ML20071K411
| ML20071K411 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/28/1983 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML15223A892 | List: |
| References | |
| BAW-1774, NUDOCS 8305270309 | |
| Download: ML20071K411 (46) | |
Text
. .
3A'a*-177 /*
February 1923 OCONIE UNIT 1. CTCLE S
- Relcad Repor -
8305270309 830519 ' Babcock & Wilcox PDR ADDCK 05000269 '"d"**""""
P PDR
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, 3Ak*-177/*
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. February 1983 I
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OCONEZ UNIT 1. CTCLE 3
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i - Reload Report -
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BA3 COCK & k'ILCOI Utility Power Generation Divisier.
P. O. Box 1260 Lynchburg Virginia 24505 I
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CONTENTS Page
- 1. INTRODUCTICN AND SUMMART . . ................... 1-1
- 2. OPERATING HISTORT ...................... . . 2-1
- 3. GENERAL DESCRIPTION ....................... 3-1
- 4. 7UEL SYSTEM DESIGN . . . . .................... 4-1 4.1. Tual Assembly Mechanical Desip .............. 4-1 4.2. Fuel 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 4.3. Thermal Design . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4. Material Design ...................... 4-3 4.5. Operating Experience . .................. . 4-3
- 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. THIRMAI.-EDRACLIC DESIGN . . . . .......... . .... . . 6-1 i
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- 7. ACCIDENT AND TRANS W ANALYSIS ....... ..... .. ... 7-1 7.1. General Safety Analysis .... ..... .... . .... 7-1 l
7.2. Accident Evaluation .................... 7-1 l
, 8. PROPOSED MODIFICATIONS TO TICENICAL SPECIFICATIONS . . . . . . . . 8-1
- 9. RI? m'NCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 l
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List of Tables Table Page 4-1. Fuel Design Parameters and Dimensions . . . . . .. . . .... 4-5 4-2. Puel Thermal Analysis Parameters - Oconee 1. Cycle 8 . .... 4-6 5-1. Oconee 1 Physics Parameters . . . . . . . ... .. . . .... 5-3 5-2. Shutdown Margin Calculation for oconee 1 Cycla 8 . . . . . . . 5-5 6-1. Thermal-Hydraulic Design Conditions . . . . . . . . . . . . . . 6-2 7-1. Comparison of Key Parameters for Accident Analysis . . .... 7-3 7-2. LOCA Limits, oconee 1, Cycle 8. After 50 ETPD . . . . . .... 7-3 7-3. LOCA Limits, Oconee 1, Cycle 8, 0-50 IFPD . . . . . . . . ... 7-4 7-4. Comparison of FSAR and Cycle 8 Accident Doses . . .. . .... 7-5 List of Figures Figure 3-1. Core Leading Diagram for Oconee 1. Cycle 8 . . . . . . .... 3-3 3-2. Enrichment and Burnup Distribution for Oconee 1, Cycle 8 ... 3-4 3-3. Control Rod Locations for Oconee 1. Cycle 8 . . . . . . . . . . 3-5 3-4. 3PRA Concentratf.on and Distribution for oconee 1, Cycle 8 . . . 3-6 5-1. Oconee 1, Cycle 8 30C Tw-Dimensional Relative Power Distribution - Full Power Equilibrium Xenon, Normal Rod Positions . ........................ 5-6 8-1. Core Protection Safety Limits for Ocense Unit 1 . . . . . . . . 8-2 8-2. Protective System Maximum Allowable Satpoints for Oconee Unit 1 ............................ 8-3 8-3. Rod Position Ltnits for Four-Pump Operation, 0-50 EFPD, Oconee 1 Cycle 8 . . . . . . . . . . . . . . . . . . . . . . . 8-4 8-4. Rod Position Limits for Four-Pump Operation After 50 EFFD, Oconee 1. Cycle 8 . . . . . . . . . . . . . . . . . . . . . . . 8-5 8-5. Rod Position Limits for Three-Pumo Operation, 0-50 EFPD, Oconee 1, Cycle 8 . . . . . . . . . . . . . . . . . . . . . . . 8-6 8-6. Rod Position Limits for Three-Pump Operation After 50 ETPD, Oconee 1. Cycle 8 . . . . . . . . . . . . . . . . . . . . . . . 8-7 8-7. Rod Position Limits for Two-Pump Operation, 0-50 IFFD.
Oconee 1. Cycle 8 . . . . . . . . . . . . . . . . . . . . . . . 8-8 8-8. Rod Position Limits for tro-Pu=p Operation After 50 EF7D.
Oconee 1 Cycle 8 . . . . . . . . . . . . . . . . . . . .... 8-9 8-9. Power Imbalance Limits for 0-50 ETPD, Oconee 1 Cycle 8 . . . . 8-10 8 -10. Power Imbalance Limits After 50 ETPD, Oconee 1, Cycle 8 . . . . 8-11 8-11. APSR Position Limits for 0-50 EF?D, Oconee 1 Cycle 8 . . . . . 8-12 8 - 12. APSR Position Limits After 50 EFFD, Oconee 1. Cycle 8 . . . . . 8-13
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- 1. INTRODUCTION AND SDefARY This report justifies the operation of the eighth cycle of Oconee Nuclear Station, Unit 1, at the rated core power of 2568 MWt. Included are the re-
. quired analyses as oytlined in the USNRC document " Guidance for Proposed License Anendments Relating to Refueling," June 1975.
To support cycle 8 operation of Oconee 1, this report employs analytical tech-niques and design bases established in reports that have been submitted to and accapted by the USNR0 and its predecessor (see references).
A brief summary of cycle 7 and 3 reactor parameters related to power capability is included in section 5 of this report. All of the accidents analyzed in the FSAR I have been reviewed for cycle 8 operation. In those cases where cycle 8 characteristics were conserweive' compared to those analyzed for previous cy-cles, no new accident analyse. vere performed.
Five of the fresh batch 10 asseublies are gadolinia lead test assemblies (LTA).
These assemblies are part of a joint Duke Power /3abcock & 'Jilcox (3&*J)/ Depart ,
ment of Energy program to develop and demonstrate an advanced fuel assembly design incorporating UO2 -Gd2 03 for extended burnup in 7WRs. Reference 2 de-scribes the L"As. your Mark 3Z demonstration fuel assemblies containing Zir-l caloy-4 intermediate spacer grids will be reinserted for a second cycle of irradiation. The Mark 3Z assemblies are described in reference 3. The gado-linia LTAs and the Mark 3Z assemblies will not adversely affect cycle 8 opera-l tion.
l The Technical Specifications have been reviewed, and the modifications required for cycle 8 operation are justified in this report.
