ML19322B861
| ML19322B861 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/30/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| References | |
| BAW-1493, NUDOCS 7912060681 | |
| Download: ML19322B861 (50) | |
Text
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RETURN TO REGillATORY CENTHAL files
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June 1978 i
OCONEE UNIT 1, CYCLE 5
-- Reload Report -
RETURN TO REG'!LAT03Y GEhIRR gBabcock &
ROUkOtB 7912060 Q / g
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BNa'-1493 June 1978 OCONEE UNIT 1 CYCLE 5
- Reload Report -
BABCOCK & W1LCOX Power Generation Group 3
Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock s. Wilcox
BLANK FRAME FOR PROPER PAGINATION
t CONTENTS Page 1.
INTRODUCTION.
1-1 2.
OPERATING HISTORY 2-1 3.
CENERAL DESCRIPTION 3-1 6.
FUEL SYSTEM DESIGN.
4-1 4.1.
Fuel Assembly Mechanical Design 4-1 4.2.
Fuel Rod Design 4-2 4.2.1.
Cladding Collap,e 4-2 4.2.2.
C1.idding Stress 4-2 4.2.3.
C1.idding Strain.
4-2 4.3.
Thermal Design.
4-3 4.4.
!bteriel Cksign 4-3 4.5.
Operating Experience 4-3 a
5.
NUCLEAR DESIGN.
5-1 5.1.
Physics Characteristics 5-1 5.2.
Analytical Input 5-2 5.3.
Changes in Nuclear Design 5-2 6.
T!!ERMAL-ilYDRACLIC DESIGN.
6-1 s
i 7.
ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.
General Safety Analysis 7-1 7.2.
Accident Evaluation 7-1 8.
PROPOSED H0DIFICATIONS TO TECHNICAL SPECIFICATIONS.
8-1 9.
STARTL'P PROGRAM - PilYSICS TESTING 9-1 REFERENCES.
A-1 J
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Babcock s.Wilcox
List of Tables Table Page
.j 4
4 '.
4-1.
Fuc l Design Parameters a-d Dimensions 4-5 4-2.
Fuel Thermal Analysis P.raseters 5-2 1
3-1.
Oconee 1. Cycle 5 Physics Parameters 5-2.
Shutdown. Margin Calculatian for Oconee 1 Cycle 5.
5-4 6-2 6-1.
Thermal-Hydraulic Design Conditions 7 'l e
7-1.
Conparison of Key Paracieters for Accident Analysis 5
7-3 7-2.
LOCA Limits. Oconee 1. Cycle 5 List of Figures 4
5 Figure 3-2 3-1.
Oconee 1 Cycle 5 - Full Core Loading Diagram.
3-2.
Enrichtent and Burnup Distribution for Oconee 1 Cycle 5 3-3 4
3-1.
Control Rod Locations for Oconee 1, Cycle 5.
3-4 3-1.
Soc. Cycle 5 Tuo-Dimensional Relative Power Distribution - Full j
5-5 P.:v e r, Equilibrius Xenon, Norr.al Rod Positions 8-1.
tore Prote. tion Safety L1=its, Oconee Unit 1 6-2 S-2.
Protective System Maximum Allowable Setpoints. Oconee Unit 1 8-3 1
6-1.
- -d Position L1=its for Four-Pu=p Operation, Oconee Unit 1 8 ~.
B '..
had Position Limits for Fcur-Pump Operation, Oconee Unit 1 8-5 l
6-5.
Rad Position Limits for Two-and Three-Pump Operation, oconee Unit 1...
8-6 6-6.
Kod Position 1.1=its for Two-and Three-Pump Operation, f
econee Unit 1.
8-7 6-7
'oser I= balance Limits. Oconee Unit 1..
8-8 b-8.
Paver Iribalance Limits, Oconee Unit 1.......
8-9 S-9.
APSR Position Limits, Oconee Unit 1 8-10 b-10.
APSR Position Limits. Oconee Unit 1...
8-11 i
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Babcock s.Wilcox
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i 1.
INTRODUCTION This report justifies operation of the Ocanee Nuclear Station, Unit 1. cycle 5 at a rated core power of 2568 MWt.
The required analyses are included as out-lined in the USSRC document, " Guidance for Proposed License Amendments Relat-ing to Refueling," June 1975. This report uses the analytical techniques and design bases documented in several reports that have been submitted to and approved by the USNRC.
Cycle 5 reactor a.,d fuel parameters related to power capability are summarized in this report and compared to those of cycle 4.
All accidents analyzed in the Oconee FSAR have been reviewed for cycle 5 operation; a detailed co: par-ison of cycle 5 characteristics to the FSAR analyses showed that no new anal-yses were necessary since cycle 5 parameters are conservative.
The Technical Specifications have been reviewed and modified where required for cycle 5 operation. Based on the analyses performed and taking into ac-I count the ECCS Final Acceptance Criteria and postulated fuel densification effcets, it is concluded that Oconee 1, cycle 5 can be safely operated at its licensed core power level of 2568 MWt.
Five fuel assemblies from batch 4 will be irradiated for a fourth cycle as part of a joint Duke Power /85W/ Dept, cf Energy program to demonstrate reliable I
fuel performance at extended burnups and to obtain post-irradiation data.
These assemblies will not adversely affect cycle 5 operation.
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I 1-1 Babcock s.Wilcox
4 2.
OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Oconee
- 1. cycle 5 is the currently operating cycle 4.
This cycle 5 design is based on a planned cycle 4 length of 235 EFPD rather than the design length of 292 EFPD.
Cycle 5 will operate in a feed-and-bleed code for its entire design length of 330 EFPD.
Initial cycle 4 operation was in a rodded mode. Ilowever, a quad-l rant power tilt was detected during cycle 4 power escalation, and the mode of operat ion was converted to feed -and-bleed to provide a larger margin for cy-cle 4 operation.2 The shuffle pattern for cycle 5 was designed to minimize the effects cf any pa.er tilts present in cycle 4.
