ML19341C587
| ML19341C587 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/26/1981 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| IR-0281-15, IR-281-15, NUDOCS 8103030790 | |
| Download: ML19341C587 (16) | |
Text
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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE RCCK. ARKANSAS 72203 (501) 371-".000 February 26, 1981 IR-0281-15 Director of Nuclear Reactor Regulation ATTU: Mr. Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Arkansas Nuclear One-Unit 1 Docket No. 50-313 License No. DPR-51 Cycle 5 Reload Report Update (File:
1510.5)
Gentlemen:
As discussed in our February 19, 1981 submittal of Cycle 5 startup information we have updated the text of the Cycle 5 Reload Report to show changes made due to a ' reshuffle of fuel assemblies to minimize the use of failed fuel. Attached are the marked up pages that have r.arkers in the right column to show where in the text changes were n.cde.
Please review the changes made and contact us on any further information you may require.
Very truly yours, bAA84b.
David C. Trimble Manager, Licensing DCT-
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MEMDEA M:OOLE SOUTH UTILITIES SYSTEM
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OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Arkansas Nuclear One, Uni: 1 is the currently operating cycle 4.
This cycle 5 design is based on a design cycle 4 length of 329 effective full power days (EFP3).
l No anomalies occurred during cycle 4 that would adversely affect fuel per-formance during cycle 5.
Babcock t. Wilcox 2-1 s
e 3.
GENERAL DESCRIPTION The ANO-1 reactor core is described in detail in section 3 of the Arkansas Nuclear Station, Unit 1, Final Safety Analysis Report (FSAR).
The cycie-5 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 comprises dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4.
The fuel asse=1bies in all batches have an average nominal fuel loading of 463.6 kg of uranium, with the exception of four batch 7 lead test assc=blies (LTAs), which have a nomins1 loading of 440.0 kg uranie=. The undensified notaral active fuel lengths, theoretical densities, fuel and fuel rod di=ensions, and other related ftiel the parameters are given in Tables 4-1 and 4-2 for all fuel asse=blies except.
LTAs; the corresponding para:eters for the LTAs are included in reference 2.
The initial en-Figure 3-1 is the f uel shuf fle diagram for ANO-1, cycle 5.
richments of batches 55, 6, and 7 are 3.01, 3.19, and 2.95 we : 23sU, respec-tively.
All the batch 1C and batch 4 asse=blies, -7 of the twice-burned batch 5 assemblies, and 4 of the once-burned batch 6 assemblies will be discharged at the end of Cycle 4.
The remaining 49 twice-burned batch 5 assecblies (desig-nated batch 5B) will be shuffled to the core interior.
The 60 once-burned batua 6 assemblies will be shuffled to locations on or near the core periphery. The 68 fresh batch 7 essemblies will be loaded in a symmetric checkerboard pattern throughout the core.
Figure 3-2 is an eighth-core cap showing the assembly burnup and enrichnent distribution at the beginning of Cycle 5.
Reactivity is controlled by 61 full-length Ag-In-Cd control rods, 64 BPRAs, and soluble boron sh1=.
In addition to the full-length control rods, ei?ht e
axiel power shaping rods (APSRs) are provided for additional control of the xial power distribution. The cycle 5 locations of the 69 control rods and 1
the group designations are indicated in Figure 3-3.
The core locations of the total pattern (69 control rods) f'or cycle 5 are identical to those of the l{
reference cycle indicated in the reload report for ANO-1, cycle 4.3 The l
Babcock & Wilcox 3-1 I
e-., - -
,.n-
,--,5,~.-.
-,,r--
an..
e-n
-, +..
---n-..
.n..
Figure 3-1.
ruel shurne for N:0-1 Cycle 5 t
- FUEL TRANSFER cidal I
I 6B 6B 7
63 63 A
K4 K2 K14 K12 63 63 6B 7
53 7
63 63 6B I
P,4 L3 N3 F5
'41 3 L13 M12 63 7
6B 7
SB 7
53 7
6S 7
63 C
E10 K5 N2 N14 K13 Fil 63 7
63 7
53 7
53 7
53 7
68 7
63 011 M3 P
A7 F8 A9 P
H5 05 0
63 63 7
53 7
59 7
53 7
53 7
63 63 E
C10 F9 LTA C5 AS a10 E13 LTA F7 C5 63 63 7*
53 7
5B 7
53 7
53 7
53.
