ML20098F135
| ML20098F135 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/31/1984 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20098F132 | List: |
| References | |
| BAW-1840, NUDOCS 8410020365 | |
| Download: ML20098F135 (64) | |
Text
l BAW-1840 August 1984 ARKANSAS NUCLEAR ONE, UNIT 1
- Cycle 7 Reload Report -
l d
i BABCOCK & WILC0X Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 8410020365 840926 babcock &WWIICOM gDRADOCK-05000 a McDermott comoany
CONTENTS Page 1.
INTRODUCTION AND
SUMMARY
1-1 2.
OPERATING HISTORY 2-1 3.
GENERAL DESCRIPTION 3-1 4.
FUEL SYSTEM DESIGN.......................
4-1 4.1.
Fuel Assembly Mechanical Design 4-1 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 4.3.
Th e rmal De si gn......................
4-2 4-3 4.4.
Material Design 4.5.
Operati ng Experience...................
4-3 4.6.
Fuel Assembly Design Changes................
5.
NUCLEAR DESIGN..........................
5-1 5-1 5.1.
Physics Characteristics 5.2.
An al y ti c al I n p ut.....................
5-1 5-2 5.3.
Changes in Nuclear Design 6.
THERMAL-HYDRAULIC DESIGN....................
6-1 7.
ACCIDENT AND TRANSIENT ANALYSIS 7-1 7-1 7.1.
General Safety Analysis 7.2.
Accident Evaluation 7-2 8.
PROPOSED MODIFICATIONS TO TECHNICN, SPECIFICATIONS.......
8-1 9.
STARTUP PROGRAM - PHYSICS TESTING 9-1 9.1.
Precritical Tests 9-1 9-1 9.1.1.
Control Rod Trip Test 9.2.
Zero Power Physics Tests.................
9-1 9.2.1.
Critical Boron Concentration...........
9-1 9.2.2.
Temperature Reactivity Coefficient........
9-2 9.2.3.
Control Rod Group Reactivity Worth........
9-2 9.2.4.
Ejected Control Rod Reactivity Worth.......
9-3 9.3.
Power Escalation Tests..................
9-3
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Babcock &WHcom a McDermott company
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5' CONTENTS (Cont'd)
Page 9.3.1.
Core Power Distribution Verification at S40, 75, and 100% FP With Nominal Control Rod Position................
9-3 9.3.2.
Incore Vs Excore Detector Imbalance Correlation Verification at $40% FP 9-5 9.3.3.
Temperature Reactivity Coefficient at s100% FP.....................
9-5 9.3.4.
Power Doppler Reactivity Coefficient at s100% FP 9-5 9.4.
Procedure for Use if Acceptance Criteria Not Met.....
9-5 10.
REFERENCES...........................
10-1 List of Tables Table 4-1.
Fuel Design Parameters and Dimensions.............
4-5 4-6 4-2.
Fuel Thermal Analysis Parameters 4-3.
Fuel Assembly Design Changes..................
4-7 5-1.
Physics Parameters for ANO-1, Cycles 6 and 7..........
5-3 2.
Shutdown Margin Calculations for ANO-1, Cycle 7...........
5-5 6-1.
Maximum Design Conditions, Cycles 6 and 7............
6-3 1.
Comparison of FSAR and Cycle 7 Accident Doses..........
7-4 7-2.. Comparison of Key Parameters for Accident Analysis 7-5 7-3.
Bounding Values for Allowable LOCA Peak Linear Heat Rates....
7-5 List of Figures Figure 3-1.
Core Loading Diagram for ANO-1, Cycle 7
~
3-3 3-2.
Enrichment and Burnup Distribution, ANO-1 Cycle 7 Off 400 EF P D Cy cl e 6... '....................
3-4 3-3.
Control Rod Locations and Group Designations for ANO-1, Cycle 7'.
3-5 3-4.- LBP Enrichment and Distribution, ANO-1, Cycle 7 3-6 5-1.
ANO-1, Cycle 7, 80C Two-Dimensional Relative Power l-Distribution - Full Power Equilibrium Xenon, No rmal Rod Posi tions.....................
5-6 8-1.
Core Protection Safety Limits -- ANO-1, Cycle 7........
8-8 8-2.
Core Protection Safety Limits -- AN0-1, Cycle 7........
8-9 l
8-3.
Protective System Maximum Allowable Setpoints -- ANO-1, l
Cycle 7.......................... r 8-10 3
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Babcock &WHcom a McDermott company
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Figures (Cont'd)
Figure-Page F
8-4.
Boric Acid Addition Tank Volume and Concentration Vs RCS Average Temperature -- ANO-1, Cycle 7........
8-11 8-5.
Rod Position Limits for Four-Pump Operation from 0 to E0C EFPD -- ANO-1, Cycle 7 8-12 8-6.
-Rod Position Limits for Three-Pump Operation from 0 to E0C EFPD -- AN0-1, Cycle 7 8-13 8-7.
Rod Position Limits for Two-Pump Operation from 0 to E0C EFPD -- ANO-1, Cycle 7 8-14 8-8.
Operational -Power Imbalance Envelope for Operation From 0 to E0C -- ANO-1, Cycle 7 8-15 8-9.
APSR. Position Limits for Operation From 0 EFPD to 1
APSR Withdrawal -- ANO-1, Cycle 7............
8-16 8-10. APSR Position Limits for Operation After Withdrawal --
ANO-1, Cycle 7 8-17 8-11. LOCA Limited Maximum Allowable Linear Heat Rate.....
8-18 i
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-iV-Babcock &Wilcou
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1.
INTRODUCTION AND
SUMMARY
This report justifies the operation of the seventh cycle of Arkansas Nu-clear 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 7 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 6 and 7 reactor parameters related to power capability are sum-marized briefly in section 5 of this report.
All of the accidents ana-lyzed in the FSAR1 have been reviewed for cycle 7 operation.
In those cases where cycle 7 characteristics were conservative compared to those analyzed for previous cycles, no new accident analyses were perfonned.
The Technical Specifications have been reviewed, and the modifications re-quired for cycle 7 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 Emer-gency Core Cooling Systems, it has been concluded that ANO-1 can be operated safely for cycle 7 at a rated power level of 2568 MWt.
The cycle 7 core for ANO-1 will contain four twice-burned lead test as-semblies (LTAs).
These assemblies are part of a Department of Energy Ex-tended Burnup Test Program. The LTA design is described in reference 2.
I h k e & Micos 1_i a McDermott company
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- c 2.
OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Arkansas Nuclear One, Unit 1 is the. currently operating cycle 6.
This cycle 7 design is based on a design cycle 6 length of 400 effective full power days (EFPD).
No anomalies occurred during cycle 6 that woul d adversely affect fuel performance during cycle 7.
l 4
hM &Mm 2-1 a McDermott company
1 T
1 3.
GENERAL DESCRIPTION The - ANO-1 reactor core is described in cetail in section 3 of the Arkansas Nuclear One, Unit 1, Final Safety Analysis Report (FSAR).
The cycle 7 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 instrument guide tube.
The fuel is comprised of dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4.
The fuel assem-blies in all batches have an average nominal fuel loading of 463.6 kg of uranium, with the exception of four batch 7B LTAs, which have a nominal loading of 440.0 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 assemblies except the LTAs; the corresponding parameters for the LTAs are included in refer-ence 2.
Figur'e 3-1 is the fuel shuffle diagram for ANO-1, cycle 7.
The initial en-235, respec-richments of batches 7B, 8 and 9 are 2.95, 3.21, and 3.30 wt %
0 tively.
All the batch 6C assemblies and 31 of the twice-burned batch 7 assemblies will be discharged at the end of cycle 6.
The remaining 37 twice-burned batch 7 assemblies (designated batch 78) will be shuffled to new locations, with 12 on the core peripnery.
Sixty of the 72 once-burned batch 8 assemblies will be shuffled to new locations on or near the core pe-riphery.
The remaining 12 will surround the center assembly.
The 68 fresh batch 9 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 7.
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3-1 Babcock &WHcom a McDermott company
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F Reactivity is controlled by 61 f011-length Ag-In-Cd control rods, 64 burn-able poison rod assemblies (BPRAs), and soluble boron ~ him.
In addition s
to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution.
The cycle 7 locations of the 69 control rods and the group designations are indicated in Figure 3-3.
