ML19343C962
| ML19343C962 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/31/1981 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19343C506 | List: |
| References | |
| BAW-1667, TAC-43369, NUDOCS 8104060334 | |
| Download: ML19343C962 (52) | |
Text
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e 3AW-1667 March 1981 RANCHO SECO UNIT 1 CYCI.E 5 RELOAD REPORT l
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B/oYO6033 y Babcock & Wilcox
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BAW-1667
-March 1981 i
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i IULNCHO SECO UNIT,1 -
4 CYCLE-5 RELOAD REPORT 3
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5 j-3ABCOCK & WILCOX Nuclear. Power Group Nuclear Power. Generation Division 1-P. O. Lox 1260 Lynchburg, Virginia 24505 3
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Babcock & Wilcox.
e CONTENTS Page 1.
INTRODUCTION AND 3UMMARY.................
1-1 2.
OPERATUG HISTORY 2-1 3.
GENERAL DESCRIPTION 3-1 4.
FUEL SYSTEM DESIGN..
4-1 4.1.
Fuel Assembly Mechanical Design 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 4-2 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.
THERMAL-HYDRAULIC DESIGN.
6-1 7.
ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.
General Safety Analysis 7-1 7.2.
Accident Evaluation 7-2 8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS.
8-1 9.
STARTUP PROGRAM - PHYSICS TESTING 9-1 9.1.
Precritical Control Rod Trip Test 9-1 9.2.
Zero Power Physics Tests.
9-1 9.2.1.
Critical Boron Concentration.
9-1 9.2.2.
Temperature Reactivity Coefficient 9-1 9.2.3.
Control Rod Group Reactivity Worth.........
9-2 9.2.4.
Ejected Control Rod Reactivity Worth...
9-3 9.3.
Power Escalation Tests.
9-3 9.3.1.
Core Power Distribution verification at s40, 75, and 100% FP With Nominal Control Rod Position 9-3
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CONTENTS (Cert'd)
Page 9.3.2.
Incore Vs Excore Detector Imbalance Correlation Verification at N40% FP'...
9-5 9.3.3.
Temperature Reactivity Coefficient at %100: FP 9-5 9.3.4.
Power Doppler Reactivity Coefficient at
%100: FP 9-5 9.4.
Procedure for Use When Acceptance Criteria Are Not Met 9-5 REFERENCES 10-1 7
List of Tabl S Table 4-1.
Fuel Design Parameters and Dimensions. Rancho Seco Cycle 5... 44 4-2.
Fuel Thermal Analyais Parameters................ 45 5-1.
Physics Parameters, Rancho Seco Cycles 4 and 5......... 5-3 5-2.
Shutdown Margin Calculation - Rancho Seco Cycle 5 5-4 6-1.
Maximum Design Conditions, Cycles 4 and 5 6-2 7-1.
BoundinE Values for Allowable LOCA Peak Linear Heat Lates 7-3 7-2.
Compari son of Key Parameters for Accident Analysis -- Rancho Seco Cycle 5..........................
7-4
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List of Figures l
Figure l
3-1.
Core Loading Diagram -- Rancho Seco Cycle 5...........
3-2 l
3-2.
Enrichment and BOC Burnup Distribution "ancho Seco Cycle 5 l
Off a 220-EFPD Cycle 4..
3-3 l
3-3.
Control Rod Locations and Group Designations for Rancho Seco Cycle 5..........................
3-4 3-4.
Rancho Seco Cycle 5 L3P Concentration and Distribution.
3-5 5-1.
30C 5 (4 EFPD) Two-Dimensional Relative Power Distribution.
l Full Power, Equilibrium Xenon, Normal Rod Positions 5-5 8-1.
Core Protection Safety Limits, Reactor Power Imbalance.
8-2 8-2.
Protective System Maximum Allowable Setpoints, Reactor Power Imbalance 8-3 8-3.
Rod Index Vs Power Level-for Four-Pump Operation, 0-60 EFPD 8-4 8-4.