3ased on the analyses performed, which account for the postulated effects of fuel densification and the final acceptance criteria for emergency core cooling syste=s, it has been concluded that Oconee Unit 1 can be operated safely for' cycle 8 at the rated power level of 2568 MWt.
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- 2. OPERATING HISTORY The reference fuel cycle for the nuclear and ther:nal-hydraulic analyses of Oconee 1, cycle S,is the currently operating cycle 7. The cycle 8 design length of 410 EFFD is based on a planned cycle 7 length of 4 0 EFFD. No op-erating anomalias have occurred during previous cycle operations that would adversely affect fuel performance in cycle S.
2-1 Babcock & Wilcox
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- 3. GENERAL DESCRI?TICN The Oconee Unit i reactor core and fuel design basis are described in detail in section 3 of the Final Safety Analysis Report for Ococee Nuclear Station.
Unit 1.1 The cycle 8 core contains 177-fuel assemblies, eac's of which is a 15-by-15 array of 208 fuel rods,16 control rod guide tubes, arid one 1 score instrument guide tube. The fuel consists of dished-end, cylisc'rical pellets ,
of uranium dioxide clad in cold-worked Zircalcy-4. ne stand. Mark 3 feel assemblies in all bacches have an average fuel loading of 463. '3 of uranium.
The undensified active fuel lengths, theoretical densi':iss, fv41 and fuel rod dimensions, and other related fuel parameters are g:,vii la Tables 4-1 and 4-2.
Figure 3-1 is the core leading diagram for Oconee 1, cycle 8. The twent.7 batch SA and four of the batch 83 assemblies vill be discharged at the end of cycle 7, along with forty batch 73 and one batch 4E assembly. The rema%1ng 44 batch 8 assemblies (designated SC) and the fresh batch 10C - with initic.1 enrichments of 3.07 and 3.41 we : 2s sU respectively - will be loaded into the central portion of the core. There are also five fresh batch 10 gado11sia LTAs 2 in the core interior. The center assembly, batch 10A, has an initial enrichment of 2.46 vt : 2 sC; the four batch 103 LTAs, with an initial enrich-ment of 4.00 vt : 2 sU are in locations symmetrical to H13. The batch 9 fuel, with an initial enrichment of 3.28 we : a ssU, will mainly occupy the core pe-riphery. Figure 3-2 is an eighth-core map showing the assembly burnup and en-richment distribution at the beginning of cycle 8.
Reactivity is controlled by 51 full-length Ag-In-Cd control rods, 60 burnable i poison red assemblies (3PRAs), and soluble boren shim. In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided fer additional control of the axial power distribution. The cycle 8 locations of the 69 control rods and the group designations are indicated in Figure 3-3.
The core locations are identical to those of the reference cycle. *he cycle 8 locacicus and concentrations of the 3PRAs are shown in Figure 3-4 3-1 Babcock & Wilcox
S.e sys:e= ;; essure is 2;;C ;sia a:d :he ::re aterage densified == 1:a1 hea:
ra:e is 5.30 W/f: a: :he ra:ed ;cve: of 2553 M::: fe the s:ardard Mark 3 fuel asse=blies.
3-2 Babcock & Wilcox
Figure 3-1. Core Leading Diagram for Oconee 1. Cycle 8 I
I A
M4 K2 CS K14 M12 9 9 9**' 9 9 L3 N3 v' V8 M14 N13 L13 g
9 9 9 10C 9 10C 9 9 9 E9 L5 36 310 L11 K3 O
9 IOC 9 10C 8C 10B* SC 10C 9 10C 9 C10 N9 R10 P8 Re K4 C6 0 9 10C 9 10C SC 10C 10C SC ICC SC 9 10C 9 C12 E10 P12 R9 R7 N2 E6 CA I 9 9 10C SC IOC SC 10C 3C 10C SC 10C 9 9 D11 311 L15 013 R8 03 L1 35 D5 F
9 9 IOC 8C 10C 8C IOC 8C IOC 8C 10C BC 10C 9 9 39 F2 K15 N14 D9 P4 K1 F14 37 8 10C SC 10C 9 SC 10C 8C 9 SC IOC SC 10C 8C 10C 9 H f -- H3 Hil H14 H15 G4 K12 El H2 H5 H13 _. y o** O 103* BC IOC 9c 4 10A* 0 9C 10C 90 103* O C**
P9 L2 G15 312 N7 D2 G1 L14 P7 K 10C 9 SC IOC 3C IOC 8C 9 8C IOC 8C IOC SC IOC 9 N11 P11 8 F15 C13 Ad C3 F1 P5 N5 L 9 9 l 10C 3C 10C 8C 10C SC 10C SC 10C SC 10C 9 9 012 M10 D14 A9 A7 34 Mo 04 4 9 9 10C 8C 10C 8C 10C 10C SC SC 10C 9 9 010 G12 A10' 38 A6 DT C6 N 9 10C 9 10C SC 100 SC IOC SC 10C 9 10C 9' G8 F5 P6 P10 711 H7 0 9 10C 9 10C SC 103* SC 10C 9 IOC 9 T3 D3 E2 ES E14 D13 713 P o o o IOC 9 10C 9 o o E4 G2 08 G14 E12 N
9 9 9** 9 9 l
Z
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1 2 3 4 5 6 7 8 9 10 11 12 13 14 15
- Contains gadolinia in 12 pins.
l ** Mark 32 demons::ation asse=blies l III CY7 Location i i
XIX 3atch
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Tii;;re 3-2. I:ri:h:e:: a:d 3 :: p Ois:ribu:i:= f:: 0 ::ee 1, 2y:2e !
3 3 13 11 12 13 la 15 2.46 3.23 3.07 3.41 3.07 4.30 3.23 3.25 H LTA 'JA Mark-32 0 17,266 22,019 0 27,420 0 17,007 16,310 3.07 3.41 3.07 3.41 3.07 3.41 3.25
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13,675 0 21,075 0 22.331 0 14,173 3.07 3.41 3.07 3.41 3.28 3.23 22,138 0 20,341 ' O 11,538 16,349 y 3.07 3.a1 3.28 3.23 15,687 0 17,352 13,121 3.23 3.41 3.23 N
17,269 0 16,257 3'2S 0
15,976
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R I. II Ini:ial E=rf h e::
IIIII 30C Sur:up, Mb'd/ :U l
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Figure 3-3. Control Rod Locations for Oconee 1. Cycle 3 X
A 4 7 4 3
2 6 6 2 C
8 5 s 7 7
0 1 1 5 2 2 5 E
7 3 a 4 F 4 8 3 1 3 3 1 6 G 6 '
5 7 -Y i H W- 7 5 7 4 7 3 3 1 6 6 1 X
7 3 8 4 4 3 3 L
l 2 5 1 l1 5 l2 M
8 5 8 7 7
N 2 6 6 2 0 l l 4 7 :.