The core locations of the total pattern (69 con-trol rods) of cycle 7 are identical to those of the reference cycle indi-cated in the reload report for ANO.-1, cycle 6.3 There is a minor differ-ence in the group designations between cycle 7 and the reference cycle.
The cycle 7 locations and enrichments of the BPRAs are shown in Figure 3-4.
2.
i i
l-l 3-2 Bakock &Micos a McDermott company
Figure 3-1.
Core Loading Diagram for ANO-1, Cycle 7 FUEL TRANSFER CANAL 1
I LO3 M04 F
P!2 L13 A
8 8
9 8
8 A07 K02 M02 F
M08 F
M14 K14 A09 8
78 8
8 9
8 9
8 8
78 HIS F
NO3 F
K04 F
K12 F
N13 F
A08 C
78 9
8 9
8 9
8 9
8 9
78 G01 F
LO9 F
A06 F
NO2 F
A10 F
K06 F
GIS 0
- 78 9
8 9
78 9
78 9
78 9
8 9
m E
809 C12 F
K10 F
E07 F
E09 F
LO7 F
C04 807 8
8 9
8 9
78 9
78 9
8 9
8 8
C10 811 F
F01 F
C03 F
M06 F
C13 F
FIS F
8'05 C06 F
8 8
9 78 9
78 9
8 9
78 9
78 9
8 8
011 F
009 F
GOS F
M10 908 LOS F
Gil F
007 F
005 G
8 9
8 9
a 9
8 8
8 9
9 8
9 8
78 78 F
H11 F
P12 F
L11 H13 N14 H03 F05 F
804 F
N05 F
y
- M 9
8 9
78 9
8 78 8
9 78 9
8 9
8 m
N11 F
N09 F
K05 F
Fil C08 E06 F
K11 F
N07 F
N05 K
8 9
8 9
9 8
8 8
9 9
8 9
8 78 79 910 P11 F
L01 F
003 F
E10 F
913 F
L15 F
POS 906 L
8 8
9 78 9
18 9
8 9
78 9
78 9
8 8
P09 912 F
F09 F
M07 F
M09 F
G06 F
004 P07 M
8 8
9 8
9 78 9
78 9
8 9
8 8
K01
'F G10 F
'R06 F
014 F
RIO F
F07 F
K15 y
78 9
8 9
78 9
78 9
78 9
8 9
78 A08 F
003 F
G04 F
G12 F
013 F
H01 0
78 9
8 9
8 9
8 9
8 9
78 p
ROT G02 E02 F
E08 F
E14 G14 R09 78 8
8 9
8 9
8 8
73 F03 E04 F
E12 F13 R
8 8
9 8
8 I
Z 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 7
Cycle 6 Location Satch !D F = Fresh Fuel Assembly
- Marit 8E8 LTA 3-3 Babcock &WHcom a McDermott company
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Figure 3-2.
Enrichment and Burnup Distribution, ANO-1 Cycle 7 off 400 EFPD Cycle 6 8
9 10 11 12 13 14 15 2.95 3.21 3.21 3.30 2.95 3.30 3.21 3.30 H
22289 16047 16564 0
22324 0
16932 0
3.21 3.30 2.95 3.30 3.21 3.30 3.21 rs 16620 0
33065 0
16397 0
15861 2.95 3.30 2.95 3.30 3.21 3.21 L
17429 0
22291 0
10859 15624
.1 3.30 3.21 3.21 M
16268 0
12456 13552 3.21 3.30 2.95 16300 0
21495 i
2.95 0
23900 P
l R
i X.XX Initial Enrichment, wt % 2ssU XXXXX B0C Burnup, mwd /mtU l
3-4 Babcock &WHcom a McDermott company L
Figure 3-3.
Control Rod Locations and Group Designations for AN0-1, Cycle 7
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A B
4 7
4 C
2 6
6 2
7 8
5 8
7 D
2 5
1 1
5 2
E 4
8 3
7 3
8 4
p G
6 1
3 3
1 6
~ ~- Y H
W-7 5
7 2
7 5
7 K
6 1
3 3
1 6
4 8
3 7
3 8
4 L
2 5
1 1
5 2
M N
7 8
5 8
7 l
2 6
6 2
C l
4 7
4 p
j R
i l
1 i
Z
", 1 2
3 4
5 6
7 8
9 10 11 12 ; 13 14 15 X
Group Number Group No. of Rods Function 1
8 Safety 2
9 Safety 3
8 Safety 4
8 Safety 5
8 Control 6
8 Control 7
12 Control 8
8 APSRs 3-5 Babcock &WHeou a McDermott comparty
Figure 3-4.
LBP Enrichment and Distribution, ANO-1, Cycle 7 8
9 10 11 12 13 14 15 H
1,4 1,4 K
1.4 1.1 0.8 l
1.4 1.4 0.8
. M 1.4 1.4 1,4 N
1.1 1.4 0.0 0
1.4 0.8 0.0 P
0.8 R
X.X LBP Concentration, wt % BSC in Al203 3-6 Babcock &Wilcon a McDermott company L
4.
FUEL SYSTEM DESIGN 4.1.
Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel parameters for ANO-1 cycle 7 are listed in Table 4-1.
All fuel assemblies are identical in concept and are mechanically interchangeable.
Retainer assemblies will be used on two fuel ' assemblies that contain the regenerative neutron sources, and on sixty-four fuel assemblies that contain BPRAs.
Sixty-two of the retainers will be exposed for a fourth cycle of irradiation during cycle 7.
This ad-ditional cycle of irradiation is justified in reference 4 based on examina-tion of retainers which have undergone three cycles of irradiation.
The re-suits of the examination meet criteria developed earlier in terms of wear and holddown force.
These criteria ensure that the retainers will perfonn in a safe and adequate manner in the areas of holddown force, stress, and fatigue during a fourth cycle of in-reactor use.
These criteria were de-veloped from analyses similar to those done in the original justification of the design and use of the retainer assemblies in references 5 and 6.
Four of the FAs in the highest burnup batch 78 are LTAs.
A description and evaluation of the LTAs is found in reference 2.
4.2.
Fuel Rod Design There has been a change in the pellet design for batch 9 fuel rods.
The fuel pellet length / diameter (L/D) ratio has been decreased from 1.63 to 1.18.
This change in L/D ratio will not adversely affect fuel perfonnance, l
and at high burnups it is expected to decrease local cladding strains.
The results of the mechanical evaluations of the fuel rods are discussed below.
4.2.1.
Cladding Collapse The batch 7 fuel is more limiting than batches 8 and 9 because of its pre-l vious incore exposure time.
The batch 7 assembly power histories were 1
i 4-1 Babcock & WNces A MCDermott (0mparty
analyzed to detennine the most limiting three-cycle power history for creep collapse.
This worst-case power history was then compared against a generic analysis to ensure that creep-ovalization will not affect fuel performance during ANO-1 cycle 7.
The generic analysis was performed based on reference 7 and is applicable for the batch 7 fuel design.
The creep collapse analysis predicts a collapse time greater than 35,000 effective full-power hours (EFPH), which is longer than the maximum ex-pected residence time of 30,394 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% on cladding plastic tensile circumferential strain.
The pellet is designed to ensure that cladding plastic strain is less than 1% at riasign local pellet burnup and heat gen-eration rate.
The design burnup ana heat generation rate are higher than the worst-case values 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 7 core is thermally similar.
The design of the four batch 7B lead test assemblies is such that the thermal performance of this 4-2 Babcock &WHcom a McDermott company
f e
fuel is equivalent to or slightly better than the standard Mark B design used in the remainder of the core.
All thermal design analyses for cycle 7 fuel used the TAC 02 code, as described in reference 8, for fuel temperature and fuel rod internal pressure prediction.
The results of the thermal design evaluation of the cycle 7 core are sum-marized in Table 4-2.
Cycle 7 core protection limits were based on a linear heat rate (LHR) to centerline fuel melt of 20.5 kW/f t as detennined by the TAC 02 code.
The LHR to melt of the LTA fuel is greater than 20.5 kw/ft.
The maximum fuel assembly burnup at E0C 7 is predicted to be less than 44,400 mwd /mtU for the Mark B fuel and less than 45,700 mwd /mtU for the LTA fuel.