No control rod interchange is planced Guring c ecle 5.
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3.
GENERAL DESCRIPTION The Oconee Unit I reactor core and fuel design basis are described in detail 3
in section 3 of the Final Safety Analysis Report for Oconee Nuclear Station, Unit 1.
The cycle 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 in-core instrument guide tube.
The fuel consists of dished-end, cylindrical pel-lets of uranium dioxide clad in cold-worked Zircaloy-4. The fuel assc=blies in all batches have an average nominal fuel loading of 463.6 kg of uranium.
The undensified nominal active fuel lengths, theoretical densitics, fuel and fuel rod dimensions, and other related fuel parameters are given in Tables 4-1 and 4-2.
Figure 3-1 is the core loading diagram for oconee 1, cycle 5.
The initial en-richment of the fresh batch 7 fuel is 3.02 wt % 235U.
The remaining batches
~ 135 4D. 5, and 6 were initially enriched to 3.20, 2.75, and 2.795 wt U, re-spectively. All the batch 4A and all but five batch 4B assemblies vili be discharged at the end of cycle 4.
The five remaining batch 4B asse=blies will be retained in cycle 5 and are redesignated as batch 4D.
The batch 4D, 5, and 6 assemblies will be shuf fled to new locations at the beginning of cycle 5.
The f resh batch 7 assemblies will occupy the periphery of the core and eight interior locations. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichsent distribution at the beginning of cycle 5.
Reactivity is controlled by 61 full-length Ag-in-Cd contral rods and by solu-ble boren shim.
In addition to the full-lencth control rods, eight axial power shaping rods are provided for additional control of the axial power dis-tribution. The cycle 5 locatiens of the 69 control rods and the group desig-nations are indicated in Figure 3-3.
The core locations of the total pattern (69 control rods) for cycle 5 are identical to those of the reference cycle indicated in the Oconee 1, cycle 4 reload report." The group designations, however, dif fer between cycle 5 and the reference cycle in order to minimize power peaking. Neither control rod interchange nor burnable poison rods are necessary for cycle 5.
Babcock s.Wi;cox 3-1
Ft;ure 3-1.
Ocence is Cycle 5 - Full Core Leading, Diagra:1 e
A 7
7 7
7 7
E7 Re E9 I
I 7
7 7
5 6
7 7
s M2 C6 F7 L2 M3 U4 79 C10 M14 7
6 S
S 6
5 6
5 5
6 7
911 C1)
L1 E3 DS C11 M13 LES 03 55 7
6 6
6 6
5 7
5 6
6 6
6 7
F3 A13 09 El 12 OS*
514 315 E)
A6 F13 7
6 5
6 6
43 6
6 3
6 5
7 66 C12 Al K1%
Db N8 710 07 A7 C4 CD 7
5 6
6 6
6
- t. )
B10 E4 512 F6 Ke F12 B4 E12 h
G11 t.
7 5
6 5
6 5
7 7
3 6
5 6
5 7
4 f
bl%
M11 Mll*
R12 M9 El.*
E7 li.
E)*
F 6
3 7
ED S
5 43 5
5 40 7
5 6
7 6i F10 F12 L4 G8 L12 F6 M12 Pg K11 7
5 6
6 5
7 9
7 5
6 5
6 1
1 to
$12 k9 09 h6 Da 510 G3 RF 04 E13 J
J 5
A 6
5 5
S 6
6 5
7 7
- 1. s
$10 Cl)
C1 02 011*
114 c15 C7 Re L13 l
7 6
5 6
6 43 6
6 5
6 3
7 g
Fil C1)
F1 DJ N3 511 D1)
F13 C3 F5 7
6 4
6 6
5 7
5 6
6 6
6 7
E2 06 L7 F2 Eli F16 L9 013 E14 7
6 6
6 7
MF t
7 7
7 F
I 7
7 7
7 7
._.I I
I I
I I
i i
i I
2 3
4 5
6 7
8 9
10 11 12 13 to IS
- 1. 4:! m t ths ice durr.eJ tatch 43 asse-b1Bes.
was Freview coet, locat tan.
a Batch En.
3-2 Babcock s. Wilcox
Figure 3-2.
Enrichment and Burnup Distribution for Oconee 1 Cycle 5 8
9 10 11 12 13 14 15 3.20 2.75 2.75 3.20 3.02 2.75 2.79 3.02 H
28,479 20,488 16,033 31,135 0
15,903 5,889 0
3.02 2.75 2.79 2.75 2.79 2.75 3.02 K
0 14,270 5,138 19,206 8,537 16,345 0
2.75 2.79 2.79 2.75 3.02 3.02 L
17,336 5,853 8,262 15,846 0
0 2.75 2.79 2.75 3.02 M
17,341 5,011 18,348 0
2.79 2.79 3.02 N
5,846 7,092 0
3.02 0
0 P
R x.xx Initial Enrichment xxxxx BOC Burnup, mwd /mtU i
Babcock a.Wilcox 3-3
Figure 3-3.
Cont rol i<od Locat ions f or oconee ' 1, Cycle 3 x
A 5
a 3
s 3
i l
C 1
7 7
1 t
.!6 l
D 6
8 4
8 E
1 5
2 2
5 1
f8 F
3 8
7 6
7 3
l 7
i C
7 2
6 4
2 5
4 6
3 6
1 4 5
-Y M
W-K l
7 2
4 4
2 7
L 3
4 7
6 7
8 3
t t
1 5
2 2
5 l 1
l 9
6 8
4 8
l6 l
1 7
7 1
0 i
8 f
I P
3 5
3
.~
R
-~
9 l
. 8 9 [ 10 1
2 3
5 6
7
,' 4 i
l l11, 12 13 to 15 i
I i
i e Croup Numbea a
b I
Croup
%.cfBMs Fec t iori I
8 Safety 2
-8 Safety di 3
9 Safety d
4 8
Safety S
8 Control 6
8 C.setrol 7
12 Contro1 8
a a5 m.