7 53 53 D9 C12 G1 E3 K3 C11 GIS C4 07 F
6B 7
53 7
53 7
53 63 53 7
53 7
53 7
63 99 312 F1 P10 C3 L2 F 1 ",
B4 97 G
7 53 7
53 7
IB 53 53 63 53 7
5B 7
53 7
y W-H L14 H14 H9 H3
';12 H13 u7 H2 F2 6B 7
53 7
53 7
53 63 53 7
53 7
53 7
40 P9 F12 L1 F14 03 35 L15 P:
, F7 6B 63 7
53 7
5B 7
53 7
53
/
53 7
33 63 N9 012 K1 05 G3 M11 (15 0:
17 L
m 63 63 7
53 7
53 7
53 7
53 7
63 C3 010 L9 LTA J *-:3 R6 RIO 311 L7A L7 C5 N
6 63 7
53 7
53 7
53 7
63 7
63 N
N11 Hil
?
R7 38 R9 P
E3
';3 63 7
53 7
53 7
53 7
63 7
63 0
M10 26 02 C14 GIO L11 63 53 63 7
53 7
53 is 63 p
E4 F3 D3 310 013 F '. 3 E12 6B 63 7
63 63 R
G4 G2 G14 G12 I
Z 1
2 3
4 5
6 7
8 9
10 11 12 13 la 15 3a tch
~
Previous Core Location O
O D' I (LTA*LeadTciti.schlies
( J%
el,.2 u.. w. '
I' P= Prech,$ rac*e r t.*0 J S t2nda rd P. ark D A;s.t,1y)
s
. Figure 3-2.
Er.richment and Burnup Distribution, AN0-1 Cycle 5 off 329 EFPD Cycle 4 8
9
'10 Ik 12 13 14 15 3.01 3.19 3.01 2.95 3.01 2.95 3.01 2.95 H
20422 12967 17814 0
20180 0
16893 0
3.01 2.95 3.01 2.95 3.01 2.95 3.19 K
16884 0
15844 0
11883 0
12205 3.01 2.95 3.01 2.95 3.19 3.19 20992 0
13712 0
10196 13476 L
3.01 2.95 3.19 3.19 M
20981 0
13778 12974 3.19 2.95 3.19 N
13754 0
13556 I
3.19 0
14292 P
R Initial Enricnment, X.XX wt%2350 BOC Burnup, XXXXX mwd /mtU D
w
?.i.
This vorst-case power history was then cc= pared against a generic analysis to ensure that creep-ovalization will not affect. fuel perfor=ance during ANO-1 cycle 5.
The generic analysis was perfor:ed based on ref erence 5 and is ap-plicable for the batch 5 fuel design.
=:
The creep collapse analyses predicts a collapse time greater than 35,000 ef-
.=
f ective f ull-power heurs (EFFH), which is longer than the =axi=u: expected residence time of 30,888 EFFH (Table 4-1).
4.2.2.
Claddin; Stress The ANO-1 stress paraceters for batch 4 and subsequent fuel are enveloped by a conservative fuel rod stress analysis.
For design evaluation, the pri=ary cembrane stress =ust be less than two-thirds of the minitun specified unir-radiated yield strength, and all stresses must be less than :he =ini=am speci-fied unirradiated yield strength.
In all cases, the =argin f.s greater than 307..
The following conservatisms with respect to the A:C-1 fuel were used in the analysis:
1.
Low post-densification internal pressure.
- w.
2.
Low initial pellet density.
3.
High system pressure.
4.
Eigh ther=al gradient across the cladding.
4.2.3.
C~addine Strain The fuel design criteria specify a limit of 1.0% on :ladding plastic tensile g
circu=ferential strain.
The pellet is designed to ensure that cladding plas-tic strain is less than 1% at design local pellet burnup cnd heat gene rat ion 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 diaceter and density and the lower tolerance value for the cladding ID.
4.3.
Thermal Design All fuel in the cycle 5 core is ther: ally similar.
The design of the four batch 7 lead test asse:blies is such that the thermal pe-icr:ance of this fuel is equivalent to or slightly better than the standard Mark B design used r
in the remainder of the fuel.. The thermal design aralysis of the LTAs using 8
the TACO-2 code is described in reference 2.
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Table 4-1.