The fuel rod internal pressures have been evaluated with TAC 02 for the highest burnup fuel rods and are predicted to be less than the nominal reactor coolant pressure of 2200 psia.
In the cycle 6 reload report (reference 3), the batch 7 and batch 8 fuel parameters of Table 4-2 were listed in a fashion compatible with the model-ing assumptions of the TAFY-3 code (reference 9).
In that report the pel-let diameter, stack height, and nominal linear heat rate were provided in Table 4-2 based on the assumption of instantaneous fuel densification.
The TACO 2 code, on the other hand, utilizes a time dependent fuel densification model.
With the implementation of the TAC 02 code for cycle 7 evaluations, the Table 4-2 parameters are provided based on nominal dimensions.
4.4.
Material Design
.'he chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 9 fuel assemblies is identical to that of the present fuel.
4.5.
Operating Experience Babcock & Wilcox operating experience with the Mark B 15x15 fuel assembly has verified the adequacy of its design.
As of April 30, 1984, the follow-ing experience has been accumulated for eight 88W 177 fuel assembly plants using the Mark B fuel assembly:
4-3 Babcock & WIfcom a McDermott company
)
Cumu tv net Max FA burnup,(a) mwd /mtU Current Reactor cycle Incore Discharged
- output, MWh Oconee 1 8
34,499 50,598 48,808,138 i
. Oconee 2 7
27,035 36,800 43,444,856 Oconee 3 8
35,123 35,463 45,200,486 Three Mile Island 5
25,200 32,400 23,840,053 Arkansas Nuclear.
. One, Unit 1 6
31,450 36,540 38,872,852 Rancho Seco 6
30,500 38,268 33,923,457 Crystal. River 3 5
23,17" 29,900 27,083,428 Davis-Besse 4
28,5 32,790 19,237,628 (a)As of April 30, 1984.
(b)As of January 31, 1984.
4.6.
Fuel Assembly Design Changes A complete list of fuel related design changes are identified in Table 4-3.
These changes will not adversely affect fuel performance.
4-4 mock &WWilcon a MtDermott company
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Table 4-1.
Fuel Design Parameters and Dimensions Batch 7B Batch 8 Batch 9 Fuel assembly type Mart B4, Mark B4 Mark B4 Mark BEB No. of assemblies 33 Mark B, 72 68 4 Mark BEB Fuel rod OD (nom), in.
0.430 0.430 0.430 Fuel rod ID (nom), in.
0.377 0.377 0.377 Flexible spacers Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Undensified active fuel 141.8 141.8 141.8 length (nom), in.
Fuel pellet 00 (mean 0.3686 0.3686 0.3686 specified), in.
Fuel pellet initial 95.0 95.0 95,0 density (nom), % TD Initi 2.95 3.21 3.30 wt % gl fuel enrichment,
'35g Average burnup, BOC, 23,992 14,910 0
mwd /mtU Cladding collapse
>35,000
>35,000
>35,000 time, EFPH Estimated residence 30,394 19,680 10,080 time, EFPH i
4-5 Babcock & WHeos a McDermott company
Table 4-2.
Fuel Thermal Analysis Parameters Batch 7B Batch 8 Batch 9 No. of assemblies 33(a) 72 68 Initial density, % TD 95.0 95.0 95.0 Initial pellet 00, in.
0.3686 0.3686 0.3686 Initial stack height, in.
141.80 141.80 141.80 Nominal linear hegt rate at 5.74 5.74 5.74 2568 MWt, kW/ft(bi TAC 02-Based Predictions Average fuel temperature at
, 1400 1400 1400 nominal LHR, F a
Minimum LHR to melt, kW/f t 20.5 20.5 20.5 (a)Four LTAs were also analyzed; the results of which are reported in reference 2.
(b) Based on a nominal stack height.
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l 4-6 Babcock &WHeem a McDermott company
Table 4-3 i
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Fuel Assembly Design Changes I
Cycle-6 Cycle-7 Part Number Part Number Description of Change Guide Tube Assy. P/N-510*
1135026-001 Improved manufacturing process; hole removed from lower end plug i
and holes in lower guide tube wall increased to maintain same G/T tower End Plug P/N 511*
1138974-001 internal flow rate.
l G/T Upper Nut P/N 103*
1135026-001 Holddown Spring P/N 563 1135021-001 Improved B10 holddown spring design maoe of Inconel 718.
Holddown Spider P/N 553 Holddown Spring Ret. Mach P/M 1134885-002 fuel Pellet P/M 1004892-001 1134918-001 GE fuel pellets with L/D ratio I
change from 1.63 to 1.18.
Spacer Sleeve A P/N 517 1i35980-001 Part number change only.
B P/N 518 1135980-002 Part number change only.
C P/N 519 1135980-003 Part number change only.
BPRA Assy. P/M 970 1125783-001 Ball locking feature in coupling spider was eliminated.
l BP Rod P/M 641 1125784-001 Short stack LBP configuration.
CRA P/N 600 1142078-001 Longer life CRA; the cladding material changed from stainless steel to Inconel absorber is slightly longer with no change in total poison mass.
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- Thirty-two of the cycle 7 fuel assemblies used this type of GT assembly, plug and nut.
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5.
NUCLEAR DESIGN 5.1.
Physics Characteristics Table 5-1 lists the core physics parameters of design cycles 6 and 7.
The values for both cycles were generated using PDQ07.
Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 7 at full power with equilibrium xenon and nominal rod positions.
Differences in cycle length, feed enrichment, BPRA load;ng, and shuffle pat-tern make it difficult to compare the physics parameters of cycles 6 and 7.
Calculated ejected rod worths and their adherence to criteria are consid-ered at all times in life and at all power levels in the development of the rod position limits presented in section 8.
The maximum stuck rod worth for cycle 7 is greater than that for the design cycle 6 at BOC and APSR a
pull, but less at E0C.
All safety criteria associated with these worths are met.
The adequacy of the shutdown margin with cycle 7 stuck rod worths is demonstrated in Table 5-2.
The following conservatisms were applied for the shutdown calculations:
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-lated using a two-dimensional model.
The reference fuel cycle shutdown mar-gin is presented in the ANO-1 cycle 6 reload report.3 5.2.
Analytical Input The cycle 7 incore measurement calculation constants to be used for comput.
ing core power distributions were prepared in the same manner as those for the reference cycle.
5-1 Babeeck &WHees a McDermott tompany
i 5.3.
Changes in Nuclear Design Core design changes for cycle 7 are the transition to a very low leakage (VLL) design and the use of "short-stack" LBPs.
For this transition cycle, twelve twice-burned assemblies are located on the core periphery to reduce fluence levels on the reactor vessel.
The LBP used in cycle 7 has a 9-inch shorter poison stack than that used with the ' standard Mark B design, i.e.,117 versus 126 inches of A10 -8 C.
The top 9 inches of the 23 4 poison stack are replaced by a Zircaloy tubular spacer.
This LBP design results in only a slight mass reduction versus the standard design, and does not change the dynamic characteristics of the LBP.
The "short-stack" design asymmetrically positions the burnable poison stack relative to the fuel column and alters the core axial power shape to create increased
" effective maneuvering room" at the beginning of the cycle.
As in cycle 6, the APSRs will be withdrawn near the end of cycle 7 The calculated stability index at 404 EFPD without APSRs is -0.052 h*1 which demonstrates the axial stability of the core.
The calculational methods used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle.
The operating limits (Technical Specification changes) for the reload cycle are given in section 8.
5-2 Md AMb8 a McDermotI company
Table 5-1.