Rod Index Vs Power Level for Four-Pump Operation From 50 to 250 EFPD.
8-5 l
8-5.
Rod Index Vs Power Level for Four-Pump Operation After 230 EFPD With APSRs Out (Cycle 3) 8-6 I
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Figures (Cont'd)
Figure-Page 8-6.
Rod Index Vs Power Level for Three-Pump Operation, O to 60 EFPD (Cycle 5) 8-7 8-7.
Rod Index Vs Power Level for Three-Pump Operation From 50 to 250 EFPD (Cycle 5) 8-8 8-8.
Rod Index Vs Power Lev 01 for Three-Pump Operation, After 230 EFPD With APRSs out (Cycle 5) 8-9 8-9.
APSR Withdrawal Vs Power Level. 60 EFPD (Cycle 5) 8-10 8-10.
APSR Withdrawal Vs Power Level, 50-250 EFPD (Cycle 5) 8-11 8-11.
Core Imbalance Vs Power Level, O to 60 EFPD (Cycle 5) 8-12 8-12.
Core Imbalance Vs Power Level 50 to 250 EFPD With APSRs In (Cycle 5) 8-13 8-13.
Core Imbalance Vs Power Level, After 230 EFPD With APSRs Out (Cycle 5).
8-14 8-14.
LOCA Limited Maximum Allowable Linear Heat Itate 8-15 l
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1.
INTRODUCTION AND
SUMMARY
This report justifies operation of the Ranc' e Seco Nuclear Generating Station, Unit 1, cycle 5, at a rated core power of 2772 MWt.
The required analyses are included as outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," June 1975. This report utilizes the analyt-ical techniques and design bases documented in several reports that have been submitted to the USNRC and approved by that agency.
Cycle 5 reactor and fuel parameters related to power capability are summarized in this report and compared to those of cycle 4.
All accidents analyzed in the Rancho Seco FSAR have been reviewed for cycle 5 operation and, in cases where 5 characteristics were conservative compared to those of cycle 4, no new analyses were performed.
The Technical Specifications have been reviewed and modified where required for cycle 5 operation. Based on the analyses performed and taking into ac-count the ECCS Final Acceptance Criteria and postulated fuel densification ef-facts, it is concluded that Rancho Seco cycle 5 can be operated safely at its licensed core power level of 2772 MWe.
l Retainera will be installed on all fuel assemblies containing burnable poison rod assemblies (3PRAs) and on the two fuel assemblies containing regenerative neutron sources. The retainers will provide positive retention during reactor operation. The effects of continued operation without orifice rod assemblies l
(ORAs) and the addition of the BPRA retainers have been accounted for in the I
analysis performed for cycle 5.
The design for cycle 5 also incorporated the following changes with respect to previous cycles:
- The design includes four axial blanket lead test assemblies (LIAs).
These assenblies are part of a Department of Energy Axial Blanket Program. The LIA design is described in reference 2.
1 1-1 Babcock & Wilcox I
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The recent concern about differences between the IAFY and TACO codes used for fuel and ECCS analysis, as discussed in reference 3, are ap-plicable to this cycle. Appropriately revised LOCA kW/f t limits are used for this cycle (Table 7-1).
As a result of the comparatively short duration of the three previous cyc:'.es, 28 of the assemblies from batch 4 are being recycled for a fcuilth cyc12 The expected incore residence times are well within previously justified limits (Table 4-1).
In addition to the use of lunped burnable poison rode (LBPRs) in all of the fresh fuel, L3 prs will be used in eight of the once-burned batch 6 assemblies to improve the power distribution (sections 3 ana 5).
As discussed in refe. ence 4, the concern that the error in the nuclear instrumentation determination of flux level in certain overcooling transients could be larger than previously assumed has been taken into account and the appropriate Technical Specificacica limits (section 8) have been correspondingly adjusted.
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2.
