ML20063G891
| ML20063G891 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/31/1982 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20063G882 | List: |
| References | |
| BAW-1658, BAW-1658-R02, BAW-1658-R2, NUDOCS 8207290240 | |
| Download: ML20063G891 (29) | |
Text
.
BAW-1658. Rev. 2 l
Stay 1982 l
1 t
s ARKANSAS NUCLEAR ONE, UNIT 1
- Cycle 5 Reload Report -
4 l
Babcock &Wilcox 8207290240 820715 PDR ADOCK 05000313 p
p.
. ~.... _. _ _ _
4 BAW-1658, Rsv. 2
'May 1982 I
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ARKANSAS NUCLEAR ONE, UNIT 1 1
4
- Cycle 5 Reload Report -
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BABCOCK & WILCOX Nuclear Power Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox j
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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 4.2.
Fue Rod Design.
4-1 4.2.1.
Cladding Collapse 4-1 4.2.2.
Cladding Stress 4-2 4.2.3.
Cladding Strain 4-2
. 4................
4.3.
Thermal Design.
4-2 4.4.
Material Design 4-3 4.5.
Operating Experience.
4-3 5.
NUCLEAR DESIGN.
5-1 i
5.1.
Physics Characteristics 5-1 5.2.
Analytical Input 5-1 5.3.
Changes in Nuclear Design 5-2 6.
THERMAL-HYDRAULIC DESIGN.
6-1 7.
ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.
GeneraI Safety Analysis 7-1 7.2.
Accident Evaluation 7-1 8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS.
8-1 9.
STARTUP PROGRAM -- PHYSICS TESTING 9-1 9.1.
Precritical Tests 9-1 i
9.1.1.
Control Rod Trip Test 9-1 9.2.
Zero Power Physics Tests.
9-2 9.2.1.
Critical Boron Concentration.
9-2 9.2.2.
Temperature Reactivity Coefficient 9-2 9.2.3.
Control Rod Group Reactivity Worth.........
9-2 9.2.4 Ej ected Control Rod Reactivity Worth........
9-3 l
t Babcock & )Milcox t
A
Rsvision 1 (4/15/81)
CONTENTS (Cont 'd)-
Page 9.3.
Power Escalation Tests..
9-3 9.3.1.
Core Power Distribution Verification at %40, 75, and 100% FP With Nominal Control Rod Position.....
9-3 9.3.2.
Incore Vs Excore Detector Imbalance Correlation Verification at 440% FP 9-5 9.3.3.
Temperature Reactivity Coefficient at
%100% FP.
9-5 9.3.4.
P(ver Doppler Reactivity Coefficient at
%100% FP.
9-5 9.4.
Procedure for Use if Acceptance criteria Not Met 9-6 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 4 and 5..........
5-2 5-2.
Shutdown Margin Calculations for ANO-1, Cycle 5 5-4 6-1.
Maximum Design Conditions, Cycles 4 and 5 6-2 7-1.
Bounding Values for Allowable LOCA Peak Linear Heat Rates 7-2 7-2.
Comparison of Key Parameters for Accident Analysis..
7-3 8-1.
Reactor Protection System Trip Setting Limits 8-17 List of Figures Figure 3-1.
Fuel Shuffle for ANO-1 Cycle 5 3-3 3-2.
Enrichment and Burnup Distribution, ANO-1 Cycle 5 Off 329 EEPD Cycle 4 3-4 l1 3-3.
Control Locations and Group Designations for ANO-1 Cycle 5 3-5 3-4 LBP Enrichment and Distribution, ANO-1 Cycle 5 3-6 5-1.
ANO-1' Cycle 5, BOC Two-Dimensional Relative Power Distribution - Full Power Equilibrium Xenon, Normal Rod Positions.
5-5 8-1.
Core Protection Safety Limits.
8-18 8-2.
Core Protection Safety Limits.
8-19 8-3.
Core Protection Safety Limits.
8-20 8-4 Protective System Maximum Allowable Setpoints.
8-21
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Babcock & \\Nilcox
/
-- ~
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t R;vi:Sen 2 (5/15/82) g Figures (Cont'd)
Figure Page 8-5.
Protective System Maximum Allowable Setpoints.
8-22 8-6.
Boric Acid Addition Tank Volume and Requirements Vs RCS Average Temperature.
8-23 8-7.
Rod Position Limits for Four-Pump Operation From 0 t o 60 E FPD -- ANO-1, Cy cle 5..................