Physics Parameters for ANO-1, Cycles 6 and 7(a)
Cycle 6(b)
Cycle 7(c)
Cycle length, EFPD 387 420 Cycle burnup, mwd /mtU 12,128 13,158 Avg. core burnup, E0C, mwd /mtU 23,009 24,238 Initial core loading, mtU 82.0 82.0 Critical boron - BOC, ppm (No Xe)
HZPLd), group 8 ins 1463 1578 HFP, group 8 ins 1273 1346 Critical boron - EOC, ppm HZP, group 8 out, no Xe 704 696 HFP, group 8 out, eq Xe 95 83 Control rod worths - HFP, BOC, % ak/k Group 6 1.13 1.20 Group 7 1.36 1.65 Group 8 0.42 0.39 Control rod worths - HFP, EOC, % ak/k Group 7 1.40 1.53 Max ejected rod worth - HZP, % ak/k(e)
BOC (N-12), group 8 ins 0.53 0.69 400 EFPD (N-12), group 8 ins 0.46 0.50 EOC (M-11), group 8 out 0.47 0.52 Max stuck rod worth - HZP, % ak/k BOC (N-12), group 8 ins 1.50 1.71 400 EFPD -(H-14), group 8 ins 1.63 1.73 EOC (H-14), group 8 out 1.43 1.29 Power deficit, HFP to HZP, % ak/k BOC 1.68 1.60 E0C 2.38 2.35 Doppler coeff - HFP,10-5 (ak/k *F)
B0C (no Xe)
-1.54
-1.53 E0C (eq Xe)
-1.82
-1.80 Moderator coef f - HFP,10-4 ( Ak/k *F)
BOC, (no Xe, crit ppm, group 8 ins)
-0.84
-0.69 EOC, (eq Xe, 0 ppm, group 8 out)
-2.89
-2.79 5-3 m a gggges a McDermott comparty
Table 5-1.
(Cont'd)
Cycle 6(b)
Cycle 7(c)
Boron worth - HFP, pps/% ak/k BOC 123 129 E0C 109 109 Xenon worth - HFP, % ak/k BOC (4 EFPD) 2.57 2.55 E0C (equilibrium) 2.69 2.68 Effective delayed neutron fraction - HFP B0C 0.0063 0.0063 E0C 0.0053 0.0052 (a) Cycle 7 data ere for the conditions stated in this report.
The cycle 6 core conditions are identified in reference 3.
(b) Based on 455 EFPD at 2568 MWt, cycle 5; actual cycle length was 446.4 EFPD.
(c) Based on 400 EFPD at 2568 MWt, cycle 6, which is the actual cycle length expected.
(d)HZP denotes hot zero power (532F Tavg), HFP denotes hot full power (579F I
Tavg *
(*) Ejected rod worth for groups 5 through 7 inserted, group 8 as stated.
I m
5-4 Babcoeft &WHcom a MtOttmott company
Table 5-2.
Shutdown Margin Calculations fo' ANO-1, Cycle 7 r
% ak/k
% ak/k
% ak/k Available Rod Worth Total rod worth, HZP 9.04 9.44 9.14 Worth reduction due to
-0.10
-0.10
-0.10 poison material burnup Maximum stuck rod, HZP
-1.71
-1.73
-1.29 Net worth 7.23 7.61 7.75 Less 10% uncertainty
-0.72
-0.76
-0.78 6.97 Total available worth 6.51 6.85 Required Rod Worth Power deficit, HFP to HZP 1.60 2.35 2.35 Allowable inserted rod 0.50 0.60 0.65 worth Flux redistribution 0.75 1.20 1.20 Total required worth 2.85 4.15 4.20 Shutdown margin (total 3.66 2.70 2.77 available worth minus total required worth)
Notet The required shutdown margin is 1.00% ak/k.
5-5 h M8ces a McDermott company
Figure S-1.
ANO-1 Cycle 7,BOC(4EFPD)s Two-Dimensional Relative Power Distribution - Full Power F
Equilibrium Xenon, Normal Rod Positions 8
9-10 11 12 13 14 15 H
1.05 1;24 1.23 1.17 1.01 1.26 1.10 0.83 a
T K
1.26 1.26
. 0.86 1.19 1.21 1.13 0.60
~
8 y
L 1.09 1.18.
0.91 1.24 0.94 0.42 M
1.22 1.25 1.14 0.68 1
,r, j
N 1.20 1.06 0.37 1
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Group'No.
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i 6.
THERMAL-HYDRAULIC DESIGN The fresh batch 9 fuel is hydraulically and geometrically similar to the previously irradiated batches 78 and 8 fuel.
The four batch 7B LTAs have been analyzed to ensure that they are never the limiting assemblies during cycle 7 operation.
The results of the thermal-hydraulic analysis for the LTAs are provided in reference 2.
The thermal-hydraulic design evaluation supporting cycle 7 marks the first implementation for ANO-1 of crossflow modeling with the LYNXT codes (refer-ences 10-12) for DNB predictions.
The crossflow modeling methods and appli-cations are described in reference 13.
A notable difference in the cycle 7 modeling is the use of a 1.71 design radial-local (FyH) power peak with a 1.65 (P8) symmetric chopped cosine design axial flux shape.
This is in comparison with the 1.71 radial-local and 1.5 axial flux shape used in cycle 6.
The cycle 7 design peaking re-suits in an allowable increase of the total peak from the cycle 6 value of 2.57 to 2.83.
The selection of the cycle 7 peaking was based on the desire to increase flexibility in the determination of operating limits (i.e., rod insertion limits).
Note that this change in design peaking has no impact on the results of BAW-182913 since that report presents the crossflow model development and justification, and not the plant specific analyses.
The thermal-hydraulic design conditions for cycles 6 and 7 are s'umarized in Table 6-1.
This table quantifies the DNB improvement for the transition to crossflow modeling with the associated design peaking for cycle 7.
The reactor protection system (RPS) setpoints for the DNB-based variable low pressure trip will remain the same for cycle 7.
DNB margin improvement gained with crossflow modeling has resulted in supporting an increase of the flux / flow setpoint up to 1.08 for cycle 7, 6-1 Babcock &WHess a McDermotI company u
t Previous fuel ' cycle evaluations included ~ the calculation of a rod bow penal-ty for each batch based on the highest fuel burnup in that batch.
A rod bow topical report (reference 14), which addresses the mechanisms and re-sulting local conditions of rod bow, has been submitted to and approved by the NRC.
The topical report concludes that rod bow penalty is insignifi-cant and is offset by the reduction in power production capability of the fuel assemblies with irradiation.'
Therefore, no departure from nucleate boiling ratio (DNBR) reduction due to rod bow need be considered for cycle 7.
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Table 6-1.
Maximum Design Conditions, Cycles 6 and 7 Cycle 6 Cycle 7 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 DNBR modeling Closed-channel Crossflow Reference design radial-local 1.71 1.71 power peaking factor Reference design axial flux shape 1.5 cosine 1.65 cosine Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Active fuel length, in.
140.7(a) 141.8 175(a) 174 Avg heat flug at 100% power, 104 Btu /h-ft 450(a) 492 Max heat flug at 100% power, 103 Btu /h-ft CHF correlation B&W-2 B&W-2 Minimum DNBR At 112% power 2.05 2.08 At 100% power 2.39 2.43 (a) Based on densified length.
6-3 Babcock &WIIcom a McDermott ccmpany
7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cycle 7 parameters to determine the effect of the cycle 7 reload and to en-sure that thennal performance during hypothetical transients is not de-graded.
The effects of fuel densification on the FSAR accident results have been evaluated and are reported in reference 15.
Since batch 9 reload fuel as-semblies contain fuel rods whose theoretical density is higher than those considered in the reference 15 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-evaluation is that, even though the FSAR dose analyses used a con-servative basis for the amount of plutonium fissioning in the core, im-provements in fuel management techniques have increased the amount of energy produced by fissioning plutonium.
Since plutonium-239 has different fission yields than uranium-235, the mixture of fission product nuclides in the core changes slightly as the plutonium-239 to uranium-235 fission ratio
- changes, i.e., plutonium fissions produce more of some nuclides and less of other nuclides.
Since the radiological doses associated with each accident are impacted to a dif ferent 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 7.
The bases used in the dose calculation are identical to those presented in the FSAR except for the following three differences:
7-1 Babcock &WHcom l
a McDermott company
1.
The fission yields and half-lives used in the new calculations are based on more current data.
2.
Updated (lowered) whole body gamma dose conversion factors.
- 3. _ The steam generator tube rupture accident evaluation considers the in-creased amount of steam released to the environment via the main steam i
relief and atmospheric dump valves because of the slower depressuriza-f tion due to the reduced heat transfer rate caused by tripping of the reactor coolant pumps upon actuation of the high pressure injection (a l
post-TMI-2 modification).
l A comparison of the radiological doses presented in the FSAR with those cal-l culated specifically for cycle 7 (Table 7-1) show that some doses are slightly higher and some are slightly lower than the FSAR values.