Total 69 1
1 3-4 Babcock & Wilcox i
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FUEL SYSTEM DESIGN 4.1.
Fuel Assembly Mechanical Design The types of fuel assemblies aru pertinent fuel design parameters and dimen-sions for Oconce 1, cycle 5 cre listed in Table 4-1.
All fuel assemblics are
!Jentical in concept and are mechanically interchangeable.
A.1 result _, ref-crences, and identified conservatisms presented in sectica 4.1 of the Oconee 1, cycle 4 reload report
- are applicable to the cycle 5 relcad core.
Five batch 4D Mark-B3 assemblies are remaining in the core f or their fourth cycle of irradiation and will experience barnups up to approxi=stely 41,000 h'd/mtU as part ot' a joint Duke Power /E6W/ Dept. of Energy program to demon-st rat e extended burnup feasibili ty in L***Rs.
The Mark-3 f uel as - ub'y Lechan-ical design will maintain ite structural integrity with these burnaps. Anal-vses of post-irradiation ex2=taation (PIE) data from two cycles of operation in the Lconee 1 reactor shov that all parameters measurea indicate that ex-tended oper1 tion is quite feasible. The parameters investigated include fuel rod and assembly growth, fuel swelling, and holddown spring force. The 'n-tended peak burnups of batch 4D fuel are within the original mechanical peak oesign limits reported in the Oconee FSAR.3 Design parameters can be affected by burnup, effective full power tire, or calendar residence tice.
Those param-eters affected most by the amount of irradiation are fuel rod and assembly growth and fuel swelling. Since burnup is within conservative design limits, growth will be acceptable. Section 4.2.3 discusses fuel swelling as it relates to cladding strain. The holddown spring force is affected by residence time as well as burnup. Evaluation of the PIE data indicates that the holddown spring will meet performance requirements through the fourth cycle of irradia-ti l
1 4-1 aabcock & Wilcox
.2.
Fuel Rod Design
-.2.1.
Cladding Co1*ipse Creep collapse analyses were performed for three-cycle assembly power histo-ries as well as f or batch 4n's f our-cycle asse:bly power histories. For cy-for 40 ele 5. the batch 5 fuel is more limiting than all other batches execpt because of its previous incore exposure time.
The batch 5 and 4D asse=bly pcwer histories were analyzed, and the cost limiting asse:bly f res each batch was de te rmined.
Die power histories f or the most limiting assemblies were used to calculate the fast neutron flux 1cvel for the energy range above 1 MeV.
The collapse limiting assembly from each batch was conservatively deter-t ime f or the most
.ined to be nore than 30,000 ef fective f ull-power hours (EFFH), which is lenger than the maximum projected batch 5 residence time of 21,456 EFFH (three I
and the maximum projected batch 4D residence time of 23,4e9 EEPH (four eycles)
The creep collapse analyses were performed based en ti e conditions 9
cycles).
-et farth in references 4 and 5.
_y.2.2.
Claddine Stress 1 stress parameters are enveleped by a conservative fuel rod stress The Ocence Since worst-case stress conditiens are a* BOL, the batch 4D fuel is analysis.
also bounded by the fuel rod stress analysis. For design evaluation, the pri-nary =embrane stress must be less than two-thirds of the sinisus specified un-irradiatsd yield strength, and all' stresses (primary and secondary) cust be l
Iess than the minimra specified unirradiated yield strength. The margin is in excess of 307. In all cases. k'ith respect to Oconee I fuel, the following cen-se r v: tis s vere used in the analysis:
1.
lew post-densification internal pressure.
2.
Low initial pellet density.
3.
High system pressure.
4.
High thermal gradient across the cladding.
e The stresses reported in reference 6 for. core 1 fuel represent conservative I
)
l values with respect to the cycle 5 core.
i 4.._2. 3.
Cladding Strain 7
l Die f uel design crit ria specify a limit of 1.0% on cladding circumferential l
plastic strain. The pellet design is established for plastic cladding strain Babcock s.Wilcox 4-2 k
i tl
of less than 12 at maximum design local pellet burnup (55,000 Wd/ met *)
and heat generation rate (20.15 kW/ft) values that are higher than the values the Oconee 1 fuel is expected to see, including batch 4D.
The strain analysis is also based oa the maximum Specification value for the fuel pellet diaseter and density and the lowest permitted Specification tolerance for the cladding ID.
4.3.
Thermal Design All fuel assemblies in this core are thermally similar. The fresh batch 7 fuel inserted for cycle 5 operation introduces no significant dif ferences in f uel thermal performance relative to the other f uel remaining in the care.
The desiRn minimus linear heat rate (LHR) capacity and the average fuel temp-erature for each batch in cycle 5 are shown in Table 4-2.
LHR capabilities are based on centerline fuel melc and were established using the tan *-3 code 7 with fuel densification to 96.5% of theoretical density. The five batch 4D fuel assemblies have an EOC burnup of about 41,000 mwd /mtt*.
The EOL =aximum pin pressure for these assemblies is well below the system pressure of 2200 psia.
6.4.
'faterial Desi :31 The hatch 7 fuel assemblies are not new in concept, nor do they utilize dif-ferent component materials. Therefore, the chemical compatibility of all pos-sible fuel-cladding-coolant-assembly interactions for the batch 7 fuel as-semblies are identical to those of the present fuel.