8-24 8-8.
Rod Position Limits for Four-Pump Operation From 50 to 200 1 10 EFPD -- ANO-1, Cycle 5...............
8-25 8-9.
Rod Position Limits for Four-Pump Operation From 200 : 10 to 400 10 EFPD -- ANO-1, Cycle 5 8-26 8-10.
Rod Position Limits for Four-Pump Operation From 400 ! 10 to 455 : 10 EFPD -- ANO-1, Cycle 5 8-27l2 8-11.
Rod Position Limits for Three-Pump Operation From 0 t o 6 0 E FPD -- ANO-1, Cy c l e 5.................
8-28 8-12.
Rod Position Limits for Three-Pump Operation From 50 to 200 ! 10 EFPD -- ANO-1, Cycle 5 8-29 8-13.
Rod Position Limits for Three-Pump Operation From 200 1 10 to 400 10 EFPD -- ANO-1, Cycle 5 8-30 8-14.
Rod Position Limits for Three-Pump Operation From 400 10 to 455 2 10 EFPD -- ANO-1, Cycle 5 8-31 l2 8-15.
Rod Position Limits for Two-Pump Operation From 0 to 60 EFPD -- ANO-1, Cycle 5.
8-32 8-16.
Rod Position Limits for Two-Pump Operation From 50 to 200 1 10 EFPD -- ANO-1, Cycle 5 8-33 8-17.
Rod Position Limits for Two-Pump Operation From 200 10 to 400 1 10 EFPD -- ANO-1, Cycle 5 8-34 8-18.
Rod Position Limits for Two-Pump Operation From 400 10 to 455 10 EFPD -- ANO-1, Cycle 5 8-35l2 8-19.
Operational Power Imbalance Envelope for Operation From 0 to 60 EFPD -- ANO-1, Cycle 5 8-36 8-20.
Operational Power Imbalance Envelope for Operation From 50 to 200 t 10 EFPD -- ANO-1, Cycle 5...
8-37 8-21.
Operational Power Imbalance Envelope for Operation From 200 10 to 400 t 10 ETPD -- ANO-1, Cycle 5........
8-38 8-22.
Operational Power Imbalance Envelope for Operation From 400 : 10 to 455 t 10 EFPD -- ANO-1, Cycle 5.......
8-39l2 8-23.
APSR Position Limits for Operation From 0 to 60 EFPD -- ANO-1, Cycle 5...
8-40 8-24.
APSR Position Limits for Operation From 50 to 200 10 EFPD -- ANO-1, Cycle 5 8-41 8-25.
APSR Position Limits for Operation From 200 : 10 to 400 10 EFPD -- ANO-1, Cycle 5..
8-42 8-26.
APSR Position Limits for Operation From 400 10 to 455 t 10 EFPD -- ANO-1, Cycle 5..
8-43l2 8-27.
LOCA Limited Maximum Allowable Linear Heat Rate.
8-44
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Babcock & Wilcox
9 4.
FUEL SYSTEM DESIGN 4.1.
Fuel Assembly Mechanical Design The type of fuel assemblies and pertinent fuel design parameters for ANO-1 cycle 5 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 4 reload reort are applicable to Mark B4 assemblies.
In addition to the assemblies listed, four lead test assemblies (LTAs) are being inserted with batch 7.
One stan-dard Mark B fuel assembly will contain annealed guide tubes to compare with the LTAs.
The analysis and justification for the LIAs and annealed 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 second cycle of operation for the retainer assemblies.
The justification for the de-sign and use of the retainers for two cycles is described in reference 4, and is applicable to ANO-1, cycle 5.
Similar retainer assemblies will be used on the two fuel assemblies containing the regenerative neutron sources.
4.2.
Fuel Rod Design The batch 7 internal fuel rod design differs from batches 5 and 6 in several respects. As outlined in Table 4-1, these include an increase in initial pellet density from 94 to 95% TD, a decrease in the nominal fuel pellet diame-ter f rom 0.3695 to 0.3686 inch, and a reduction in stack length from 142.25 to 141.8 inches.
These combined changes were implemented to improve fuel perfor-mance as well as maintain a constant assembly uranium loading.
The mechanical evaluation of the fuel rod is discussed below.
4.2.1.
Cladding Collapse The batch 5 fuel is more limiting than batches 6 and 7 because of its previous incore exposure time.