F I I I R I I II I
I 9 10 11 12 13 14 15 3 4 5 5 7 8 1 2 GROUP NO. OF R003 FUNCTION 1 8 SAFETY 2 8 SAFETY X GROUP NUMBER 3 8 SAFETY 4 9 SAFETY 5 8 CONTROL 6 8 CONTROL 7 12 CONTROL 8 9 AP SRs TOTAL SS 3-5 Babcock & Wilcox
l ip Pigure 3-4 3PRA Concen::stion and 31s::1bu:1:n for Oc: nee 1, Cycle 5 j 3 9 10 11 12 13 14 15 I
I,;A 1.10 I.!A g
K 1.40 1.10 0.20 L 1.40 1.40 0.50 M 1.10 1.40 1.40 i
N 1.10 1.40 0.20 0 LA 0.50 0.20
=
E 0.20 R
X,XX 3PRA Concentration, vc : 3.C
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- 4. FL'EL SYSTIM DESIGN 4.1. Fuel Assembiv Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for Oconee 1, cycle 3 are lis'ted in Table 4-1. All the fuel assemblies are mechanically interchangeable. Tour once-burned Mark 32 fuel assemblies are included in batch 95. The Mk 3Z uses Zircaloy as the material for the six intermediate spacer grids. The Mark 3Z assembly is dese 1 bed in reference 3, which de.non-strates that reacecr safety and performance are not adverseiy affected by ths-presence of the four demonstration assemblies.
The five fuel assemblies in batches 10A and 103 are gadolinia LTAs. The me-chanical design of the LTAs is described in reference 2. -
Retainer asse=blies will be used on the two fuel assemblies that contain re-generative neutron source (355) assemblies and on the 60 batch 10C assemblies that contain 3PRAs. The justification for the design and use of the retainers is described in references 4 and 5.
4.2. Fuel Rod Design The mechanical evaluation of the fuel rod is discussed below.
4.2.1. Cladding Collapse The fuel assemblies of batch SC are more limiting than those of other batches because of their longer previous incere exposure time. The power history and fuel design parameters for the most limiting batch SC fuel assembly were com-pared with those used in the generic Mk-B creep collapse analysis and were found to be enveloped. The generic analysis was based on the methods and pro-cedures described in reference 6 and is applicable to the batch 8C fuel design.
The generic analysis predicts a collapse time of more than 35,000 E7PH, which exceeds the maximum projected residence time of 29112 E7?H (Table 4-1).
4-1 Babcock & Wilcox
l A detailed creep analysis was perfor=ed en the gadolinia bearing fuel rods in the LTAs. The collapse time for these rods was greater than the max 1=um pro-jected residence time.
4.2.2. Cladding Stress The stress parameters for the Oconee i standard fuel rods and the gadolinia bearing fuel rods are enveloped by a conservative fuel rod stress analysis.
The following four assumptions were used in this analysis:
- 1. A lower post-densification internal pressure.
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- 2. A lower initial pellet density. " '
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- 3. A bigher system pressure.
- 4. A higher thermal gradient acrees the cladding.
For design evaluation, the primary membrane stress must be less than two-thirds of the mini:mun specified unirradiated yield strength, and all stresses (primary and secondary) must be less then the sini=us specified unirradiated yield strength. In all cases, tne margin is in excess of 30%.
4.2.3. Cladding Strain '
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The fuel design criteria specify that the cladding average circumferential strain is not to exceed 1.0: inelastic strain. The pellet design is estab-lished for plastic cladding strain of less than 1% at maximum design local pellet burnup and heat generation race values that are higher than the values the Oconee 1 CO 2 fuel is expected to see. Strain analysis of the gadolinia fuel showed that the calculated strains for these rods are also below design limits.
i Thus, fuel red cladding strain will not affect cycle 8 fuel perfor-mance.
4.3. Thermal Desien
{ All fuel in the cycle 8 cora is ther= ally similar except the five LIAs. The fresh batch 10C fuel insertad for cycle 8 operation introduces no significant differences in fuel thermal performance ralative to the fuel remaining in the core. The fresh batch 10A and 10B fuel containing the gadolinia LIA demonstra-tien assemblies have different fuel performance characteristics, but are not
! more limiting than the re=ainder of the' core.
The cycle 8 thermal analyses represent a change in analytical method in that the fresh batches of fuel have been analyzed with the IACO27 code using the i
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nethodology described in" reference S. As shown in Table 4-2, the analysis e
utas necinal undensif:.ed input para =eters. The TACO 2 code densification model accounts for densification effects. TACO 2 analyses also apply to reinserted I
batch SC and 9 inel sface this fuel is identical in design to batch 10C.
Results of the thermal design evaluation for .he cycle 8 core are summarized in Table 4-2. The TACO 2 fuel performance code was used to determi=e linear heat rate to melt capabilities for batch SC, 9, and 10 fuel (95*. TD nomiaal initial density). Maximum linear heat race to centerline =elt was determined as a function of fuel burnup. The lowest maximum linear heat race was 20.5 kW/ f t for 8C, 9, and 10C bacches of fuel. The lowest maximum linear heat race for the batch 10A and 10B LTA gadolinia fuel i$ 17.6 kW/ft. ,
The maximum fuel rod burnup at EOC 8 is predicted to be 40,233 MEd/stU. Fuel
, t red internal pressure was evaluated with the TACO 2 computer code for thc/aigh-est burnup fuel rod and is. predicted to be less than the ne=inal RC systa=-
pressure of 2000 psia. ,
e 4.4 Material Design The batch 10 fuel asse=blies are not new 'in concept, nor do they still:e dif-ferent component.=aterials, except for the Zircaloy grids of the four Mark 3Z asse=blies and the UO. -Gd.0, pellets in the LTAs. Thereftre, the chemical co=-
patibility of all possible fuel-cladding--coolant-assembly in5eractions for the batch 10 fue14 assemblies is acceptable.