Fuel Desien Parameters and Dimensions Batch 5
' Batch 6 Batch 7 Fuel assembly type Mark B4 Mark 34 Mark B4, Mark BE3 No. of assemblies (*)
49 60 64 Mark B4, 4 Mark SEB 1
i.
Fuel rod OD (nom), in.
0.430 0.430 0.430 m.
f.,
Fuel rod ID (nem), in.
0.377 0.377 0.377
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Flexible spacers.
Spring Spring Spring Zr-4 Zr-4 Zr-4
.i Rigid spacers, type L'ndensified active fuel 142.25 142.25 141.S0
.i f
length (nom), in.
l Fuel pellet OD (=ean 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 we %
U Average burnup, 30C, W d/ntU 16,467 12,892 0
Clade.ing collapse ci-e,
>35,000
>35.000 35,000 EFFH Estimated r asidence time, 25,560 28,680 30.888 I
EFFH (max)
(a)Four lead test assemblies (Mark SES) make up a total batch 7 reload of l
68 fuel assechlies.
These LTAs were analyzed and reported in ref erence j
2.
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l Babcock & Wilcox l
424 1
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Table 4-2.
Fuel Ther=al Analysis Parameters Batch SB Batch 6 Batch 7 64(a)
No. of asse=blies 49 60 Initial density, % TD 94.0 94.0 95.0 Pellet dianeter, in.
0.3695 0.3695 0.3686 Stack height, in.
142.25 142.25 141.80 Dens'ified Fuel Parameters Pellet dianeter, in.
0.3646 0.3646 0.36t.9 Fuel r. tack height, in.
140.5 140.5 140.74 Nominal linear heat rate 5.80 5.80 5.79 at 2568 st, kW/ft Avg fuel te=perature at 1320 1320 1310 nominal LER, F LHR capability, kW/f t(b) 20.15 20.15 20.15 Nominal core avg LHR = 5.80 kW/f: at 256S W.
~
(*)Four LTAs vert also analyzed; the results are reported in reference 2.
(b) Centerline fuel melt based on fuel specification values.
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Babcock s Wilcox 4-5
o 5.3.
Changes in Nuclear Design _
core design changes between the reference and reload There are no significant The calculational methods and design information used to obtain the cycles.
nuclear design parameters for this cycle were the same as those used important for the reference cycle. There are two significant operational changes f rom the reference cycle: the withdrawal of the APSRs during the last 35 EFFD of The cycle 5 and a change f rom a rodded to a f eed-and-bleed node of operation.
stability and control of the core in the f eed-and-bleed = ode with /PSRs recoved The calculated stabiiity index without APSRs is -0.0334 h-1, aave bean analyzed.
which demonstrates the axial stability of the core.
The operating limits (Tech-nical Specification changes) for the t '1oad cycle are given in section S.
Physics P srameters for ANO-1, Cycles 4 and 5(#
Table 5-1.
Cycle 4(b) Cycle 5(c) 387 435 Cycle leng:h. EFPD Cycle burnup, mwd /ntU 12,111 13,633 Avg core burnep, EOC, mwd /stU 20,505 22,562 f
Initial core loading, stU 82.1 82.0 Critical boron - EOC, ppm (no Xe)
HZP (d), grcup 8 ins 1562 1538 1246 1370 HFP, group 8 ins Critics' boron - EOC, ppm HZP, group 8100% ud, no Xe 413 598 HFP, group 8 100% wd, eq Xe 86 22 Control red wcrths -- RFP, BOC, % ak/k 1.18 1.26 Group 6 1.02 1.47 Group 7 0.37 0.46 Group 8 Control red worths - HFP, 435 ETP3, % ak/k 1.00 1.49 Group 7 Max ejected rod worth -- HZP, % ak/k
- BOC (N-12), group 8 ins 0.76 0.53 400 E"'D (N-12), group 8 ins 0.82 0.53 Max stu-rod worth -- EZP, % ak/k BOC,T-12) 1.92 1.57 400 EiPD (N-12) 1.86 1.67 5-2 Babcock & 'Nilcox
r Table 5-1.