- However, l
with the exception of the maximum hypothetical accident (MHA) all doses are I
bounded by the values represented 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 RiA the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the exclusion area L
boundary (EAB) is 157.7 Rem (53% of the 10 CFR 100 limit) and the 30 day i
thyroid dose at the low population zone (LPZ) is 73.1 Rem (24% of the 10 CFR 100 limit).
The small increases in some doses are essentially offset by reductions in other doses.
Thus, the radiological impact of accidents during _ cycle 7 is not significantly different than that described in Chapter 14 of the FSAR.
1 7.2.
Accident Evaluation l
The key parameters that have the greatest effect on detenntning the outcome of a transient can typically be classified in three major areas:
core ther-mal parameters, thermal-hydraulic parameters, and kinetics parameters, in-cluding the reactivity feedback coefficients and control rod worths.
l Core thermal properties used in the FSAR accident analysis were design op-erating values based on calculational values plus uncertainties.
Thermal parameters for fuel batches are 7B, 8, and 9 given in Table 4-2.
The cycle
- 7. thermal-hydraulic maximum design conditions are compared with the previ-ous cycle 6 values in Table 6-1.
These parameters are common to all the ac-cidents considared in this report.
The key kinetics parameters from the FSAR and cycle 7 are compared in Table 7-2 7-2 liBatacoc8r &WIlcos l
a McDermott company b
'The implementation of crossflow modeling (see section 6) for DNB analyses has identified addition'al DNB margin over that of closed channel modeling used in previous cycle analyses.
This additional margin has been incorpo-rated into the DNB-based core protective safety limits for cycle 7.
The flux / flow protective system setpoint, which is established by the core DNBR performance during the limiting Condition II transient (two RC pump coast-down), has now increased to 1.08 %FP/% flow for cycle 7 as a result of cross-flow modeling.
In addition to the gain in the flux / flow setpoint, the mini-mum DNBR during the limiting transient has increased by over 15 DNB points (where 1 DNB point = 0.01).
A generic LOCA analysis for a B&W 177-FA, lowered-loop NSS has, been per-formed using the Final Acceptance Criteria ECCS Evaluation Model,(reported in BAW-10103).16 This analysis is generic since the limiting val'ues of key parameters for all plants in this category were used.
Furthermore, the com-bination 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 reported in BAW-10103 and substantiated by reference 17 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 7 fuel.
These LHR limits include the effects of NUREG 0630 with offsetting credit taken for FLECSET.
It is concluded from the examination of cycle 7 core thermal and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the AN0-1 plant's ability to operate safely during cycle 7.
Considering the previously accepted design basis I
used in the FSAR and subsequent cycles, the transient evaluation of cycle 7 l
1s considered to be bounded by previously accepted analyses.
The initial conditions for the transients in cycle 7 are bounded by the FSAR, the fuel densification report, and/or subsequent cycle analyses.
l 7-3 Babcock & WHcom a McDermott company l
Table 7-1.
Comparison of FSAR and Cycle 7 Accident Doses _
FSAR doses, Cycle 7 doses, Rem Rem Fuel Handling Accident Thyroid dose at EAB (2 h) 0.92 1.24 Whole body dose at EAB (2 h) 0.54 0.21 Steam Line Break Thyroid dose at EAB (2 h) 1.'6 1.71 0.008 Whole body dose at EAB (2 h)
Steam Generator Tube Failure Thyroid dose at EAB (2 h) 0.0087 6.15 Whole body dose at EAB (2 h) 0.16 0.52 Waste Gas Tank Rupture Thyroid dose at EAB (2 h) 0.22 0.054 Whole body dose at EAB (2 h) 1.53 Control Rod Ejection Accident j
Thyroid dose at EAB (2 h) 11.4 3.42 Whole body dose at EAB (2 h) 0.014 0.003 Thyroid dose at LPZ (30 d) 8.3 2.55 Whole body dose at LPZ (30 d) 0.0099 0.002 LOCA i
Thyroid dose at EAB (2 h) 3.6 4.10 Whole body dose at EAB (2 h) 0.057 0.026 Thyroid dose at LPZ (30 d) 1.66 1.02 Whole body dose at LPZ (30 d) 0.043 0.008 i
Maximum Hypothetical Accident Thyroid dose at EAB (2 h) 153 157.7 Whole body dose at EAB (2 h) 10 4.73 Thyroid dose at LPZ (30 d) 64.1 73.1 Whole body dose at LPZ (30 d) 3.4 1.54 7-4 hock &MIcom a MCDermott CompJny
)
Table 7-2.
Comparison of Key Parameters for Accident Analysis FSAR and densification ANO-1 Parameter report value cycle 7 Doppler coeff (B0C),10-5 Ak/k/*F
-1.17
-1.53 Doppler coeff (E0C),10-5 ak/k/*F
-1.30
-1.80 Moderator coeff (B0C),10-4 Ak/k/'F 0.0(a)
-0.69 Moderator coeff (E0C),10-4 Ak/k/*F
-4.0(b)
-2.79 All-rod group worth (HZP), % ak/k 12.9 9.04 Initial boron concentration, ppm 1150 1346 Boron reactivity worth (HFP),
100 129 ppm /% Ak/k Max ejected rod worth (HFP), % ak/k 0.65 0.39 Dropped rod worth (HFP), % ak/k 0.65 0.20 (a)+0,5 x 10-4 Ak/k/*F was used for the moderator dilution analysis.
(b)-3.0 x 10-4 ak/k/*F was used for the steam line failure analysis.
Table 7-3.
Bounding Values for Allowable LOCA Peak Linear Heat Rates Allowable Allowable Core peak LHR, peak LHR, elevation, first 1000 mwd /mtu, balance of cycle, ft kW/ft kW/ft 2
14.0 15.5 4
16.6 16.6 6
17.5 18.0 8
17.0 17.0 10 16.0 16.0 7-5 mock &Wilcon a McDermott company
,. -..., ~ -.... _. -... _. _..... - - -. _ _ _ _
8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 7 operation to ac-count for changes in power peaking and control rod worths.
These changes are a result of the ve'y. low leakage fuel cycle design and the implementa-r tion of crossflow in the analysis.
The LOCA limits used to develop the normal operating Technical Specifications inc,1ude the impact of NUREG 0630 with offsetting credit taken for FLECSET.
Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be ex-ceeded, nor will the thermal design criteria be violated.
The following pages contain the revisions to previous Technical Specifications.
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8-1 a McDermott company
.. ~.
DNBR of 1.3 corresponds to a 95 percent probability at a 95 percent confi-dence level that DNB will nct oce"";. this is considered a conservative mar-gin to DNB for all cperating conditions.
The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.
The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure was actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR greater than 1.3 is predicted.
The curve is the most restric-tive combination of 3 and 4 pump curves, and is based upon the maximum possible thermal power at 106.5% design flow per applicable pump status.
This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects:
F" = 2.83 Fyg=1.71; F$=1.65 The curves of F';ure 2.1-2 are based on the more restrictive of two thennal limits and include the effects of potential fuel densification:
1.
Tge 1.3 DNBR limit produced by a nuclear power peaking factor of Fq = 2.83 or the combination of the radial peak, axial peak, and the l position of the axial peak that yields no less than 1.3 DNBR.
2.
The combination of radial and axial peaks that prevents central fuel melting at the hot spot.
The limit is 20.5 kW/ft.
l Power peaking is not a directly observable quantity, and therefore, limits have been established on the basis of the reactor power imbalance produced by the power peak ing.
The flow rates for curves 1, 2 and 3 of Figure 2.1-3 correspond to the ex-pected minimum flow rates with four pumps, three pumps, and one pump in j
each loop, respectively.
8 8-2 Babcock & Wilcou a McDermott company
The curve of Figure ?.1-1 is the most restrictive of all possible reactor coolant pump maximum thermal power combinations shown in Figure 2.1-3.
The curves of Figure 2.1-3 represent the conditions at which a minimum DNBR greater than 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operatf or..
The local quality at the point of minimum DNBR is less than 22 percent.(1)
Using a local quality limit of 22 percent at the point of minimum DNBR as a basis for curve 3 of Figure 2,1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.
The DNBR as calculated by the BAW-2 correlation conti,nually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.
The maximum thennal power, as a function of. reactor coolant pump operation is limited by the power level trip produced by the flux-flow ratio - ( percent flow X flux-flow ratio), plus the appropriate calibration and instrumentation errors.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a UNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particu-lar reactor coolant pump situation.