4.5. Oseratine Exeerience / ,sy ,
, t 36W operating experience with thr4 Mark 3'15-by-15 fuel assembly has verified the adequacy cf its design. As of October 31,,1982, the following experience
., i has been accumu]Jeed for i.he eight operattag S&W 177-fuel assembly plants us-ing the Mark B indi assembly:
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Max assenbiv * **i"* ***
beman (a) . Wd'/n=" electrical cu ;u:,(d"', J Current Reactor evele Incore Discharged MWh Oconte 1 7 44,850 40,000 38,723,077 ccones 2 6 23,750 36,300 34,354,735 Oconee 3 7 20,200 35,450 36,772,920 TMI-1 5 25,000 32,400 23,840,053 ANO-1 5 36,429 33,220 32,834,786 Kancho Seco 5 35,821 37,730 28,636,196 Crystal River 3 4 24,360 29,900 19,803.456 Davis Besse 3 25,742 25,326 12,021,378
(*)As of October 31, 1982.
)As of May 31, 1982.
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Table 4-1. Fuel Desirn Partneters and Dimensions Batch 3atch 3 acch 8C 9 10A/10B/10C EA type Mark 34 Mark 34/ Mark Gd3/ Mark Gd3/
Mark 3Z Mark 34
,No. of FAs 44 64/4 1/4/60 Fuel rod CD, in. 0.430 0.430 0.430 Fuel red ID, in. 0.377 0.377 0.377 Flex spacers, type Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Undensified active fuel length 141.38 141.8 141.8/143.5/141.8 (nominal), in.
Fuel pellet initial density 95 95 95 (nominal), TD Fuel pellac CD (mean specifi- 0.3686 0.3686 0.3686 cation), in.
Initial fuel enrichment. 3.07 3.28 2.46/4.0/3.41 wt : 235 U 30C burnup (avg), mwd /=tU 21,602 15,488 0 Cladding collapse ti=e, EFFH > 35 ,000 >35,000 >35,000 Esti=ated residence cine, 29,112 19,920 10,080
(=ax), ETPE l
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l Table 4-2. Fuel Thernal Analvsis Para =eters -- Oconee 1. Cvele 3 3atch 8C 9(*) 10A( ) 10B(*) 10C No. of assemblies 44 68 1 4 60 Nominal pellac density : TD 95 95 95 95 95 Pellac diamacer, in. 0.3686 0.3686 0.3686 0.3686 0.3686 Stack height, in. 141.38 141.8 141.8 143.5 141.8 Nominal LER @ 2568 MWe kW/f: 5.76 5.74 5.74 5.68 5.74 LHR to G, fuel melt, kW/ft 20.5 20.5 17.6(d) 17.6(d) 20.5 (a) Includes fcur Mark 3Z demonstration assenblies.
One gado11sia LTA.
(")Four gadolinia LTAs.
(d)Cadolinia bearing rods culy. Uraniun rods have li=1:s 1 20.5 kW/ft.
4-6 Babcock & Wilcox
- 5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of design cycle S vich those of the reference cycle 7. The values for both cycles were generated using Because of its shorter length, the average cycle 8 burnup will
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FDQ07'~11 be lower than that of design cycle 7. Figure 5-1 illustrates a representa-tive relative power distribution for the begi=ning of cycle 8 at full power with equilibrium xenon and normal red positions.
Since the core has not yet reached an equilibrium cyc1S, differences in core physics parameters are to be expected between the cycles. The critical boron concentrations for cycle 3 are lower because of the shorter cycle length and the higher average burnable poison enrichment. The control red worths differ between cycles due to changes in radial flux and burnup distributions. Tc.is also accounts for the larger ejected and stuck rod worths in cycle 8 compared to cycle 7 values. Calculated ejected rod worths and their adherence to cri-teria are considered at all times in life and at all power levels in the de-velopment of the rod position limits presented in sectica 3. These red worths
=eet all safety criteria. The adequacy of the shutdown nargin with cycle 3 stuck red worths is demonstrated in Table 3-2. The following assu=ptions were applied for the shutdown calculations:
- 1. poison =aterial depletion allevance.
- 2. 10: uncertainty on net red worth.
- 3. Flux redistribution penalty.
Flux redistribution was accounted for since the shutdown analysis was calcu-laced using a two-dimensional model. The reference fuel cycle shutdown margin is presented in the reload report for Oconee 1, cycle 7 11 The cycle 8 power deficits, differential boren worths, and effective delayed centren fractions differ from those for cycle 7 because of the shorter cycle length and lower critical boren concentrations.
5-1 Babcock a.Wilcox
5.2. Analvtical Inou:
The constants used to compute core power distributions from incere detector measurements were obtained in the same nanner for cycle 8 as for the reference cycle 7. The monitoring of power discributions in the LTAs is discussed in reference 2.
5.3. Changes in Nuclear Design There are five fresh fuel assemblies, with 12 gadolinia fuel pins each, that are fully described in reference 2. Their effect on the core design is not significant, because the cycle 8 design nests all criteria including thosa applicable to radial power peaking, ejected rod' worths, and shutdown nargin.
5-2 Babcock & Wilcox
Table 5-1. Oconee 1 Physics ?aramecers(*)
Cvela 7(b) Cvele 8(*)
Cycle length, EFFD 427 410 Cycle burnup, mwd /scU 13,363 12,858 Average core burnup, ECC, Mud /scU 22,505 24,183 Initial core loading, scU 82.1 82.1 1
Critical boron, 30C (no Xe), ppm EZ?(d), group 8 inserted 1628 1602 EF? (d) , group 8 inserted 1464 1365 Critical boren, EOC (eq Xe), ppm EZP, group 8 inserted 380 401 EFP, group 8 inserted 68 60 Control rod worths, EF?, 30C, ik/k Group 6 0.97 0.98 Group 7 1.45 1.47 Group 8 0.47 0.42 Control rod worrhs, EFP, ECC, I ik/k Group 7 1.54 1.54 Group 8 0.53 0.49 Max ej ected red worth, EZ?, I ik/k(*}
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30C (N-12) 0.55 0.59 EOC (N-12) 0.62 0.48 Max stuck rod worth, HZ?, ik/k 30C (N-12) 1.44 1.68 ECC (N-12) 1.65 1.72 Power deficit, EZP to EF?, ik/k 30C 1.35 1.62 ECC 2.25 2.36 Doppler, coeff., 10~8 (ik/k/ 7) 30C, 100% power, no Ie -1.52 -1.54 EOC, 100% power, eq Ia -1.62 -1.78 Moderator coeff, ETP,10~' (ik/k/F) 30C (O Ie, crit pps, gp 8 ins) -0.48 -0.67 ZOC (eq Ie,17 ppm, gp 8 ins) -2.87 -2.35 Boron worth, EF?. pp:/* ik/k 30C (1300 pps) 123 129 ECC (17 ppm) 105 110 5-3 Babcock & Wilcox
Table 5-1. (Conc ' d)
Cvele 7(b) Cycle SI ")
Ienen worth, EFP, 2 ak/k BOC (4 EFPD) 2.58 2.54 ECC (equilibrium) 2.74 2.67 Eff delayed neutron fraction EF?