4.5 Operating Experience Babcock & Wilcox operating experience with the Mark-B, 15 by 15 fuel assembly has verified the adequacy of its design. As of February 23, 1978, the exper-ience described below has been accumulated for the eight operating B&*.* 177-fuel assembly plants using the Mark-B fuel assembly. In addition, Three Mile Island 1* nit 2 achieved initial criticality on March 28, 1978, and is currently in the startup testing phase that precedes cotanercial operation.
Max assembly Cumulative burnup, mwd /mtU Current not elect.
Reactor evele Incore Disch.
output, mWh g
Oconee 1 4
27,200 25,300 20,385,249 Oconce 2 3
26,700 26,800 15,2^8,595 l
Oconee 3 3
27,140 27,200 16,182,813 j
Babcock s Wilcox 4-3 J
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Max assembly Cumulative burnup, %'d/ntt, Current net elect.
Re : tor evele incore Disch.
output,rGh TMI-1 3
31,720 25,860 18,430,506 AN0-1 2
28,290 17,650 14,575,320 Rancho Seco 2
22,3sJ 17,170 10,297,637 4,936,412
. Crystal River 3 1
10,430 1,009,741 l
Davis-Besse 1 1
2,490 1
Table 4-1.
Fuel Design Parameters and Disnensions 1
1 Thrice-Twice-Once-burned burned burned Fresh i
FA type Mark-B3 Mark-B4 Mark-B4 Mark-B4 No. of FAs 5
60 56 56
]
Fuel red OD, in.
0.430 0.430 0.430 0.430 l
Fus:1 rod ID. in.
0.377 0.377 0.377 0.377 Flex. spacers, type Spring Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 j
t'ndensif active feel 142.0 142.6 142.25 13.2.25
)
length (nom), in.
Fuel pellet initial
$94.5 93.5 94.0 94.0 density (non), ?. TD Fuel pellet OD (mean 0.3685 0.3700 0.3695 0.3695 specif), in.
Initial fuel igrich.,
3.20 2.75 2.79 3.02 WU
\\
i wt 1
BOC burnup (avg),
30,604 17,011 6,539 0
mwd /mtU l
Cladding collapse
>30,000
>30,000
>30,000
>30,000 time. EFPH Estimated residence 28,469 21,456 22,440 26.496 time (max). EFPH I
Babcock & Wilcox 4_4 i
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Table 4-2.
Fuel Thermal Analysis Para eters Batch
- p(a)
$(a!
6( }
7 No. of assemblics 5
60 56 56 Noninal pellet density, 2 TD 95.5 93.5 94.0 94.0 Pellet diameter, in.
0.3685 0.3700 0.3695 0.3695 0)
Stack height, in.
141.0 142.6 142.25 142.25 Densi f ied Fue l Parameters Pellet diameter, in.
0.3b40 0.3645 0.3646 0.3646 Fuei stack height, in.
140.30 140.46 140.47 140.47 Nominal LilR at 2568 %*t, kW/ft 5.80 5.80 5.80 5.80 Av,c fuel temp at nominal LiiR. F 1320 1320 1320 1320 LIIR to rt fuel melt, kW/ft 20.15 20.15 20.15 20.15
(" Data from reference 4.
(
Censervative calculational parameter.
(C)Densification to 96.54 TD assumed.
4-5 Babcock & Wilcox
BLANK FRAME FOR PROPER PAGINATION
5.
NL' CLEAR DESIGN 5.1.
Physics Characteristics Table 5-1 compares the core physics parameters of design cycle 5 with those of reference cycle 4.
The values for both cycles were generated using PDQ07.
The average cycle burnup will be higher in cycle 5 than in the design cycle 4 because of the longer cycle 5 length. Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 5 at full power with equilibrium xenon and nornal rod positions.
l The critical boron concentrations for cycle 5 are cotip rable to those of the design cycle 4, The control rod worths for hot full power differ between cy-cles due to changes in group designations as well as changes in radial flux distributions and isotepics. The ejected rod worths in Table 5-1 are the max-imun calculated values within the allowable rod insertien limits. Calculated ejs etud 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 presented in section 8.
The maximum stuck rod worth for cycle 5 is greater than that for the design cycle 4 at BOC and approxinately the same at EOC.
All safety criteria associated with these warths ate met.
The adequacy of the sSutdown margin with cycle 5 stuck rod worths is demonstrated in Table 5-2.
The following conservatis=s were applied for the shutdown calculation =:
1.
Poison material depletien allowance.
2.
10 uncertainty on net rod worth.
3.
Flux redistribution penalty.
Flux redistribution was accounted for since the shutdewn analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown mar-gin is presented in the Oconee 1. cycle 4 reload report."
The cycle 5 power deficits from hot zero power to hot full power dif fer f rom those for the design cycle 4 because of the longer cycle 5 design length.
The differential boron worths and total xenon worths for cycle 5 are greater 5-1 Babcock s.Wilcox
than or equal to thone for the design cycle 4 because of fuel depletion and the associated buildup of fission products. Effective delayed neutron frac-tlans for both cycles show a decrease with burnup.
j X2. 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 reference cycle.
- 5. 3.
Changes in Nuclear Design There were no relevant changes in core design between the reference and re-load cycles. The same calculational methods and design information were used to obtain the important nuclear design parameters. The only algnificant oper-ational procedure change from the reference cycle is the operation in a feed-and-bleed mode.
The reference cycle began operation in the rodded mode but was subsequently modified for operation in the feed-and-bleed mode.
There-fore, since nearly the entire reference cycle 4 was cperated in the feed-and-bleed mode, this is not actually a new mode of operation.
Table 5-1.