Curves 1 and 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.
REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May 1976.
(2) FSAR, Section 3.2.3.1.1.c.
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l 8-3 Babcock &Wilcox a McDermott cotr,9any
The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor pow-er level increases or the reactor coolant flow rate decreases.
The power level trip setpoint produced by the power-to-flow ratio provides overpower
- DNB protection for all modes of pump operation.
For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate.
Typical power level and low flow rate
. combinations for the pump situations of Table 2.3-1 are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if pow-er is 101 percent and reactor flow rate is 100 percent or flow rate is 93.5 percent and power level is 100 percent.
2.
Trip would occur when three reactor coolant pumps are operating if pow-er is 80 percent and reactor flow rate is 74.7 percent or flow rate is 70 percent and-power level is 75 percent.
3.
Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52 percent and reactor flow r5te is 49.2 percent or flow rate is 45.8 percent and the power level is 49.0 percent.
The flux / flow ratios account for the maximum calibration and instrumenta-tion errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a i
conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum cali-bration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded.
These thermal limits are either power peaking kW/ft limits or DNBR limits.
The reactor power imbalance (power in top half of core minus power in bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced.
The power-to-flow ratio reduces the power level trip associated with reactor power-to-reactor power imbalance boun-daries by LD7 percent for a 1 percent flow reduction.
4 B.
Pump Monitors In conjunction with the power imbalance / flow trip, the pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s).
The pump monitors also restrict the power level for the number of pumps in operation.
C.
RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high-pressure trip setpoint is reached before the nuclear overpower trip setpoint.
The trip setting limit 12 8-4 Babcock &WHcom a McDermott compacy
Reactor Pmtection Systan Trip Setting Limits (Specifications Td)le 2.3-1)
Four RC pmps operating Thr& RC pmps operating One RC pmp operating in (noMnal operating (nominal operating each loop (noninal Shutdown pmer,1000) peer,75%)
operating pmer, 49%)
typass Nuclear power, % of 104.9 104.9 101.9 5.0a rated, max Nucigar power based on 1.07' times flm mims 1.07 times flm mirus 1.07 times flm minus Bypassed l
fla/' and inbalance, %
mduction die to im-ruiuction die to im-ruiuction die to im-of rated, max balance (s) balance (s) balance (s)
Nuclear power based on NA tA 55 Bypassed punp nonitors, % of rated, nexc High RC systan presstre, 2300 2300 2300 17203 9'
psig, max Lw RC systen pressure, 1800 1800 1800 Bypassed i
psig, min Variable lw RC systen 11.75 Tout - 51031 11.75 Tout - 51031 11.75 Tout - 5103d Bypassed 1
pressum, psig, min l
RC taip, F, nax 618 618 618 618 i
g liigh reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 psia)
=m presstre, psig, max EI ER aAutonatically set den other serjents of the RPS (as specified) are bypassed.
!h bRextor coolant systen fim.
CIhe pmp monitors also prodre a trip on (a) loss of two RC ptsps in one RC log), and (b) loss of ore or two RC pmps during j,:::
tw operation.
- y d
i Iout s given in &grees Fahrerheit (F).
i
~
6.
If a control - rod in the regulating or axial power shaping groups is declared inoperable per Specificatit,.. 4. '.1.2 opera-
. tion above 60 percent of the thennal power allowable for the reactor coolant pump combination may continue provided the rods in the group. are positioned such that the rod that was declared inoperable is contained within allowable group aver-
' age position limits of Specification 4.7.1.2 and the with-orawal limits of Specification 3.5.2.5.3.
3.5.2.3 The. worth of _ single - inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Con-trol 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 3.1% power N I
shall be redu'ced immediately to below the power Ievel cutoff U
(92% FP), Moreover, the power level cutoff value shall be reduced 2% for each 1% tilt in excess of 3.1%.
For less than 4 pump operation, thermal power shall be reduced 2% of the thermal power allowable for the reactor coolant pump combination for each 1% tilt in excess of 3.1%.
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 reduced to less than 3.1% except for ph'ysics tests, or the
~
following adjustnients in setpoints and limits shall be made:
a.
The protection system maximum allowable setpoints (Figure i
2.3-2) shall be reduced 2% in power for each 1% tilt.
b.. The control rod group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of 3.1%.
i c.
The operational imbalance limits shall be reduced 2% in power for each 1% tilt in excess of 3.1%.
i l-3.
If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testirig, 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 combination 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.
47 8-6 mgayWIIcom N
4 A4CDermott Company
.9 3.5.2.5 Control rod positions:
l'.
Technical Specification 3.1.3.5 (safety rod withdrawal) does not1 prohibit. the exercisirg of individual safety. rods as - re-quired by Table 4.1-2 or _ apply to inoperable safety rod limits in Technical Specification 3.5.2.2.
2.
Operating rod group overlap shall be 207,
+5 between two sequential groups, except for physics tests.
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a McDermott Company
Figure 8-1.
Core Protection Safety Liraits -- AM0-1 (Tech Spec Figure 2.1-2)
Thermal Power evel, % FP
- - 120
(-3,1,112)
(27.7,112)
(-42,102,1 1 ACCEPTABLE l
l 4 PD!P l
l l OPERATIO::
-- 100 I
l 1(-31,90.6)
(27.7, 90.6k
- (35.9, 96.4)
(-42,89.7)I l ACCEPTABLE I
I i-4.c o 3 I
I l
PL?!P I
I OPERATIO!:S l
(35.9,75) g
[(-31,63.3)
(27.7. 63.3)I.
I I
i I
l ACCEPTABLE
-- 60 g
(-42,53.4 l
4, 3 A::D 2 i
i I
[ PD!P l
1 l OPERAE10:'
l 35.9,47.7) 1 l
l
- 40 I
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- 20 I
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-60
-40
-20 0
20 40 60 Reactor Power Imbalance, %
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Figure 8-2.
Core Protection Safety Limits -- ANO-1,
( Tec h R')ec rigure 2.1-3) 2600 2400
.?
E I
2200 5
V 2000
/
cm 5
3 1800 1600 560 580 600 620 640 660 Reactor Outlet Temper.ature, 'F CURVE GPM POWER PUUPS OPERATING (TYPE OF LlMIT) 1 374,880 (1005)*
112%
FOUR PUMPS (ONER LIMIT) 2 280,035 (74.75) 90.6%
THREE PUMPS (DNBR LIMtT) 3 184,441 (49.25) 64.1%
ONE PUNT IN EACH LOOP (OUALITY LIMIT)
- 106. 55 0F DESI GN F! 01 8-9 9
e en-
--'"'-T'
" ' ' " " ' ' ~ ~
Figure 8-3.
Protective System Maximum A!'owable Setpoints --
A!;0-1, (Tech Spec Figure 2.3-2)
Thermal Power Level, % FP
- 120
(-12,_107)
_ (12, 107)
{
l
-100l i
l ACCEI TABLE l
(-28,91T I 4 PU: P
[
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l OPERiTION I l
(22, 34.13) l(-12,8s)
-80 (12,330) 1 ACCEI TABLE I l
l
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l 3 ANI 4
l 8
I PWT l
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(-28, 64) i lOPERITION l l
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- 60 l 1
1 l
l (22, 57.13)
I(-12,59 1(12,852) l I
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(-28, 36) I ACCEPTABLE __ 40 [
l l
2, 31, AND 4
I g
I PUMPl I
(22, 29.13)
I OPERhTION l
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- 20 l l
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-60
-40
-20 0
20 40 60 l
l Reactor Power Imbalance, %
l l
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B.abcock & Wilcox g.jQ a McDermott ccr..pany L-
o Figure 8-4.
Boric Acid Addition Tank Volume and Concentration Vs RCS Average Temperature -- ANO-1 (Tech Spec Figure 3.2-1)
OPERATION ABOVE AND TO THE LEFT 6000 0F THE CURVES IS ACCEPTABLE 8700 PPM 5000 J
5 4000 9500 PPM E
- e c
3000 10,000 PPM O
B
,E o
2000 - 12,000 P M a<
i
/
m 1000 0
f i
200 300 400 500 600 700 RCS Average Temperature, F TEMP. F REQUIRED VOLUME, GAL.
