Calculating DNER Reduction Due to Rod Bow," October 18, 1979.
ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev.1, Babcock II
& Wilcox, September 1975.
J. H. Taylor (B&W Licensing) to R. L. Baer (Reactor Safety Branch, USNRC),
12 Letter, July 8,1977.
I Babcock &)Milcox A-1
-.. ~j
. c -:.,,. : :... ~
- -2,..
~. _..
_=
=- --
l L
23 J. H. Taylor (B&W Licensing) to L. S. Rubenstein (USNRC), Letter September 5, 1980' 1
Note: All Babcock & Wilcox reports mentioned are from NPGD, Lynchburg, Virginia.
i i
1 1
1 a
1 t
4 i
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A-2 Babcock & Wilcox
_ _