Oconee 1, Cycle 5 Physics Parameters "
Cycle 4 Cycle 5*
Cycle length, EFPD 292 330 Cycle burnup, SG'd/mtU 9,136 10.327 Average core burnup, toc, M'Jd/mtU 19,034 19,027 Initial core loading, etU 82.1 82.1 Crit ical boron, BOC (no Xe), pps IlZP, group 8 37.5% wd(d) 1415 145g liZP, groups 7 and 8 insertad 1335 1324 IIFP, group 8 inserted 1145 1276 Crit ical boron, EOC (eq Xe), ppm
!!ZP group 8 37.5% vd 373 343
!!FP, group 8 37.5% wd 88 44 Control rod worths, liFP, BOC, % Ak/k Group 6 1.07 1.21 Croup 7 0.93 1.45 Croup 8 37.5% wd 0.50 0.43 Control rad worths HFP, EOC, % Ak/k Group 7 1.16 1.53 Croup 8 37.5% wd 0.47 0.48 1
Babcock s.Wilcox 5-2 l
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l Table 5-1.
(Cont'd)
Cvele 4(b) Cycle 5
- Max ejected red worth, HZP, I ak/k(*
i BOC (N-12) 0.68 0.57 EOC (N-12) 0.61 0.70 Max stuck rod worth, HZP, % ak/k BOC (N-12) 1.74 2.17 EOC (N-12) 2.02 2.01 Power deficit, liZP to HFP, % Ak/k j
BOC 1.49 1.11 EOC 2.07 2.12 Doppler coef f,10 5(4k/k *F) l BCC, 100% power, no Xe
-1.45
-1.45 EOC, 100% power, eq Xe
-1.55
-1.62
)
Moderator coeff HFP, 10-4(ak/k *F) j BOC (0 Xe, crit 9pm, gp 8 ins)
-1.00
-0.45 COC (eq Xe, 17 ppm, gp 8 ins)
-2.55
-2.64 Boron worth, HFP, ppm /1 Ak/k BOC (1150 ppm) 109 109 EOC (17 ppm) 101 97 Xenon worth, llFP, % ak/k BOL (4 EFpD) 2.60 2.62 EOC (equilibrium) 2.61 2.73 Eff delayed neutron fraction. HFP BOC 0.00593 0.00593 EOC 0.00530 0.00521 I"} Cycle 5 data are for the conditions stated in this report.
The cycle 4 core conditions are identified in reference 4 Pased on 292 EFPD at 2568 MWt, cycle 3.
I* Cycle 5 data are based on a " planned" cycle 4 length of 235 EFPD; the cycle 4 " design" lifetime is 292 EFPD.
(
llZP denotes hot zero power (532F Tavg), HFP denotes hot full power (579F T vg)*
a I* Ejected rod worth for groups 5 through 8 inserted.
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t Table 5-2.
Shutdown Margin Calculation for Ocence 1. Cycle 5 BOC, % ik/k EOC, 7. Ak/k l
Available rod worth j
Total rod worth. HZr 8.91 8.79 4
Worth reduction due to burnup
-0.36
-0.
2 of poison material j
5 taxi =um stuck rod, HZP
-2.17
-2.01 Net worth 6.38 6.36
)
j 1.e s s 10% uncertainty
-0.64
-0.64 i
Total available worth 5.74 5.72 j
'temired rod worth Power deficit, HFP to HZP 1.31 2.12' k
Stax allowable inserted rod 0.40 0.60 J
worth j
Flux redistribution 0.59 1.20 Total required worth 2.30 3.92 1
}
Shutdotin margin (total available 3.44 1.80 I
worth minus total required worth)
Note: Required shutdown raargin is 1.00*. Ak/k.
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Ba'ocock & Wilcox 5-4 4
4 1
t w
-r-v4 u
+ - - - - -.,
-e---,.
w
--s
Figure 5-1.
BOC (4 EFPD). Cycle 5 Two-Dimensional Relative Power Distribution - Full Power, Equilibrit= Xenon, Norrul Rod Positions (Group 8 Inserted) 8 9
10 11 12 13 14 15 H
0.83 0.93 0.96 0.90 1.37 1.03 1.09 0.87 7
K 1.35 1.07 1.21 0.98 1.09 0.93 0.83 7
8 L
1.05 1.25 1.03 0.95 1.15 7.67 M
1.09 1.23 0.89 0.91 N
1.21 0.94 0.61 0
0.70 P
R Inserted Rod Group No.
x.xx Relative Power Density 5-5 Babcock s.Wilcox l
BLANK FRAME FOR
~
PROPER PAGINATION
6.
THERMAL-hfDRAUlIC DESIGN The thermal-hydraulic design evaluation so. sporting cycle 5 operation utilized the rathods and models described in references 3, 4, and 6.
The fresh batch 7 fuel is hydraulically and geometrically similar to batch 6 fuel.
The cycle 4 and 5 maximum design conditions and significant parameters are shown in Ta-ble 6-1.
The minimum DNSR shown at the desigr. overpewer is unchanged for cy-cle 5 and is based on 106.5% of RC design flow and on the Mark-B4 fuel assen-bly and includes the effects of incore fuel densification.
The potential ef fect of fuel rod bow on DNBR was considered by incorporating suitable margins into DNS-limited core safety limits and RPS setpoints. The maximum rod bow was calculated from the equation ff=0.065+0.001449/BU o
LC = rod bow magnicude, mils, Co = initial gap (138 mils),
BU = maximum assembly burnup,15'd/mtU.
The fuel cycle design calculations show that the maximum radial-local peak during cycle 5 is always located in the batch 7 fuel assembly with the maxi =um burnup. This maximum peak (1.527) is 17 below the 1.78 reference design peak.
Since this fuel assembly is limiting for DNBR analysis, the rod bow penal *ty associated with batch 7 is applied to cycle 5 operation. This method for cal-culating the maximum core rod bow penalty has been reviewed and approved for acceptability by the USNRC.0 The Oconee 1. cycle 5 calculated rod bow penalty is S.0; based on the maximum burnup in batch 7, 13,667 mwd /mtU. No credit is claimed for the difference between calculated cycle 5 peaking and the refer-ence design peaking used for the analysis. An 11.2% rod bow penalty is con-servatively applied to all analyses that define plant operating limits and to design transients.