30C 0.00626 0.00625 ECC 0.00520 0.00526
(*) Cycle 8 data are for the condi: ions stated in this report. The cycle 7 core conditions are identified in reference 12.
O)3ased on 372 EFFD at 2568 MWe, cycle 6.
(*) Cycle 8 data are based on a cycle 7 length of 420 EFFD.
(d)EZP denoces hoe zero power (5327 T ) ; EF? denotes hoc full
- 3 power (579F T**3).
(e) Ejected rod worth for groups 5 through 8 inserted.
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o .
Table 5-2. Shutdown Margin Calculation for oconee 1. Cvele 8 30C. I ak/k EOC, I ak/k Available Rod Worth Total rod worth, HI? 8.53 9.19 Worth reduction due to burnup of poison material -0.42 -0.42 Maximum stuck rod, 3:7 -1.68 -1.72 Nec worth 6.43 7.05 Less 10% uncertaincy -0.64 -0.71 Total available worth 5.79 6.34 Recuired Rod Worth
?over deficit, HF7 to HZ? 1.62 2.36 Max allowable inserted rod verth 0.30 0.50 Flux redistribution 0.76 1.20 Total required worth 2.68 4.06 Shutdven nargin (cocal available worth minus total required worth) 3.11 2.28 Note: Required shutdown margin is 1.00% ak/k.
5-5 Babcock a, Wilcox t
Figure 5-1. Oconee 1. Cycle 3 3CC (4 IF?D) ?.ro-Df=ensional Relative ?cuer Distribution - Full Power, Equilibriu=t Xenon, Nornal Rod Positions 8 9 10 11 12 13 14 15 0.968 1.120 1.063 1.237 0.979 1.159 1.015 0.567 H ,
LTA j LTA Mark-3Z g 1.065 1.270 1.088 1.236 1.016 1.173 0.574 6
- 1. 1.082 1.259 0.980 1.287 0. 96 2 0.420 M 1.128 1.266 1.097 0.718 N . 1.190 1.071 0.466 .
0 0.579 P
R 1 INSERTED R00 GROUP NO.
X. XXX RELATIVE POWER DENSITY 5-6 Babcock & Wilcox
- 6. IdEDfAI.-R'DRAti C 07.5:3 Incoming ba:ch 10 fuel is hydraulically and gec=e::1cally sd d'a to the fuel re=aining 1: the core from previous cycles. "hermal-hydraulic design evalua-tion supporting cycle 3 operation used :he methods and models described is references 3, 4, and 13.
Cycle 3 =azd-=- design condi: ions re=ain unchanged f m cycle 7 and are shown in Table 6-1. The four Mark 3Z demonstration asse=blies and five I.TA demon-stracion assemblies will be conservatively limi:ed to a design peak of 1.61 (6: peaki:3 redue: ion) :s ensure : hat they are co: :her= ally Iisi:ing. A 1.71 design radial-local paak ramm M valid for all c:her asse=blies is this .;
cycle.
A rod bov penal:7 has been calculated according to the procedure approved in reference 14 The maxi =um fual assembly burnup of the ba:ch : hat contains the limiting (=azi=um radial-local peak) fuel assembly is used. yor cycle S, this burnup is 17,511 is a batch 10C assembly. The resul: ant net rod bow penal:y, after inclusion of the 1: flow area redue:1on factor credit, is 0.2% redue:1on 1: DNBR. Thermal-hydraulic design for cycle S includes a margin greater than 0.2% above the m4"d- - DN3R of 1.30.
6-1 Babcock & Wilcox
Table 6-1. Ther=al-Hvdrauli: Design Conditions Cycle 7 Cvele 3 Power level MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, 2 design flow 106.5 106.5 7essel inlet coolant camp, 1002 power, 7 555.6 555.6 Vesstl outlet coolant temp, 1002 power, 7 602.4 602.4 Ref design axial flux shape 1.5 cos 1.5 ces Ref design radial-local power peaking factor 1.71 1.71 Active fuel length, in. (a) (a)
)
Average heat flux, 1002 power, 10 3Stu/h-fe 176 0) l 176 CEF correlation BAW-2 BNJ-2 Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Minimum DN3R vich densification penalty 2.05 2.05 f")See Table 4-2.
b)3ased ou densified length of 140.3 in.
6-2 Babcock & Wilcox
. . j
- 7. ACCIDENT AND TRANSIINT ANALYSIS 7.1. General Safety Analvsis Each FSARI accident analysis has been examined with respect to changes in cycle 8 parameters to determine the effect of the cycle 8 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 3AW-1388." Since batch 10 reload fuel assemblies contain fuel rods with higher theoretical density than those considered in the reference 13 report, the conclusions in that reference are still valid.
7.2. Accident Evaluation The key parameters in determining the outcoms of a transient can typically be classified in three rrjor areas: core thermal parameters, thermal-hydraulic parameters, and kinetics pararecers, including the reactivity feedback coeffi-cients and control rod wor.hs.
Core ther=al 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 8 are compared in Table 4-2. The cycle 8 thermal-hydraulic * * *"* design conditions are compared to the pre-vious cycle 7 values in Table 6-1. These parameters are common to all of the accidents considered in this report. The key kinetics parameters from the
?SAR and cycle 8 are compared in Table 7-1.
A generic LOCA analysis for the 36W 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." The analysis in 3AW-10103 is generic since the limiting values of key parameters for all plants in this category were used.
Furthermore, the combination of average fuel te=perature as a function of LER and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the 7-1 Babcock & Wilcox
analysis and :he LOCA limi:s reported in 3AW-10103 provida conservative resul:s for the operation of Oconee 1, cycle a fuel.
Table 7-2 shows the bounding values for allowable LOCA peak L3Rs for Oconee 1, cycle 8 fuel after 50 EFFD. The LOCA kW/ft limits have been reduced for the first 50 E7FDs in order to account for mechanistic fuel densification. The i
redue:1on will ensure that conservative limits are maintained while a transi '
tion is being made in the performance codes that provide input to the ECCS analysis:s. The reduced limits for the first 50 EFFD are shown in Table 7-3.