~
8700 PPM 9500 PPM 10,000 PPM 12,000 PPM 579 6436 5863 5554 4589 l
532 5289 4817 4564 3769 500 4488 4087 3872 3199 400 2434 2218 2101 1737 300 986 898 851 705 200 0
0 0
0 i
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8-11 Babcock &Wilcox a venermott remony
Figure 8-5 Rod Position Limits for 4 -Pump Operation From 0 - EOC ------ AN0-1 (Tech Spec Figure 3.5.2-1 )
110 100 (231.2,102)
(280' (300,102) 90 (275,90) 80 (265,78) y 70 SHUTDOWN MARGIN LIMIT e
y 60 i
OPERATION RESTRICTED
)
0
- OPERATION IN (158.6,48)
(230,45)
THIS REGION IS 40 t
- NOT ALLOWE0 8
30 PEPMISSIBLE 20 OPERATING REGION (80,10 10 i(0, )
e i
0 f
+
0 20 40 60 80 100 120 140 160 180 200 220 240 250 280 300 0
20 40 60 80 100 i
f I
f f
I GROUP 7 0
20 40 60 80 100 i
f f
f f
f GROUP 6 0
20 40 60 80 100 L
f f
f f
r GROUP S Rod Index. % WD 8-12 Babcock & WIIcom a ueO. en.tt e,mm..anv
Ff gure 8-6 Rod Position Limits for 3 -Pemp Operation
\\
From 0 - E0C AN0-1 (Tech Spec Figure 3.5.242 )
110 100 90 80 (233,77)
(280.3,77), (300,77) g 70 SHUTDOWN MARGIN LIMIT (270,67 m
g 60 (260,58) 50
- OPERATION IN THIS REGION IS OPERATION 40 b
- NOT ALLOWEC RESTRICTED 5
(158.5,36) 30 (230,35)
~
PERMISSIBLE (31.5,9 OPERATING 10
~
REGION 0 i-0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0
20 40 60 80 100 f
f f
f f
f GROUP 7 0
20 40 60 80 100 t
t f
f f
I GROUP 6 0
20 40 6,0 8,0 1,00 t
t-t GROUP 5 Rod Index % WD 8-13 m.; -
Sabcock & Wilcox s udnwrwm w
o Figure 8 Rod Position Limits for 2,-Pump Operation From 0 EFPD to E00 - ANO-1 (Tecti Spec Figure 3.5.2-3 )
110 100 90 80
.y 70 60 50
(
,52)
W.W OPERATION IN 6 (300, 52)
THIS REGION Is SHUTDOWN OPERATION (275, 46) 40 g
- NOT ALLOWED MARGIN RESTRICTED (265,43) g LIMIT 30
~
(158.5,26) 20 (235,25)
OPERATING 10 REGION (0,0)
(81 5,.8 5),
3
'd 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300
?
2,0 4,0 69 89 170 GROUP 7 0
20 40 60 80 100 t
r f
f f
f GROUP 6 0
2,0 4,0 60 80 100 GROUP 5 Rod index, % WD
\\
8-14 Babcock A Wifenw
Figure 8-8 Operatio...I Power Imbalar.ce Envelope for Operation From 0 to E0C EFP0 -- ANO-1, (Tech Spec Figure 3.5.2 4 )
110
(-15,102) ()
Q (9, 102) 1gg
(-15,92) gg)(10,92)
(-20,80) 8Q) (10, 80) 70 RESTRICTE0 RESTRICTED REGION REGION 60 5
(-20, 50()
[-
50() (12, 50) 8 N
t-40
==
30 2
20 10
(-20,0)
(12,0)
O i
en
-50
_-40
-30
-20
-10 0
10 "
20 30 40 30 Axial Power Imbalance, %
8-15 Babcock & Wilcox
. um e r
i Figure 8-9 APSR Positiorv Limits for Operat... Fr.-
0 EFPD to APSR Withdrawal --- At10-1 (Tech Spec Figure 3.5.2-5A) 110 (9.5,102)
(35,102) f>
cp RESTRICTED 100 REGION 90 (p(35,90) 80 4
7 (40,75) 3 70 (
(0,70) 8 0
o PERMISSIBLE fa, OPERATIfiG b
50 p
40
>(100,40) 30 20 10 I
I I
8 I
O 0
10 20 30 40 50 60 70 80 90 100
% Withdrawn,
8-16 Babcock & Wilcox
Figure 8-10.
APSR Position Limits for Operation After APSP. Withdrawal------ ---- ANO-1 (Tech Spec Figure 3.5.2-SB )
110 100 90 80 APSR IftSERTION NOT ALLOWED 70 y
IN THIS TIME INTERVAL
=c 60 g
a 50 C
ig 40 30 20 10 0
0 10 20 30 40 50 60 70 80
' 90 100
% Withdrawn 8-17 Babcock & Wilcox a McDermott company
~. -..
Figure 8-11.
LOCA Limited Maximum Allowable Linear Heat Rate (Tech Spec Figurt 3. F.' " '.
d 21 x
20 aI 19 c:
ag 18
- =
Balance of h
17 g
Cycle 3
16 3
"g 15
~
First 1,000 S
mwd /mtU 14 q
e i
13 T
12 0
2 4
6 8
10 12 Axial Location of Feak Power from Bottom of Core, ft i
1 8-18 Babcock &Wilcox a McDermott company
_.. _ ~
3.
Except for physics tests or exercising control rods, (a) the con-trol rod withdrawal limits are specified on Figures 3.5.2-),
3.5.2-2 and 3.5.2-3 for 4, 3 and 2 pump operation respectively; and (b) the axial power shaping control rod withdrawal limits are specified on Figures 3.5.2-5A and 3.5.2-58.
If any of these control rod position limits are exceeded, corrective measures shall be taken immediately 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 />.
4.
Except for physics tests, power shall not be increased above the power level cut-off of 92% of the maximum allowable power level unless one of the following conditions is satisfied:
a.
Xenon reactivity is within 10% of the equilibrium value for operation at the maximum allowable power level and asymptot-ically approaching stability.
b.
Except for xenon free startup, when 3.5.2.5.4a applies, the reactor has operated within a range of 87 to 92% of the maxi-mum allowable power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to ex-ceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 40% rated power.
Except for physics tests, imbalance shall be maintained within the envelope defined by Figure 3.5.2-4.
If the imbalance is not within the envelope defined by Figure 3.5.2-4, corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalar,ce 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 reduced unt.1 imbalance limits are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with
. limited access to be authorized by the shrift supervisor.
Bases The power-imbalance envelope defined in Figure 3.5.2-4 is based on (1) LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-6) such that the maximum cladding temperature will not exceed the,F,inal Accept-ance Criteria and (2) the Protective System Maximum Allowable Setpoints (Figure 2.3-2).
Corrective measures will be taken immediately 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 improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbal-ance) must be at their limits while 8-19 Babcock &Wilcox a McDermott company
The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the definition of quadrant power tilt given in Technical Specifications Section 1.6.
These limits, in conjunction with the control rod position limits in Specification 3.5.2.5.3, ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.
The quadrant tilt and axial imbalance limits in Specification 3.5.2.4 and 3.5.2.6, respectively, apply when 'using the plant computer to monitor the limits. The 2-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service. Additional uncertainty is applied to the limits when other monitoring methods are used.
During the physics testing program, the high flux trip setpoints are administratively set as follows to ensure that an additional safety margin is provided.
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Babcock & Wiscom 8-20 a skoennocr company
1 9.
STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is out-lined below.
These tests verify that core perfonnance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of the unit.
9.1.
Precritic.
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.
Accep-l' table critieria 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 dron ' from fully withdrawn to two-thirds inserted.
Since the most accurate position indication is obtained from the zone reference swi tch at the l
75%-inserted position, this position is used instead of the two-thirds in-serted position for data gathering.
I 9.2.
Zero Power Physics Tests 9.2.1.
Critical Boron Concentration Criticality is obtained by deboration at a constant oilution rate.
Once criticality is achieved, equilibrium boron is obtained and the critical baron concentration determined. The critical boron concentration is calcu-lated by correcting for any rod withdrawal required to achieve equilibrium boron.
The acceptance criterion placed on critical boron concentration is that. the actual boron concentration must be within 1100 ppm boron of the predicted value.
l 9-1 Batscocir&WWIIcom A MCDermott Company e
l 9.2.2.
Temperature Reactivity Coefficient Th9 isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limit.