Babcock & Wilcox 6-1
s Tae pressure-temperature limit curve shown in Figure 2.1-3A of the Oconee Technical Specifications provides the basis for the variable low-pressure trip I
setpoint.
The curves shown for four-and three-pu=a operatien represent a i
Iacus of points for which the calculated minteum DNsr. is equal to 1.30 (BAW-2)
I plus a suitable cargin to offset the DNBR reduction ove to red bow (discussed in the previous paragraph).
The flux / flow tri; setpoint was determined on the basis that the Oconee 1 plant has the pump nonitor trip function set to trip the reactor upon loss of one pump during four-pump operation if the indicated reactor power is greater than 80% of f ull power.9 The flux / flow trip setpoint of 1.055 established for cycle 5 yields a minimum DNBR of 1.68 and a 30% DNBR credit to offset the rod bow penalty.
Table 6-1.
Thermai-Hydraulic Design Conditions 4
Cy-le 4 Cycle 5 Power level, >Nt 2563 2568 System pressure. psia 22'0 2200 Peactor coolant flow, % design flow 106.5 106.5 Vessel inlet coolant temp, 555.6 555.6 100% power, F Vessel outlet coolant teep, 602.4 602.4 100% power, F Pef design radial-local power 1.733 1.783 peaking factor Ref design axial flux shape 1.5 cos 1.5 cos Active fuel length, in.
(a)
(a)
Average heat flux, 100% power, 176(b) 176(
103 Bto/h-ft2 CHF correlation Bri-2 RAW-2 Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flcw area 0.98 0.98 Minimum DNBR with densif'n penalty 1.91 1.91 I' See Table 4-2.
I Based on densified length of 140.3 inches.
6-2 Babcock & Wilcox
I I
7.
ACCII;ENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analysis Each FSAR3 accident analysis has been exanined with respect to changes in cy-cle 5 parameters to determine the ef fect of the cycle 5 reload and to ensure that thermal performance during hypothetical transients is not degraded.
f 5
The ef fects of fuel densification on the FSAR accident results have been eval-uated and are repnrted in reference 6.
Since batch 7 reload fuel assemblies contain fuel rods whose theoretical density is higher than those considered in the :lerence 6 report, the conclusions in that reference are still valid.
7.2.
Accident Evaluation The key para =cters that hate the greatest effect on deternining the outconc of a t rar.s i en t can typically be classified in three major areas: care thermal pa-rameters, 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 operat-ing values based on calculational values plus uncertainties. Fuel thermal analysis values for each batch in cycle 5 are compared in Table '.- 2. The cy-cle 5 thercal-hydraulic maxi =am design condi' sus are ecmpared to the previous cycle 4 values" in Table 6-1.
These parameters are cocnon to all the acci-dents considered in this report. A comparison of the key kinetics parameters from the FSAR and cycle 5 is provided in Table 7-1.
A generic LOCA analysis for the B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model. This study is re-ported in SAW-10103, Rev. 1.10 The snalysis in BAV-10103 is generic since the limiting values of key parameters for all plants in this category were used.
Furthermore, the combination of overage fuel temperature as a function of LHR and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for th's reload. Thus, the 7-1 Babcock & Wilcox
.1
W h
b.
2 4
analysis and the LOCA limits reported in EA-10103 provide conservative results for the operatien of Oconce 1 cycle 5 fuel.
Table 7-2 shows the bounding values for a110wable LOCA peak 61Rs for Oconee 1, cycle 5 fuel.
It is concluded f ro:n the exa=ination of cycle 5 core thereal and kinetics prop-ertles, with respect to acceptable previous cycle values, that this core re-I lead will not adversely af fect the Oconce 1 plant's ability to operate safely daring cycle 5.
Considering the previously accepted design basis used in the FSAR and subsequent cycles, tha transient evaluation of cycle 5 is considered to be bounded by previously ac(?pted analyses. The initial conditions for the transients in cycle 5 are bounded by the FSAR, the fuel densification report 6, 3
and/or subsequent cycle analyses.
4 i
i i
s Babcock &Wilcox l
7-2 i
4
i i
I i
Table 7-1.
Comparison of Key Parameters for Accident Analysis FSAR and Predicted
^densification cycle 5 Parameter report value value Doppler coef f, ak/k/*F BOC
-1.17 a 10-5
-1.45 x 10-5 EOC
- 1. 3 3 = 10- 5
- 1. 62 = 10- 5 Nderator coef f, Ak/k/*F BOC
+0.5 a 10 '
-0.45 = 10-'
roc
-3.0 x 10 "
-2.64 x 10-"
All-rod group worth, HZP Z ak/k 10 8.91 Initial boron conc'n, HFP, ppm 1400 1276 Boron reactivity worth at 70F, ppa /1% ak/k 75 76 Max ejected rod worth, HFP, %
ak/k 0.65 0.25 Dropped rod worth (HFP), %
ak/k O.46 0.20 Table 7-2.
LOCA Limits. Oconee 1. Cycle 5 i,
Elevation, LHR limits, ft kW/ft 2
15.5 4
16.6 6
18.0 i
8 17.0 10 16.0
)
7-3 Babcock & Wilcox
BLANK FRAME FOR l
PROPER PAGINATION
--.-----.----,-..--,,.,--.--,------,,,,-,------a--
- - - - ~ ~ - -,,
1 l
8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 5 operation. Changes were the results of the following:
1.
The Technical specification limits based on DNBR and LHR criteria include appropriate allowances for projected fuel rod bow penalties, i.e.,
poten-tial reduction in DNBR and increase in power peaks. A statistical combi-nation of the nuclear uncertainty factor, engineering hot channel factor, and rod bow peaking penalty was used in evaluating LHR criteria, as ap-proved in reference 11.