(Cont ' d) l Cycle 4
, Cycle 5 Power deficit, EZP to HFP, % Sk/k BOC 1.38 1.33 EOC 2.28 2.34
[
Doppler coef f - BOC,10-5 (ak/k/*F)
(
100% power (no Xe)
-1.57
-1.52 Doppler coef f -- EOC,10-s (Ak/k / *F) 100% power (eq Xe)
-1.71
-1.84 l
Moderator coeff - EFP, 10~"
ak/k/*F)
BOC, (no Xe, crit pp=, group 8 ins)
-0.48
-0.49 l
EOC, (eq Xe, O pps, group 8 out)
-2.78
-2.87 Boron worth - HFP, ppm /% Ak/k BOC 118 122 EOC 105 105 I
Xenon vorth -- HFP, % ok/k BOC (4 EFFD) 2.59 2.58 EOC (equilibriu=)
2.75 2.66 Eff ective delayed neutron fraction -- HFP BOC 0.00617 0.00626 0.00517 0.00515 EOC
("} Cycle 5 data are for the conditions stated in this report.
The cycle 4 core conditions are identified in ref erence 2.
(b) Based on 294 EFFD at 2568 MWt, cycle 3.~
(c) Based on 329 EFPD at 2568 MWt, cycle 4.
(d)HZP denotes hot zero power (532F T""8), HF7 deno te s ho t f ull power (579F Tavg).
i
(*)Ej ected rod worth 'for groups 5 through 7 inser ted, group S as stated.
l 4
i 5-3 Babcock & Wilcox
(,
Table 5-2.
Shutdown Maroin Calculations for ANO-1, Cycle 5 l
% ok/k
% ak/k
% ak/k Available Rod Worth Total rod worth, HZP 9.05 9.42 9.15 Worth reduction due to
-0.42
-0.42
-0.42 poison material burnup Maximum stuck rod, HZP
-1.57
-1.67
-1.48 Net worth 7.06 7.33 7.25
)
Less 10% uncertainty
-0.71
-0.73
-0.72 Total available worth 6.35 6.60 6.53 Required Rod Worth Power deficit, HFP to HZP 1.33 2.36 2.34~.
r Allowable inserted rod 0.39 0.58 0.30 worth Flux redistribution 0.59 1.19 1.20 i
Total required worth 2.31 4.13 3.84 Shutdown margin (total 4.04 2.47 2.69 available worth minus total required worth)
Note: The required shutdown margin is 1.00% ak/k.
ie l
l
(
Figure 5-1.
ANO-1 Cycle 5, B0C (4 EFrni Two-Dimensional Relative Power Distribution - Full Power Equilibrium Xenon, Normal Rod Positions l
i 8
9 10 11 12 13 14 15 1.09 1.21 1.16 1.27 1.02 0.75 H
K 1.08 1.17 1.15 1.25 1.24 1.13 0.61 1.05 1.19 1.06 1.17 0.97 0.44 g
1.06 1.20 1.08 0.68 g
1.17 0.92 0.44 N
0.52 -
0 P
R X
Inserted Rod Group No.
XXX Relative Power l Density
r 4
5 6.
THER.%L-EYDPAULIC DESIGN The fresh batch 7 fuel is hydraulically and geometrically si ilar to the pre-viously irradiated batch 53 and 6 fuel.
The four batch 7 LTAs have been ana-ly:ed to ensure that they are never the li=iting asse:blies during cycle 5 The results of the thermal-hydraulic analysis of the LTAs are in-op eration.,
~
cluded in reference 2.
The thermal-hydraulic evaluation of cycle 5 incorporated the zechods and codels described in references 1, 3, an.
3.
The cycle 5 nuclear design al-loved 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 over7ower =ini=cs DN3R intreased frc= 1.53 to 2.05.
Table 6-1 su=sarizes the cycle 4 ar.d 5 =axi=u:
design conditions.
gg 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 g
applicable cycle tinus 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 8A) that are
=cre conserv:2tive than those that would be obtained with new approved procedures given in Reference 9.
For the current cycle of operation, this subrogation
- 2 results in a DNBR margin in excess of 3.8%, which is cartially used to offset the reduction in DNBR due to fuel rod bowing.
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Babcock & Wilcox cr:
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l 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 deter =ine the ef fect of the cycle 5 reload and to en-sure that ther=al perfor=ance during hypothe*'..1 transients is not degraded.
The effects of fuel densification on tS FEAR accident results have been eval-usted and are reported in reference S.
Since batch 7 reload fuel asse=blies 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 =ajor FSAR Chapter 14 accidents using the cycle 5 iodine and noble gas inventories concluded that the thyroid and whole body doses are less than 12.7% of the 10 CFR 100 limits f or all accidents except the FSA.