The batch 5 assembly power histories were analyzed to determine the most limiting three-cycle power history for creep collapse.
4-1 Babcock & Wilcox
g o.
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This worst-case power history"was then compared against a generic analysis to ensure that creep-ovalization will not aff ect fuel performance during ANO-1 cycle 5.
The generic analysis was performed based on reference 5 and is ap-plicable for the batch 5 fuel design.
The creep collapse analyses predicts a collapse time greater than 35,000 ef-fective full-power hours (EFPH), which is longer than the maximum expected residence time of 30,288 EFPH (Table 4-1).
l2 4.2.2.
Cladding Stress The ANO-1 stress parameters for batch 4 and subsequent fuel 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 unir-radiated yield strength, and all stresses must be less than the minimum speci-fled 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 i
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 design burnup and 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 5 core is thermally sinilar. 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 in the remainder of the fuel.
The thermal design analysis of the LTAs using 8
the TACO-2 code is described in reference 2.
4-2 Babcock & Wilcox
s The results af the thermal design evaluation of the cycle 5 core are summarized in Table 4-2.
Cycle 5 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 maximum fuel rod burnup at EOC 5 is predicted to be less than 42,000 mwd /mtU.
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 7 fuel assemblies is identical to that of the present fuel.
4.5.
Operating Experience Babcock & Wilcox operating experience with the Mark B, 15 x 15 fuel assembly has verified the adequacy of its design.
As of July 31, 1980, the following experience has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly:
Max FA burnup *,
Cumulative net Current electrical output,
Reactor cycle Incore Discharged MWh Oconee 1 6
23,300 40,000 32,457,943 Oconee 2 5
26,100 33,700 27,786,436 Oconee 3 5
30,200 29,400 28,483,452 TMI-l 4
32,400 32,200 23,840,053 ANO-1 4
28,100 33,222 25,006,003 Rancho Seco 4
27,900 37,730 22,625,102 Crystal River 3 3
20,530 23,194 12,113,632 Davis-Besse 1 1
14,884 7,654,365 Babcock & VVilcox 4-3
Table 4-1.
Fuel Design Parameters and Dimensions Batch 5 Batch 6 Batch 7 Fuel assembly type Mark B4 Mark B4 Mark B4, Mark BEB No. of assemblies (*}
49 60 64 Mark B4, 4 l1 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 Er-4 Zr-4 Zr-4 Undensified active fuel 142.25 142.25 141.80 length (nom), in.
Fuel pellet OD (mean 0.3695 0.3695 0.3686 specified), in.
Fuel pellet initial 94.0 94.0 95.0 density (nom), % TD Initial fuel enrichment, 3.01 3.19 2.95 235 wt %
U Average burnup, BOC, mwd /mtU 16,467 12,892 0
Cladding collapse time,
>35,000
>35,000 35,000 EFPH Estimated residence time, 25,560 28,680 30,288 l2 EFPH (max)
(
Four lead test assemblies (Mark BEB) make up a total batch 7 reload of 68 fuel assemblies.
These LTAs were analyzed and reported in reference 2.
l 4-4 Babcock & Wilcox
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5.
NUCLEAR DESIGN g
- ?
- E 5.1.
Physics Characteristics Table 5-1 lists the core physics parameters of design cycles 4 and 5.
The 1
values for both cycles were generated using PDQ07.
Since the core has not
}
yet reached an equilibrium cycle, differences in core physics parameters are w
to be expected between cycles.
Figure 5-1 illustrates a representative rela-tive power distribution for the beginning of cycle 5 at full power with equi-librium xenon and nominal rod positions.
Operational changes as well as differences in cycle length, feed enrichment, d
BPRA loading, shuffle pattern, and rod group designations make it difficult to compare the physics parameters of cycles 4 and 5.
Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits
.:a presented in section 8.
The maximum stuck rod worth for cycle 5 is less than that for the design cycle 4 at BOC and EOC. All safety criteria associated with
]
these worths are met.
The adequacy of the shutdown margin with cycle 5 stuck S
rod worths is demonstrated in Table 5-2.
The following conservatisms were 1
applied for the shutdown calculations:
bd 1.
Poison material depletion allowance.
2.
10% uncertainty on net rod worth.
3 i.d 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 4 reload report.
)3 5.2.
Analytical Input The cycle 5 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the g
reference cycle.
5-1 Babcock & Wilcox w
/
R:vicion 2 (5/15/d2) 5.3.