2.
Per reference 12, the power spike penalty due to fuel densification was not used in setting the DNBR-and ECCS-dependent Technical Specification licits.
3.
The allowable quadrant tilt limit for cycle 5 is 5.0".
Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated. Figures 8-1 through F-10 illustrate revisions to previous Technical Specification limits.
i
}
Babcock & Wilcox 8-1
,a
)
figure t-1.
Core Protection Satecy Limits. Oconce L' nit 1 Thermal Power Level, *.
- 120
(-28.1I2)
(32.1.12)
I ACCEPTABLE
(-40.98)
- - I 4 PUMP (50.98)
OPERATION (32.85.3) 2
(-28,85.3) 80 ACCEPTABLE 354 PUMP (50.71.3)
(-40.71.3)
OPERATION 3
-~
60 (32.58.2)
(-28,58.2)
ACCEPTABLE (50.44.2) 40 2.3&4 PUMP
(-40,44.2)
OPERATION 20 i
e i
-83
-60
-40
-20 0
20 40 60 80 Reactor Power imbalance, $
CURVE RC FLOW (GPM)
This is proposed new Tech-nical Specification 1
374,880 Figure 2.1-2A.
2 280,035 3
383,690 Babcock s. Wilcox 8-2
Figurre 8-2.
Protect Ive Syste:2 Maximum Allevable Setpoints, Oconee Unit 1 Thermal Power, %
(-18,105.5)
(20,105.5)
M2 = -0. 75 Mg = 0.568
,,g 4 PUNP l
(30.98)
(-40,93)
OPERATION I
e
(-18,78.8)
_ _ 80 (20.78.8)
=
3 E
z 3&4 PUMP l
(30,71.3) g
(-40.66.3)
OPERATION n.
y
(-18. 51,. 69)
- - 60 l
a.
l w
(20.51.,9) y m
6 b
b 8
I I
(30.44.19) 8 g
(-40.39.19) l
- - 40 g
i i
4 I
l UIz I
-- 20 o l o
o o
bs0 e a l,,
,, l M n e,,t e,
E, E
i i
-60
-40
-20 0
20 40 60 Reactor Power imbalance. ?,
This is proposed new Technical Scecification Figure 2.3-2A.
Babcock & Wilcox 8-3
i
?! cure 8-3.
Rod Position L1=1ts for Pour-Purp 0peration.
Oconee Unit I to to 100 : 16 EFPD)
(125.102)
(274.1.102)
O 100.
CPER ATf 3 (278.8.92) 40T ALLSID Postt 60 (2s8.2.80) g, st$1tKTED
$~;TD0mu 14ARGin LIMIT agGion to (214.7.60) f (70.50) 3
'.0
}
PEht t $$l8LE C'1RATileG
,,5) 20 sEGION (O.42.5' (0.0) i et 50 1to a50
- CO 250 J T.
25 50 75 s o,o 0,
25, 5,0 i
i Grcsa 5 Group 7 1
i 0
25 50 75 400 j
Group 6 Tnis is proposed new Technical Specification Figure 3.5.L-1A1.
8-e.
Babcock s.Wilcox
Figure 8-4 Rod Position Limits for Four-Pump Operation.
Oconee t* nit 1 (After 100 1 10 EFPD)
(2'41. 0 C2 )
(274.e.lC4 U
800 OPIRAfton (274.1.92) 40T Altowf3 248.2.80)
LEVEL 90 CUTOFF
- 92T F P 4(214.7.60) 60 j
m n
(179.50)
I PERH15SISLC
=.
40 CPenafi4G SMU100wn 84ARGlu REGION g
LIMIT -
=
20 (403.15)
(0.6.8) 0 r(0.C1,
0 50 IOC 650 200 250 300 0
25 50 75 0
25 50 75 800 0
i f
f 1
E f
f Gro.o 5 Group 7 75 400 0
2,5 5,:
i Gro.o 6 This is proposed new Technical Specification Figure 3.5.2-1A2.
8-5 Babcock & Wilcox
I 4
Figure a-5.
Rod Position 1.f mits for Two-and Three-Pu=p Operation. Ocence Unit 1 (0 to 100 : 10 EFPD)
(125.102)
( t H.102 (265.2.102)
ICO OPltAttom NOT AttCtetD g
//
(214.7.76)
(120.61) t g
.E
~ 5"'JTDOWN MARGiu 8
tl4ti N
A; (70.50) r j 40 C PER AT ING PisW15518LE
,]
REGION t*
(16.15) 1 20
- h. t2.5 9 t
i f
f f
I f
0g O
$3 100 150 200 250 300 Rod indes. % WitNrawn 0
25 50 75 1,00 C
25 5,0 7,5 Ip Group 5 Gro p 7 0
2,5 50 75 100 G*owp 6 This is proposed new Technical Specification Figure 3.5.2-2A1.
8-6 Babcock & Wilcox
Figure 8-6.
Rod Position Limits for Twr and Three-Pump Operation, Oconee Unit 1 (After 100 : 20 EFPD)
(2.s.lc2)
(26s.2.102)
OPEauton 100 -
NOT ALLowEO eESTRICTED FOR 3 PUMP 60 g 214.7.76 )
&{ 60 (l79.50)
PERul5slBLE OPERAftoG
^
$MUT;,0wn etARGlu LIMIT REGICN
". to t
E
~
(103.15)
(0.G.I)
O t
i i
e t
t t
t i
'50 200 210 300 1
O 50 400
. mitadra i Rod leden.
0 25 50 75 100 0
25 50 75 100 t
t t
1 i
i 1
I I
Group 5 Gro.c ?
O 25 50 75 600 t
I f
I Gro.a 6 This is proposed new Technical Specification Figare 3.5.2-2A2.