For the IEA, the 2-hour dose to the thyroid at the exclusion area boundary did not change from 153 Re=, which represents 51% of the 10 CFR 100 li=its.
The cor-responding-2-hour whole body dose for the !F.A decreased by 33.0% to 6.7 Re=,
which represents 27% of the 10 CFR 100 limits.
7.2.
Accident Evaluation The key parameters that have the greatest effect on deter =ining the outec=e can typically be classified in three =ajor areas: core ther=al of a transient para =eters, ther=al-hydraulic para =eters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.
Core ther=al properties used in the FSAR accident analysis were design oper-First-core ating values based on calculational values plus uncertainties.
values (FSAR values) of core ther=al parameters and subsequent fuel batches are co= pared to para =eters used in cycle 5 analyses in Table 4-2.
The cycle 5 ther=al-hydraulic =aximum design conditions are co= pared to the previous cycle 4 values in Table 6-1.
These parameters are co==on to all the accidents considered in this report. The key kinetics parameters fro = the FSAR and i
cycle 5 are co= pared in Table 7-2.
Babcock & Wilcox 7-1 1
(
o Table 7-2.
Comparison of Key Parameters for Accident Analysis FSAR and.
densification ANO-1 report value cycle 5 Parameter Doppler coeff (BOC), 10-s ak/k/*F
-1.17
-1.52 Doppler coeff (EOC),10-s Ak/k/*F
-1.30
-l.84 Moderator coeff (BOC), 10~" ak/k/*F 0.0(*
.49 Moderator coeff'(EOC), 10~" ak/k/*F
-4. 0 (b)
-2.87 All-rod group worth (H2P), % ok/k 12.9 9.05 Initial boron concentration, ppm 1150 1370 Boron reactivity worth (HFP),
100 122 ppm /% ak/k Max ejected rod worth (HFP), % ak/k O.65 0.30 Dropped rod worth (HFP), % Sk/k 0.65 0.20
(*}+0.5 x 10-" ik/k/*F was used for the moderator dilution analysis.
F
(
10-" ak/k/*F was used for the steam line failure analysis.
-3.0 x L
Babcock & Ylilcor 7-3 i
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REFERENCES 1 Arkansas Nuclear One. Unit 1 -- Final Saf ety Analysis Report, Docket 50-313, Arkansas Power & Light.
2 T. A. Coleman and J. T. Wills t, Extended Burnug Lead Test Assembly Irradi-ation Program, BAW-1626, Babcc ek & tilcox, October 1980.
3 Arkansas Nuclear One, Unit 1 - cycle 4 Reload Report, 3AW-1504, Babcock &
Wilcox, October 1978.
J. H. Taylor to S. A. Varga, Letter, "3PRA Retainer Reinsertion," January 14, 1980.
?rogram to Deter =ine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10054. Rev. 1, Babcock & Wilcox, November 1976.
Y. H. Hsii, et al., TACO-2 -- Fuel Pin Performance Analysis, EAW-10141P, Babcock & Wilcox, January 1979.
C. D. Morgan and H. S. Kao, TAFY -- Fuel Pin Temperature and Gas Pressure 7
Analysis, BAW-10044, 3abcock & Wilecx, May 1972.
Arkansas Nuclear One, Unit 1 - Fuel Densification gecort,,3 AW-1391, Babcock
& Wilcox, June 1973.
8A D. F. Ross and D. G. Eisenhut (NRC), memorandum to D. S. Vassallo and K. R.
Goller (NRC) on " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations For Light Water Reactors," dated December 8, 1976.
- L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Let ter, " Evaluation o f In-teri Procedure for Calculating DNSR Reduction Due to Rod' 3cv," October 18, 1979.
18 ECCS Analysis of 3&W's 177-FA Lowe red-Loop NSS, B AW-10'_03. Rev. 1, Babcock
& Wilcox, September 1975.
11 J. H. Taylor (B&W Licensing) to R. L. Baer (Reactor Safety Brancu, USNRC),
. Letter, July 8, 1977.
12 J. H. Taylor (B&W Licensing) to L.' S. Rubenstein (USSRC)
Letter, Septether 5, 1980.
Note:
All Babcock & Wilcox reports mentioned above are f rom K?GD, Lynchburg, Virginia.
Babcock t \\Vilcox A-1
_