The average coolant temperature is varied by first increasing and then de-creasing temperature by 5'F.
During the change in temperature, reactivity l
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 calculator strip chart recorder) associated with the tempera-ture change.
Acceptance criteria state that the measured value shall not differ from the predicted value by more than i0.4 x 10-4 (Ak/k)/*F (pre-dicted value obtained from Physics Test Manual curves).
The mcderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement.
After the temperature coefficient l
I has been measured, a predicted value of fuel Doppler coefficient of reac-l tivity is added to obtain moderator coefficient.
This value must not be in f'
excess of the acceptance criteria limit of +0.5 x 10-4 (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 zero power conditions using the boron / rod swap method.
This technique I
consists of establishing a deboration rate in the reactor coolant system and compensating for the reactivity changes of this deboration by inserting control rod groups 7, 6 and 5 in incremental steps.
The reactivity changes that occur during these measurements are calculated based on 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 controlling groups are then summed to obtain integral rod group worths.
The acceptance criteria for the control bank group worths l
- are as follows:
l l
1.
Individual bank 5, 6, 7 worth:
predicted value - measured value x 100 -< 15 measured value 2.
Sum of groups 5, 6 and 7:
predicted value - measured value x 100 < 10 measured value 9-2 Babcock &WWilcon a McDermott company k
9.2.4.
Ejected Control Rod Reactivity Worth Af ter CRA groups 7, 6 and 5 have been positioned near the minimum rod insertion limit, the ejected rod is borated to 100% withdrawn and the worth obtained by adding the incremental changes in reactivity by bora-tion.
After the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is swapped in versus the controling rod group, and the worth is detennined by the change in the control rod group position.
Acceptance criteria for the ejected rod worth test are as fol-lows:
1.
predicted value - measured value x 100
-< 20 measured value 2.
Measured value (error-adjusted) < 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 S40, 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 pl ateau.
Rod index is established at a nominal full-power rod configura-tion at which the core power distribution was calculated.
APSR position l
is established to provide a core power imbalance corresponding to the im-l balance at which the core power distribution calculations were performed.
f The following acceptance criteria are placed on the 40% FP test:
i 1.
The worst-case maximum LHR must 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 next power plateau overpower trip setpoint must be greater than 1.30, or the extrapolated value of imbalance must fall outside the RPS power /im-balance / flow trip envelope.
9-3 Batacock &WHcom a McDermott company
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 out-side the RPS power / imbalance / flow trip envelope.
5.
The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.
6.
The highest measured and predicted radial peaks shall be within the following limits:
predicted value - measured value measured value
-< 8.
x 100 7.
The highest measured and predicted total peaks shall be within the fol-lowing limits:
predicted value - measured value measured value
-< 12.
x 100 Items 1, 2, 5, 6 and 7 are established to verify core nuclear and thermal calculational models, thereby verifying the acceptabili ty of data from these models for input to safety evaluations.
Items 3 and 4 establish the criteria whereby escalation to the next power plateau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR and linear heat rate.
The power distribution tests perfomed at 75 and 1005 FP are identical to the 40% FP test except that core equilibrium xenon is established prior to the 75 and 100% FP tests.
Accordi ngly, 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:
predicted value - measured value x 100 <5 mease ed value 2.
The highest measured and predicted total peaks shall be within the following limits:
predicted value - measured value measured value
-< 7.5 x 100 9-4 a McDermott company
9.3.2.
Incore Vs Excore Detector Imbalance Correlation Verification at s40% FP Imbalances, set up in the core by control rod positioning, are read simul-taneously on the incore detectors and excore power range detectors.
The excore 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.
9.3.3.
Temperature Reactivity Coefficient at S100% FP The average reactor coolant temperature is decreased and then increased by about 5'F at constant reacy.or po'etc.
The reactivity associated with each temperature change is CXa{ned from the change in the controlling rod group position.
Controlling rod group worth is measured by the fast in-sert / withdraw method.
The temperature reactivity coefficient is calcu-lated from the measured changes in reactivity and temperature.
Acceptance criteria state that the moderator temperature coefficient shall be nega-tive.
9.3.4.
Power Doppler Reactivity Coefficient at m100% FP Reactor power is decreased and then increased by about 5% FP.
The reactiv-ity change is obtained from the change in controlling rod group position.
Control rod group worth is measured using the fast insert / withdraw method.
Reactivity corrections are made for changes in xenon and reactor coolant temperature that occur during the measurement.
The power Doppler reactiv-ity coefficient is calculated from the measured reactivity change, ad-justed as stated above, and the measured power change.
The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual.
Acceptance criteria state that the measured value shall be more negative than -0.55 x 10-4 (A k/k)/%FP.
9.4.
Procedure for Use if Acceptance Criteria Not Met If the acceptance criteria for any test a~re not met, an evaluation is per-fonned before the test program is continued.
The results of all tests 9-5 Babcock &Wilcom a McDermott company
will be reviewed by the plant's nuclear engineering group.
If the accep-tance criteria of the startup physics tests are not inet, an evaluation will be perfonned by the plant's nuclear engineering group with assistance from general office personnel, Middle South Services, and the fuel vendor, as needed.
The results of this evaluation will be presented to the On-site Plant Safety Committee. Resolution will be required prior to power i
escalation.
If a safety question is involved, the Off-site Safety Review Committee would review the situation, and the NRC would be notified if an unreviewed safety question exists.
l 1
i l
l l
4 9-6 N EMb8 a McDermott company
- 10. 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 Irradiation Program, BAW-1626, Babcock & Wilcox, Lynchburg, Virginia, October 1980.
3.
Arkansas Nuclear One Unit 1, Cycle 6 Reload Report, BAW-1747, Babcock
& Wilcox, Lynchburg, Virginia, November 1982.
4.
J. H. Taylor (B&W) to J. F. Stolz (NRC), Letter, " Extension of Re-tainer Lifetime to Four Cycles," July 24, 1984.
5.
BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.
4 6.
J. H. Taylor (B&W) to S. A. Varga (NRC), Letter, "BPRA Retainer Rein-sertion," January 14, 1980.
7.
Program to Determine In-Pactor Performance of B&W Fuels -- Cladding Creep Collapse, BAW-10084A, Rev.
2, Babcock & Wilcox, Lynchburg, Virginia, October 1978.
8.
Y. H. Hsii, et al., TAC 02-Fuel Pin Performance Analysis, BAW-10141P-A, Rev.1, Babcock & Wilcox, Lynchburg, Virginia, June 1983.
9.
C. D. Morgan and H. S. Kao, TAFY -- Fuel Pin Temperature and Gas Pres-sure - Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1972.
10.
B. R. Hao and J. M. Alcorn, LYNX 1:
Reactor Fuel Assembly Thermal Hy-draulic Analysis
- Code, BAW-10129, Babcock
- Wilcox, Lynchburg, Virginia, October 1976.
11.
LYNX 2:
Subchannel Thermal-Hydraulic Analysis
- Program, BAW-10130, Babcock & Wilcox, Lynchburg, Virginia, October 1976.
10-1 Babcock &Wilcom a McDermott company
o 12.
J.
H. Jones, et al.,
LYNXT -- Core Transient Thermal-Hydraulic Pro-gram, BAW-10156, Babcock & Wilcox, Lynchburg, Virginia, February 1984.
13.
R. L. Harne and J. H. Jones, Thennal-Hydraulic Crossflow Applications, BAW-1829, Babcock & Wilcox, Lynchburg, Virginia, May 1984.
- 14. Fuel Rod Bowing in Babcock & Wilcox Fuel Designs, BAW-10147P-A, Rev.
1, Babcock & Wilcox, Lynchburg, Virginia, May 1983.
- 15. Arkansas Nuclear One, Unit 1-Fuel Densification Report, BAW-1391, Babcock & Wilcox, Lynchburg, Virginia, June 1973.
1, Babcock & Wilcox, Lynchburg, Virginia, September 1975.
17.
J.
H. Taylor (B&W Licensing) to R.
L. Baer (Reactor Safety Branch, USNRC), Letter, July 8,1977.
10-2 Babcock & WI8com a McDermott company
-)
l BAW-1829 April 1984 Thermal-Hydraulic Crossflow Applications k
Babcock &Wilcox a McDermott company
_ _ _ _ _ _ _ _