Changes in Nuclear D sign 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 are two significant operational changes f rom the reference cycle: the full insertion of the APSRs during the last 55 EFPD of cycle 5 and a change from a rodded to a feed-and-bleed mode of operation.
The stability and control of the core in the feed-and-bleed mode with APSRs removed have been analyzed.
The calculated stabiiity index without APSRs is -0.0334 h~,
which demonstrates the axial stability of the core.
The operating limits (Tech-nical Specification changes) for the reload cycle are given in section 8.
Table 5-1.
Physics Parameters for ANO-1, Cycles 4 and 5(*)
Cycle 4(b)
Cycle 5(c)
Cycle length, EFPD 387 455 Cycle burnup, mwd /mtU 12,111 14,259 Avg core burnup, EOC, mwd /mtU 20,505 2,3,188 Initial core loading, mtU 82.1 82.0 Critical boron -- BOC, ppm (no Xe)
HZP (d), group 8 ins 1562 1538 HFP, group 8 ins 1246 1370 Critical boron -- EOC, ppm HZP, group 8 100% wd, no Xe 418 487 HFP, group 8 100% wd, eq Xe 86 17 '
Control rod worths - HFP, BOC, % ak/k 2
Group 6 1.18 1.26 Group 7 1.02 1.47 Group 8 0.37 0.46 Control rod worths -- HFP, 455 EFPD, % Ak/k Group 7 1.00 1.59 Max ejected rod worth - HZP, % ak/k(*}
BOC (N-12), group 8 ins 0.76 0.53 455 EFPD (N-12), group 8 ins 0.82 0.58_
Max stuck rod worth -- HZP, % Ak/k BOC (N-12) 1.92 1.57 455 EFPD (N-12) 1.86 1.74 5-2 Babcock & Wilcox
--N Tcble 5-1.
(Cont'd)
Cycle 4 Cycle 5 Power deficit, HZP to HFP, % ak/k BOC 1.38 1.33 EOC 2.28 2 39_
Doppler coef f -- BOC,10-8 (ak/k/*F) 100% power (no Xe)
-1.57
-1.52 Doppler coeff -- EOC,10-3 (ak/k/*F) 100% power (eg Xe)
-1.71
-1.82 Moderator coef f -- HFP, 10~"
ak/k/*F)
BOC, (no Xe, crit ppm, group 8 ins)
-0.48
-0.49 EOC, (eq Xe, 0 ppm. group 8 out)
-2.78
-3.00 Boron worth -- HFP, ppm /% ak/k BOC 118 122 E0C 105 103 Xenon worth -- HFP, % ak/k BOC (4 EFPD) 2.59 2.58 EOC (equilibrium) 2.75 2.70 Effective delayed neutron fraction -- HFP BOC 0.00617 0.00626 EOC 0.00517 0.00517
(*} Cycle 5 data are for the conditions stated in this report.
The cycle 4 core conditions are identified in reference 2.
(
Based on 294 EFPD at 2568 MWe, cycle 3.
(" Based on 329 EFPD at 2568 MWt, cycle.4.
l1 (d)HZP denotes hot zero power (532F T"#8), HFP deno tes ho t full power (579 T,y ).
(*} Ejected rod worth for groups 5 through 7 inserted, group 8 as stated.
5-3 Babcock & Wilcox
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l}
dcv808on 3 (5/15/82) s Tabis 5-2.
Shutdown' Mergin CJ41culationn for ANO-1. Cycle 5
% ak/k
% ak/k
% ak/k Available Rod Worth Total rod worth, HZP 9.05 9.42 9.53 Worth reduction due to
-0.42
-0.42
-0.42 poison material burnup Maximum stuck rod, HZP
-1.57
-1.67
-1.74 Net worth 7.06 7.33 7.37 Less 10% uncertainty
-0.71
-0.73
-0.74 Total available worth 6.35 6.60 6.63 2
Required Rod Worth Power deficit, HFP to HZP 1.33 2.36 2.39 Allowable inserted rod 0.39 0.58 0.30 worth Flux redistribution 0.59 1.19 1.20 Total required worth 2.31 4.13 3.89 Shutdown margin (total 4.04 2.47 2.74 available worth minus total required worth)
Note: The required shutdown margin is 1.00% ak/k.
5-4 Babcock & Wilcox
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Mcv8cBsn 8 YM/XM/@3) 6.