~
The Oconee 1, cycle 8 core contains four Mark 3Z demonstration assemblies and five gadolinia LTAs. As a result of material and geometrical differences, these nine assemblies have LOCA kW/f: limits that are lower in some cases :han the standard Mark 3 h t:s. 3och the Mark 3Z assemblies and LTAs are being loaded in the core in a manner to ensure that there is sufficient margin to offset any negative impact on the LOCA kW/f: limits.
It is concluded frem the examination of cycle 8 core thermal and kinetics properties, with respect :o acceptable previous cycle values, that this core reload will not adversely affect the ability of the Oconee 1 plant to operate safely during cycle 8. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 8 is con-sidered to be bounded by previously accepted analyses. The ini:ial condi: ions for the transients in cycle 8 are bounded by the FSAR 1, the fuel densification repor:13, and/or subsequent cycle analyses.
The radiological dose consequences of the acciden:s presented in chapter 15 of the FSAR were recalculated using the specific parameters applicable :o cycle 8. The bases used in the dose calculations are identical ec : hose in the FSAR except that updated dose conversion factors were used. The use of the updated dose conversion factors resul:ed in reduced whole body dose values. l
! Table 7-4 compares the revised FSAR dose values with those calculated specifi-cally for cycle 8. As can be seen from the table, some cycle 8. doses vary slightly from the FSAR values. However, all cycle 8 doses are either bounded !
by the values presented in the FSAR or are a small frac:1on of the 10 CFR 100 l limits, i.e. below 30 REM :o che thyroid or 2.5 REM to the whole body. Thus, the radiological impact of the accidents during cycle B are not significantly different than those described in chapter 15 of the FSAR.
l I
7-2 Babcock & Wilcox i i
l
l l
1 Tabla 7-1. Coccarison of Kev ?araceters for Acciden: Analvsis FSAR and Fredic:ed densification cycle 8 Parameter report value value Doppler coeff, 10-8 Ak/k/F BOC -1.17 -1.54 ECC -1.33 -1.78 Moderator coeff, 10~" ak/k/F 30C ' +0.5 -0.67 EOC -3.0 -2.85 All-rod group worth at EZP, 10 8.53
- Ak/k Initial boren conc's at HTP, ppm 1400 1365 Boron reactivity worth at 70F, 75 91 pps/1: Ak/k Max ejected red worth at EF?, 0.65 0.35
- Ak/k Dropped red worth (EFP), Ak/k 0.46 0.20 Tabla 7-2. LOCA Li=its. Oconee 1, Cycle 8, After 50 EFFD Elevacion, LER limits, f: kw/f:
2 15 .5 4 16.6 6 18.0 8 17.0 10 16.0 i
1 7-3 Babcock & Wilcox
Table 7-3. LOCA Li=1:s, ocenee 1, Cycle 3, 0-30 E7?D Elevation, LER li=1:s, ft W/ft 2 14.5 4 16.1 6 17.5 8 17.0 10 16.0 l
1 l
l 7_4 Babcock & Wilcox i
f
Table 7-4 :_= arise: :f FIA2 and i:le ! A::ident Doses FEAR doses, (*) Cycle 8 doses, ram rem
- 1. Fuel hm' ling Ac=ide=t Thyroid dose at EA3, 2 h. 0.50 0.51 b le body dose at IA3, 2 h 0.028 0.010
- 2. Steam Line 3:aak Thyroid dose at EA3, 2 h 0.20 0.20 Whole body dose at IA3, 2 h 0.002 0.001
- 3. Steam Generator Tube Failure Thyroid dose at EA3, 2 h 0.31 0.32 b le body dose at EA3, 2 h 0.058 0.027 i 4 Waste Gas Tank Rupture Thyroid dose at EA3, 2 h 0.27 0.2B h is body dose at EA3, 2 h 0.17 0.079
- 5. Control Rod Ejection Accident Thyroid dose at EA3, 2 h 1.44 1.38 Whole body dose at EA3, 2 h 0.004 0.002 Thyroid dose at L?:, 30 days 1.57 1.53 Whole body dose at L?I, 30 days (b) 0.002.
- 6. Loss of Ciolant Accident
~hyroid dose at IA3, 2 h 5.0 4.94 h is body dose at EAB, 2 h 0.010 0.005 Thyroid dose at L?!, 30 days 5.5 5.48 Wie body dose at L? , 30 days 0.010 0.007
- 7. Maximzm Hypothetical Accident Thyroid dose at EA3, 2 h 193 193 b le body dose at EA3, 2 h 1.4 1.12 Thyroid dose at L72, 30 days 180 180 b le body dose at L72, 30 days 0.62 0.44 (a)FSAR changed since cycle 7 reload.
(b)Not listed in FSAR.
7-5 Babcock & Wilcox
- 8. PROPOSED MCDIFICATIONS TO IICHNICAI, SPECITICATIONS The Technical Specifications have been revised for cycle 8 operation in accor-dance with the nethods of references 17-19 to account for changes in power peaking and control rod worths.
Sased on the Technical Specifications derived from the analyses presented in this reporr, the final acceptance criteria ECCS 11 nits will not be exceeded, and the thernal design criteria vill not be violated. Figures 8-1 through 8-12 are revisions to previous Technical Specification 11 nits.
I t
8-1 Babcock & Wilcox
.n. - . . . .
Figure 3-1. Core Protec:ica Safety Li=1:s for Oconee Uni: 1
' ~ ~
Thermal Power Level, 5
, , 120 i (-41.0,112.0) (33.0.112)
E; = 1.571 l ACCEPTABLE
" 110 32**b#
(-48.0,101.0) 4 PullP l l lg OPERATION -> 100 1 (48.0, 95.5) I l
(-41.0,89.899) 90 s
ACCEPTABLE I U
(-48*0*78*899) o 80 l j l0PIRAT10N g l-(48.O,73.399) ,!
j < - 70 I i
(-41.0,62.73)
(33.0.62.73) !
l ACCEPTABLE 4,3&2 PUMP
(- 48.O,51.73) .. 50 i
l lOPERAT10N g (48.O,46,23)
- i I .. 40 I i i
i l l l >
.. 30 l '
- I
.I .