Babcock a.Wilcox 8-7
Figure 8-7.
Pcwer Imbalance Limits. 0:enee L~ nit 1 (0 to 100 10 EFPD)
Power. % of 2568 mrt
" ~ ' '
RESTRICTE0 REGION (15.9.102)
(-26.9.102),
(-28.2.92)
- - 90 (15.9.80) 80
(-29.2.80) 70 PERMISSIBLE -.
60 OPERATING REGION 50 40 30 20 10 I
t I
f f
f 3
1 g
-50
-40
-30
-20
-10 0
+10
+20
+30
'40
+50 Axial Power imbalance. %
This is proposed new Techr.ical Sp'ecification Figure 3.5.2-3A1.
l l
l 8-8 Babcock & Wilcox l
Figure 8-8.
Power Imbalance Limits, Oconee Unit 1 (After 100 : 10 EFPD)
Power, % of 2568 MWt
.-110 RESTRICTED REGION
(-28.0.102)
(16.4.102) r
.-100
(-26.5.92)
- - 90 4
(-29.2,80)<
- 80
' '(16.4,80)
PERMISSIBLE
- - 70 OPERATING REGION
-- 60
- - 50
' 40
- ' 30
- - 20
=- !O
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power imbalance, %
This is proposed new Tech 9ical Specification Figure 3.5.2-3A2.
l s-9 Babcock s. Wilcox
Tigure 8-9 A?SR Position 1.imits. Oconee Unit 1 (Fron 0 to 100 : 10 IFPD)
(16.3.102)
(45.0.102) 100 f
RESTRICTED (45.0,92)
(14.1.92) 80 (0.0.80)
(100.60) a f
60 SN PERMISSIBLE s$
OPERATING 40 REGION c
I2 20 This is proposed r.ew Technical S ecification Figure 3.5.2-4A1.
0 O
20 40 60 80 100 Bank 8 Position, % Withdrawn i
i 8-10 Babcock 8.Wilcox
Ti:;ure 6-10.
APSR Positior: 1.1=its, Oconee Unit 1 (After 100 : 10 EFP3)
(9.3.102)
(45.0.102)
I RESTRICTED
~
REGION (7.6,92)
(45.0.92) 80 (0.0.80)
(57.9.80) 100.60) a N
60 o
k PERMISSIBLE E
OPERATING J
'(
REGION b
40 l
20 This is proposed new Technical Specification Figure 3.5.2-4A2.
0 e
i i
0 20 40 60 80 100 Bank 8 Position, % Withdrawn 8-11 Babcock a. Wilcox
4 i
9.
START'.T PROGRA.'t - PliYSICS TESTIS; i
The planned startup test progr.la associated with core performance is outlined belo. These tests verify that core perfornance is within the assumptiens of the safety analysis and previde the necessary data for continued safe opera-tion.
a "reeritical Tests i
1.
Cont rol rad trip test.
Xero Power Pntstes Tests i
1.
Critical baron eeneentration.
i
.!. Te::Terature eactivity coefficient.
a.
All rods out, group S in.
i b.
Groups 5 through 8 inserted, groups 1 through 4 out.
1.
Control rod groun reactivtty worth.
4.
1:jected control rod reactivity worth.
I Power le.ets I.
Core po.cr di.stribut ion scrification at approximatelv !.0, 75, and 100?.
tull parr with normal control rod group configuration.
2.
in. ore strsus out-of-core detector imbalance correlation verificatien 4
l at ~ less t h.an f ull power.
I 3.
Powsr Deppler reactivity coefficient at approximately 100% full power.
4.
Temperature reactivity coef ficient at approximately 1007. full power.
i a
=w-
REFERENCES I"
Oconee 1. Cycle 4 Quadrant Flux Tilt. BAW-1477, Babcock & Wilcox, Lynchburg, Virginia, January 1978.
2 A. C. Thies (Duke Power Co.) to Edson G. Case (USNRC). Le tter, October 26, 19'17. Docket No. 50-269.
3 Oconee Nuclear Station, Units 1, 2, and 3 -- Final Safety Analysie Reports, Docket Nos. 50-269, 50-270, and 50-287, Duke Power Company.
4 Oconec Unit 1 Cycle 4 Reload Report. BAW-1447, Babcock & Wilcox, Lynchburg, Virginia, March 1977.
Program to Determine In-Reactor Performance vi B&W Fuels -- Cladding Creep Collapse, BAW-10084, Rev.
1, Babcock & Wilcox, Lynchburg, Virginia, Decem-her 1976.
6 Oconee 1 Fuel Densification Report, BAW-13SS, Rev. 1. Babcock & Wilcox, Lynchburg, Virginia, July 1973.
7 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Cas Pressure Analysis, BAW-10044, Babcock & Wilcox Lynchburg, Virginia, May 1972.
8 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment 14 to Facility operating License No. DPR-54, Rancho Seco Nuclear Cencration Station, Sacramento Municipal Utility District, Docket No.
50-312.
9 Duke Power Company to E. C. Case (Acting Director, Of fice of Nuclear Reac-tor Regulation), Letter, " Revision of Oconee Nuclear Station Tech. Spec. to Modity Pump Monitor Trip Setpoint," Septecher 14, 1977.
10 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 1 Babcock
& Wilcox, Lynchburg, Virginia, September 1975.
Babcock & Wilcox A-1
'?
~
,i S. A. Varga (USNRC) t o J. 11. Tayler (B5W), Letter, "Coments en 35W's sub-nit tal on Conbination of Pe.n., ng Fac tors," May 13, 1977.
I'S.A.
Va r g,.i (USSRC) to J. H. Taylor (B5W), Letter, " Update of BAW-10055 -
Fue ! Dens i f icat ion Report," L'ecember 5, 1977.
1 r
l 1
A-2 Babcock 8. Wdcox