THERMAL-HYDRAULIC DESIGN The fresh batch 7 fuel is hydraulically and geometrically similar to the pre-viously irradiated batch SB and 6 fuel.
The four batch 7 LTAs have been ana-lyzed to ensure that they are never the limiting assemblies during cycle 5 operat ion.
The results of the thentel-hydraulic analysis of the LTAs are in-cluded in reference 2.
The thermal-hydraulic evaluation of cycle 5 incorporated the methods and models described in references 1, 3, and 8.
The cycle 5 nuclear design al-lowed a reduction of the design radial-local peak from 1.78 to 1.71.
As a re-sult of this peaking reduction, the steady-state design overpower minimum DNBR increased from 1.88 to 2.05.
Table 6-1 summarizes the cycle 4 and 5 maximum design conditions.
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 ap-plicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis.
All plant operating limits are presently based on an original method of calculating rod bow penalties (reference 9) that are more conservative than those that would be obtained with new approved proce-dures given in reference 10.
For the current cycle of operation, this subro-gation results in a DNBR margin sufficient to offset the 4% reduction in DNBR due to fuel rod bowing.
2 6-1 Babcock & \\Milcox
t
- e Table 6-1.
Maximum-Design Conditions, Cycles 4 and 5 Cycle 4 Cycle 5 Design power level, MWe 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 1.78 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 Flow 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 2(b) 100% power, 468 449 3
10 Btu /h-ft CHF correlation BAW-2 BAW-2 Minimum DNBR At 112% power 1.88 2.05 At 108% power 2.01 2.18 At 100% power 2.30 2.39
(*) Heat flux was based on densified length (in the hottest core location).
( } Based on average heat flux with reference peaking.
Babcock & \\Milcox 6-2
R2 vision 2 (5/15/82)
I 7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cycle 5 parameters to determine the effect of the cycle 5 reload and to en-
~
sure that thermal performance during hypothetical transients is not degraded.
The effects of fuel densification on the FSAR accident results have been eval-usted and are reported in referende'8.
Since batch 7 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.
A study of the major FSAR Chapter 14 accidents using the cycle 5 iodine and noble gas inventories concluded that the thyroid and whole body doses are less than 4.1% of the 10 CFR 100 limits for all accidents except the HHA.
For the MHA, the 2-hour dose to the thyroid at the exclusion area boundary increased to 157 Rem, which represents 52% of the 10 CFR 100 limits.
The corresponding 2-hour whole body dose for the MHA increased by 6% to 7.07 Rem, which repre-sents 28% of the 10~CFR 100 limits.
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 uncertaintiec.
First-core values (FSAR values) of core thermal parameters and subsequent fuel batches are compared to parameters used in cycle 5 analyses in Table 4-2.
The cycle 5 thermal-hydraulic maximum design conditions are compared to the previous cycle 4 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 5 are compared in Table 7-2.
i i
Babcock & \\Vilcox l
7-1 l
l
~ - - -
Mov8sion R (6/13/@l) w A generic LOCA analysis for a B&W 177-FA, lowarsd-loop NSS has been performsd using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103).1 This analysis is generic since the limiting values of key parame-l1 ters for all plants in this category were used.
Furthermore, the combination of average fuel temperatures as a f unction 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 12 provide conservative l1 results for the operation of the reload cycle.
Table 7-1 shows the bounding values for allowable LGCA peak LHRs for ANO-1 cycle 5 fuel.
The basis for two sets of LOCA limits is provided in reference 13.
l1 It is concluded from the examination of cycle 5 core thermal and kinetics properties, 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 5.
Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 5 is considered to be bounded by previously accepted analyses. The initial conditions for the transients in cycle 5 are bounded by the FSAR, the fuel densification re-port, and/or subsequent cycle analyses.
Table 7-1.
Bounding Values for Allowable LOCA Peak Linear Heat Rates 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 7-2 Babcock & \\Milcox
=,
Rrvision 2 (5/t5/82)
Table 7-2.