UNACCEPTABLE g l!;;; l
.. 20 "l
- w
- 8 UNACCEPTABLE
, OPERATION " "
OPERATION e 10 "l".~ l
.I .~l t . i t I*i I
, 60 -40 -20 0 20 40 60 Reactor Poser Imaalance, 5 CURVE RC FLOW (GPN) 1 374,880 2 280,035 3 183,690 8-2 Babcock f. Wilcox
Figure S-2. Protective Systes Maxi =t:= Allevable Satpoints for Oconee Unic 1 inermal Power Level, 5 l
120
(-17.O,107.0) - 110 (17.O,107.0) !
j Ej = 1.0625 ' l l ACCEPTABLE
- 100 I h = -1.3750 ,
j 4 PUhP I
- (-33.0,90.0) oeEnArion
= 90 80(17,79.92) % 0, 85. 0 l g(17,79.92)
UNACCEPTABLE l UNACCEPTABLE l
OPERATION OPERATION ACCEPTA81.E l l !
4.3 PUMP
(-33.0,62.92)l ,PinATion l 60 l
1 (33.0,57.92)
(-17,52. 43) (17 52,43) l l ACCEPTABLE 40 l i l 4.3.2 Puwe
(-33.0,35.43) l
,,,,37;a, l (33. O,30. /43, I
I
". . - - 20 o l .!
9l 3l i, g dl dI ii ii ii j 1 t a;l :(tI i
!1 :! 1 I
-60 -4G -20 0 20 40 60 Reactor Power imoalance, 5 '
I i
I 8-3 Babcock & Wilcox
- l. .
Figure S-3. Rod ?csi: ion Li:1:s for Four-?t7 Cpera: ion.
0-50 .=?D, Ocenee 1. Cycle 3 (204,102) (280,102)
. OPERATION 100 -
NOT ALLOWED SHUT 00NN I MARGIN (275.92) 6 ,
- LIMIT i
!: b- 80 - (273,80) l l3 i
OPERATION RESTRICTED
'E "
, 60
% (139,50) (200,50) :
m:
= !
" i
- 40 j r.
3 OPERAT10N !
ACCEPTABLE 20 -
(S4,15)
(0.7.5) ' ' '
0 O 100 200 300 Roc Incer, 5 fitnarawn GR 5 O 75 100
.- GR 6 0 25 .
75 100
- * -I
~ GR 7 0 25 100 s-4 Babcock & Wilcox
l j
Figure 8-'*. Red Posi: ion L1=1:s for Four-Pu=p Opera:1cn Af:er 50 EF?D, Oconee 1. Cycle 8 (238,102) (275,102) 4W 100 OPERATION NOT Al.LOWE0 OWN 3 (275.92)
M I MARGIN
'E 80 . Lluli (256,80) -
'S !
=
- 3. 60
{
a i
's (173,50) (200,50)
- a
% 40 -
- OPERATION 5 ACCEPTABLE D 20 . -
(0,6.8) (81,15) a.
5 0 ' ' '
O 100 200 300 Acc Inces, 5 Intnarawn GR 5 ' ' '
O 75 100 GR 6 ' ' '
O 25 75 100 GR 7 '
0 25 100 t
a-5 Babcock f. Wilcox
Figure 3-5. Rod Position Lini:s for ~hree-Punp Operation, 0-50 EFPD, oconee 1, cycle 5 ;
i l 100 - !
OPERATION NOT ALLOWED i
! 80
" ( ' } ( ' }
3HUTDOWN O EARGIN
,E LIEli ,
- 80 -
%gsD
= (200,50) 2 40 -
(139,38)
= OPERATION E ACCEPTABLE
- 20 - -
- (64,11)
E ( . 6.1 ) , , ,
O 0 100 200 300 Roc innex, 5 Titnerawn GR 5 8 ' '
O 75 100 GR 6 ' ' ' '
0 25 75 100 GR 7 ' ' '
O 25 100 e
8-6 Babcock & Wilcox
Figure S-6. Red Posi: ion Limi:s f:r hree-?u=p Operation Af:er 50 ETPD, Oc::ee 1. Cycle 3
, 100 -
OPERATION RESTRICTED O
= 80 = 259.4,77)
OPERATION (238.77)
?
40T ALLOWED
. SHUT 00NN 5 MARGIN 5 50 = '
.E 2 (200,50) 40 .
j (373,33) 6 OPERATION
- 20 ACCEPTABLE 5
- ( 81,11 )
(0,5.6) , , ,
0 0 100 200 300 Roa inaex, 5 Wi tnarawn GR 5 ' ' '
O 75 100 GR 6 ' ' ' '
O 25 75 100 GR 7 ' ' '
O 25 100 i
t 8-7 Babcock & Wilcox
71gtre S-7. Rod ?csition I.ici:s for ?;o-?u=p Opera: ion, 0-50 I??D, Oconee 1, Cycle 3 100 -
l
.n
= 80 OPERATION tZ NOT ALLOWED IE
,3 SO =
i* SHUTDOWN .
j MARGIN -
a (204.9,52)
LIMIT
- 40 -
E 3
- 5. (139,26) OPERATION 20 ACCEPTABLE E (0,4.8)
- (64,8) r O'_ ' ' '
0 100 200 300 Roa Incex, 511 tnarawn GR 5 O 15 100 GR 6 ' ' ' '
0 25 75 100 GR 7 ' ' '
0 25 100 8-8 Babcock & Wilcox
l 1
1 Figure 3-3. Rod ? si: ion L1=its for Tve-?u=p Operation Af:er 30 IFFD, Oconee 1. Cycle 3 100 -
f L
II iEa 80 -
- r. OPERATION NOT ALLOWED l% "
li -
60
=
SHUT 00fN (238,52) jE MARGIN l; LIMIT
!E 40 -
- 3 2
~ OPER ATION
!- ACCEPTABLE
= 20 - (173,25) 2 (0.4.4)
, (81,8) 0 1 ' t 0 100 200 300 Rsa incer, 5 Witnarann GR 5 ' ' '
O 75 100 GR S ' ' ' '
j 0 25 75 100 GR 7 ' ' '
0 25 100 t
8-9 Babcock & Wilccx
Figure 3-9. Power != balance L1=1:s fcr 0-30 E7?D, Oconee 1 Cycle 3 OPERATION RESTRICTED
(-17,102) (20j02) 100
}
4 (-25,92) -- 90 (25.92)
' 80 , 4 (30,80) a
.- 70 *-
OPERATION -
m -
ACCEPTABLE 5 60 3 . _
= .
s-o.
50 =
-- 40 =
.u. .
. . s G. '
30 - -
w, ,
o ., .-
a- '
{ \
'^ I
-- 20 '
s ,
1 sa ,
s
-- 10 -
x ,
..i m,
i t t I t 1 1 30 -20 -10 0 10- 20 30 40 N.
Axial Power imoalance, 5 s \,- . b.