Comparison of Ksy Parrmet0rs for Accident Analysis FSAR and densification ANO-1 Parameter report value cycle 5 Doppler coeff (BOC), 10-3 ok/k/*F
-1.17
-1.52 Doppler coeff (EOC), 10-s ak/k/*F
-1.30
-1.82 l2 Moderator coeff (BOC), 10-" ak/k/*F 0.0(*)
-0.49 9
Moderator coeff (EOC), 10~" ak/k/*F
-4.0
-3.00 l2 All-rod group worth (HZP), % ak/k 12.9 9.05 Initial boron concentration, ppm 1150 1370 Boron reactivity worth (HFP),
100 122
- pps/% ak/k Max ejected rod worth (HFP), % ak/k O'65 0.32 l2 Dropped rod worth (HFP), % ak/k 0.65 0.20
(*)+0.5 x 10-" ak/k/*F was used for the moderator dilution analysis.
( ) 3.0 x 10-" ak/k/*F was used for the steam line failure analysis.
7-3 Babcock & \\Milcox
Figure 8-6.
BORIC ACIO ADDITION TANK VOLUME AND CONCENTRATION REQUIREMENTS VS RCS AVERAGE TEMPERATURE (TECH SPEC FIGURE 3.2-1) 8700 PPM OPERATION AB0VE AND 5000 TO THE LEFT OF THE CURVES IS ACCEPTABLE 9500 PPM
/ 10,000 PPli a
f /
4000 j
3
/,
p12,000 PPM 2
l
/
/
/
3000
/
5 7
f
/
/
'b
/
f
~
//
/
o s' f
=
a 2000 j
f l f
4
/
/
/
//
l 1000
/
/
/
l 0
i i
200 300 400 500 600 RCS Average Temperature, F i
Figure 8-7.
Rod Position. Limits for Four-Pump Operation From 0 to 60 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.3.2-1A) 110 100 (134.102)
(271.102) u POWER LEVEL SHUTOOWN CUT 0FF 90 MARGIN
@.D LIMIT 80 (258.S0)
OPERATION IN 70 THIS REGION IS NOT RESTRICTED
[
ALLOWED REGION g
60
~
50 (67.50 (175.50)
G REGION 40 J'
30 20 0.13) 10 (0,0) 0 0'
20- 40 60 80 100 120 140 160 180 200 220 240 260 280 300
,0 2,0 4,0 60 g0100 GROUP 7 0
g0 4,0 6,0 SQ 100 GROUP 6 0
20 40 80 100 I
i
?
t GROUP 5 Rod Inaex, % WO
/
a-24 Babcock & Wilcox
Fipure 8-10.
90D POSITION LIR!TS FOR FGUR-PUM? OPERATICN FROM 400 10 TC 455-t t0 EFPD - ANG-1, CYCLI 5 (IECH SPEC FIGURE 3.5.2-10) 110 1
( 85,102)-
~
(230,102)
(283,92) 90 SHUIDOWN MARGIN RESTRICTED 80 LIMIT REGION (255,80) 70 g
OPERATION IN THIS 35 REGION IS NOT ALLOWE0 60 g
N 50 (162,50)
(181,50) o f
40 E
30 PERMISSIBLE 20 OPERAT1NG 10 -
(92,15)
(0,8.5) 0 i
e i
e i
e i
i i
i e
i i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Ron Index, % WD y
2,0 4,0 60 8,0 100 0
20 40 60 80 100 I
i I
I i
t 0
20 40 60 80 100 GROUP 6 e
i e
t i
t GROUP 5 t
1 s
~
o Figure 8-11.
Rod Position? Limits for,Three-Pump Operation e h S ec F ur 3.
k) x 110 100 90 80 SHUT 00WN (134,77)
MARGIN l
(250,77)\\
70
- OPERATION IN LIMIT THIS REGION
/
E 60 IS NOT RESTRICTED ALLOWED REGION
=2~ 50 (175.50)
PERMISSIBLE
- , 40 (67,38)
OPERATING t
REGION
$ 30 20 10 0
0' 20 40 60 80-100 120 140 160 180 200 220 240 260 280 300 0
2p 4,0 6,0 8E 190 s'
GROUP 7
,0 2p 40 6,0 8,0
, ' 10,0 GROUP 6 0
20 40 80 100 f
f 9
t GROUP 5 Rod Incex, % WD l
l l
\\
^
1 l
8-28 Babcock & Wilcox
~
~
Figure 8-14.
R00 P3S:T!CN LIWITS FOR THREE-pud? 0? ERAT!DN FRCM 400 i 10 TO 455 i 10 EFPS-ANO-1, CICLE 5 (TECH SPEC FIGURE 3.5.2-20) s 110
/
0 100
\\
90 RESTRICTED REGION 80 (230,77)
(247.6,77)
=
]
70 OPERATION IN 1
SHUTDOWN g
THIS REGION IS NARGlH E
NOT ALLOWED 60 LIMIT e
50 J
PERMISSIBLE
,E OPERATING 40 (162,38)
REGION 30 20 (0.6.9)
! \\(92,11.8) 10 0
e i
e i
i e
i i
i i
e i
i i
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index, 5 WO O
- 2,0 ',40 60 80 1,00 0
20 40 60 80 100 GROUP 7 e
i i
I i
0 20 40 60 80 100 GROUP 6 i
i i
i GROUP 5
Redision 1 (4/ 15/81) w Figure 8-15.
Rod Position Limits for Two-Pump Operation From 0 to 60 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2E) 110 100 90 80 70 C
60
~
OPERATION IN SHUTOOWN (134.52)
THIS REGION MARGIN (182.52) 50 IS NOT
~
LIMIT se ALLOWED
%g 30 (67.26)
PERMISSIBLE OPERATING 20 REGION (0,7) 1 (0,0)O.
r 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 O,
2,0 40 6,0 8,0 10,0 GROUP 7 0
20 4,0 6,0 80 10,0 l
GROUP 6 l
0 2,0 4,0 60 80 100 GROUP 5 Rau Inaex, % WO 8-32 Babcock & Wilcox
~
~
tigure 8-18.
ROU POSITION LIMITS FOR TWO-PUNP DPERATION FROM 400 1 10 TO 455 i 10 EF?D-ANO-1, CYCLE 5 (TECH SPEC FIGURE 3.5.2-2H) 110 100 90 80 E
70
=
N O
60 e
50 (230,52)
E 40 OPERATION IN SHUTOOWN TNIS REGION MARGIN LIMIT IS NOT ALLOWED 30 i
(162,26)
PERMISSIBLE 20 OPERATING REGION 10
- (0,5.3)
- (92,8.5) 0 i
i i
e i
i e
i 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 O
20 40 60 80 100 R
R I
I t
0 20 40 60 80 100 I
I I
e i
B 0
30 4,0 60 80 1,00 S
GROUP 5 a
N 4
Figure 8-19.
Operational Power Imbalance Envelope for Operation From 0 to 60 EFPD -- ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-3A)
. 110
(-18,102) o (25,102)
, igg
(-19,92)
(25,92)
. 90
(-26,80) 80 (32,80)
PERMISSIBLE RESTRICTE0 OPERATING RESTRICTED REGION REGION REGION 60 50
- - - 40
=
5 ga -
30 de -
20
^
a; W
g3 - - 10 t
a f
f f
e-I f
-40
-30
-20
-10 0
10 20 30 40 Axial Power Imoalance, %
l 8-36 Babcock & Wilcox
--~~-
\\
I d'
+
i4 1
Figure 8-22.
OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 400 1 10 TO 455 1 10 EFPD-ANO-1, CYCLE 5 l
(TECH SPEC FIGURE 3.5.2-30)
-- 110
(-19,102) a (20,102) 100
(-21, 92 )
90
(-26,80) 80 RESTRICTED REGION PERMISSIBLE OPERATING REGION RESTRICTED 60 REGION 50 C
=
40 ro
'N 30 e
h d-20 i
2 10 y
i I
i t
i I
I
?
-40
-30
-20
-10 0
10 20 30 40 Axial Power imoalance, ",
ll t
1, s
y t,
Figure 8-27.
LOCA Limited Maximum Allowable Linear Heat Rate (Tech Spec Figure 3.5.2.4) 21 20 2
19 1
a J
18
[l BALANCE OF CYCLE E
F 16 o
s
/
FIRST 50 EFPD
/
3
/
E 14 13 12 0-2 4
6 8
10 12 Axial Location of Peak Power from Bottom of Core, ft I
l t
i 8-44 Babcock & Wilcox l
]
\\1
/ **
s M
Fi gure 8-26.
APSR POSITION LIMITS FOR OPERATION FROM 400 t 10 TO 455 10 EFPD - ANO-1, CYCLE 5 (TECH SPEC FIGURE 3.5.2-40) 110 100 (20,102) 90 80 si RESTRICTED 70 g
REGION N
60
==
d 50 d?
PERMISSIBLE 40 OPERATING REGION l
30 20 10 0
i i
i i
i i
i 6
0 10 20 30 40 50 60 70 80 90 100 APSR Position, % Withdrawn
~,
._ _