.x
' 's' (
5 _ c-
~
- q. l\ -r w -
, x. .,
\ Y ~'- -
\ ~,
, s T
. _ ,c g u,
s' t \ 's * \ \
\ \ g N, - i
\ -
. 3 - -
x x \w.sT.
IN \ ' '
y 3 , .
\
', 13-10 -
Babcock & Wilcox er g s, s r
.s . . . . .
ki *< . o b ...,
n
Figure 8-10. Fever I: balance L1:1:s Af:er 50 EF?D, Oconee 1, Cycle 8 OPERATION RESTRICTED
(-17,102) , (20']O2)
-- 100 (30,92)
I (-29,92)
. .. 90 I
l .
80 6(35,80)
OPERATION -- 70
- ACCEPTABLE 2
-- 60 G
E a
-- 50 a
E
-- 40 .
=
s
-- 30 2 -
B
.. 20 g 10 t s i I f
- I '
l I f I
-40 -30 -20 10 0 10 20 30 40 :
x Axial Power Imaalance, 5 s ,
.s
.~.
t ,
- s g
4
\
.,,..s. . . s s, s
t: ;. ~ ,
- ,,- .N lC, ,_ ,i .. 8-L1 Babcock & Wilcox
. t
,,. %=T ., g ;- i *-
- W *e s %-
s t+ .
s ._ A
i e' .
e I
.E.
e e.
e e
r e w a e
- e e
. e
, c E w-c u
w 4 y -
as - .
,e -
Iw g, w
- *w ~
t 0 < =
u a. e a
o . -
=
.
. o
.r. e .
- .o
= m o e a
v . o
o ><
e.
u ,i-
. a. *
- e =
a a.
-u . a. w u ~
n
,. e - e, .
m
== = n,.,
- m.
a:
A. W . .
- 4. 4
<= m w aw y
- w U
. = =
- m
- e ,
I - #
= .-
u .
w w o a. . -
w ~. e ~.
~
w e
= a
- e
~ -
- e. .-
e n o.
- e
- =
a e t
w i . e f , g ,
f f ,
e o e e e e o e e e, _e e e
e . - e o c. ; ~
Ii (Janad lecJaul satry to lua:Jad) 2:aca f
/
/
f /
/
/
8-12 Babcock & Wilcox
L -
! ! ' l m s i e s
)
e 5
s e
l e a
( s s
i e .
S I E
. O I l C D l I P A R x F R T E E S a a P E l 0 O S
. 5 r
e t n f I e v A s a r
a i
t s
q n t
a i
mS s
B L e ) . e l
8 S 5 nc oy ) 8, ,
2 5 n i C 9 . e t 8 t l . 8, t
( e u1 3 o ( i e s P e )
) 4 o c 2 ( P s t i 0 i s 1 R S o , . S P c 4 P AO 3 A m (
a o a
2 1
8 E e WL o
r 0B i
_ ) l A 2 u 2 l I g
i _
9, A P RE F ) 0 E C 2 1 P C 0 ( OA 1 '
a e
_ I
, m )
S l
I S
( .
_ S k
- ~- - - . - - - - a
(
0 0
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0 5
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0 3
0 2 t o 8 EE E E oSE ;
- n;t ;E TC rRgx e gg*
.. 4 s
- 9. RETIKENCES 1
Oconee Nuclear Station, Uni:s 1, 2, and 3 -- Final Safety Analysis Report, Docket Nos. 50-269, 50-270, and 50-287, Duke Power Company.
- Design Report for Oconee 1 Lead Test Assemblies, BAW-1772, Babcock & Wil- .
cox, Lynchburg, Virginia (to be published).
Mark 3Z De=onstration Assemblies in oconee 1, Cycles 7, 8, and 9, 3AW-1661, Babcock & Wilcox, Lynchburg, Virginia, March 1981.
3PRA Retainer Design Report, 3AW-1496, 3abcock & Wilcox, Lynchburg, Virginia, May 1978.
s J. H. Taylor (3&W) to S. A. Varga (NRC) . Lat:er, "3PRA Retainer Reinser-tion," January 14, 1980.
A. F. J. Eckert, H. W. Wilson, and K. E. Yoon, Program to Determine In-t.
i . Reactor Performance of 3&W Tuels -- Cladding Creep Collapse, 3AW-1COS4A, Rev. 2, 34bcock & Wilcox, Lynchburg, Virginia, October 1978.
7 Y. H. Esii, et al. , TACO 2 -- Fuel Pin Performance Analysis, 3AW-10141?,
3abcock & Wilcox, Lynchburg, Virginia, January 1979.
J. H. Taylor to J. S. Berggren, Lat:ar, "3&W's Responses to TACO 2 Ques-tions, Babcock & Wilcox, April 8, 1982.
H. A. Hassan, W. A. Wittkopf, and W. A. Mullan, 3&*=* Version of PDQ07 Code, 3AW-10117A, 3abcock & Wilcox, Lynchburg, Virginia, January 1977.
11 J. J. Romano, Core Calcula:1onal Techniques and Procedures, 3AW-10113A, 3abcock & Wilcex, Lynchburg, Virginia. December 1979.
11 M. R. Gudorf, G. E. Hanson, and J. R. Lojek Assembly Calculations and Fi :ed Nuclear Data, 3AW-10116A Babcock & Wilcox, Lynchburg, Virginia, May 1977.
9-1 Babcock & \Vilcox
~ '
13 Oconee 1 fuel Densification Report, 3AW-1388. Rev. 1, 3abcock & Wilcox, Lynchburg, Virginia, July 1973.
I' L. S. Rubenstein (NRC) to J. H. Taylor (3&W), Letter, " Evaluation of In-teria Procedure for Calculating DN3R Reductions Due to Rod 3cv," October 18, 1970.
is R. C. Jones, J. 3. Biller, and 3. M. Cunn, ECCS Analysis of B&W's 177-TA Lowered-Loop NSS, 3AW-10103, Rev. 1, Sabcock & Wilcox, Lynchburg, Virginia, September 1975.
l' J. H. Taylor (3&W) to L. S. Rubenstein (NRC) Letter, September 5,1980.
l' H. A. Hassan, et al. , Power Peaking Nuclear Reliability Factors, BAW-10119, Sabcock & Wilcox, Lynchburg, Virginia, January 1977.
18 G. E. Hanson, Normal Operating Controls, BAW-10122, Babcock & Wilcox, Lynchburg, Virginia, August 1979.
l' C. W. Mays, Verification of the Three-Dinansional TLAME Code, 3AW-10125A, 3abcock & 'Jilcox, Lynchburg, Virginia August 1976.
l l 9-2 Babcock t iVilcox
__ _