OPERATING HISTORY Cycle 4, the current Rancho Seco Unit 1 operating '.ycle, is-the reference fuel cycle for the nuclear and tharmal-hydraulic analyses performed for cycle 5 opera ion. Cycle 4 achieved initial criticality en May 9, 1980, and power escalation began en May 12, 1980.
No operating anomalies occurred during cycle 4 operation char would adversely affect fuel performance during cycle 5.
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3.
GENERAL DESCRIPTION The Rancho Seco reactor core is described in detail in Chapter 3 of the FSAR.s The core consists of 177 fuel assamblies, each of which is a 15 by 15 array containing 208 fuel rods,16 control rod guide tubes, and one incore instru-ment guide tube. The fuel consists of dished-end, cylindrical pellets of ura-nium dioxide clad in cold-worked Zircaloy-4. All fuel assemblies in cycle 5.
Lactuding the four axial blanket lead test assemb3.ies (LIAs) described in ref-erence 2, maintain a conscant nominal fuel loadinI of 463.6 kg of uranium.
The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod limensions, and other related fuel param aters may be found in Tables 4-1 and 4-2 of this report.
Figure 3-1 is the core loading diagran for Rancb s Seco cycle 5.
Nine once-23s burned batch 1 assemblies with an initial enrictment of 2.01 wt %
U will be reloaded into the core. Batches 4B, 5, and 6, with initial enrichments of 235 3.19, 3.04, and 3.21 we %
U, respectively, vill be shuffled to new locations.
235 Batch 7, with an initial enrichment of 3.14 wt :
U, will be loaded in a checkerboard pattern. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle S.
Reactivity is controlled by 61 full-length Ag-In-Cd centrol rods, 40 LBP clus-ters, and soluble boron shim.
In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of 1
l the axial power distribution. The cycle 5 location of the 69 control rods and the group designations 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 ref-erence cycle described in the Rancho Seco cycle 4 reload report.s The group designations, however, differ between cycle 5 and the reference cycle. The l
cycle 5 locations and LBP concentrations are shown in Figure 3-4.
l LBP clusters are used in all 32 fresh batch 7 assemblies and in 8 once-burned batch 6 assemblies.
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Figure 3-1.
Core Leading Diagra::1 - Rancho Seco Cycle 5 I
tcEL *Ub7ER CA.4AL t
I 6
6 6
6 5
4 43 S4 E3
'*12 413 6
6 5
7 7
5 6
6 8
(2 L5 54 313 512 Lil r4 t.3 5
5 5
5 43 5
43 5
5 6
5 C3
'.3
'*2 46 C7 C3 C$
413 v14 L13 C13 6
5 5
7 IC 7
5 7
It
?~
5 5
0 39 C13 36
,;5, 35
,;13, F14
- 5 37 6
5 7
5 5
4B 5
43 5
5 7
5 6
E E13 511
' ;3 47 C5 13 C11 49 L6 85 E6 5
5 5
it 5
48 7
48 7
28 5
1C 5
5 6
F
'1 02 el E3,,
il r4 03
- 12 115 E3 FIS 014 C4 6
7 48 7
43 7
43 5
AS 7
43 7
48 7
6 3
011 13 I3 47 43 39
!!3 113 05
,.3 6
5 6
5 5
28 5
it 5
48 6
5 5
5 6
~5 F2 m3 4
47 4
- 15
%)
al M12
-9 914
-13 L14 wil t
i 6
7 48 7
43 7
43 5
43 7
43 7
49 7
6 E
%11 (3
13 C
Aa O
v13
';U 95 gg c
5 5
1C 5
at 7
45 7
45 5
it 5
5 6
L 212 12 L1 v3
<1 L4 N3 L12 (15 r 13..
.15 414 M
l
,3, l
6 5
7 5
5 49 6
43 5
5 7
5 6
l M
w!3 Pil r10 47 25 (3
311 29 F6 35 46 5
5 5
7 1C 7
5 7
- C 7
5 6
6 P9 313 L2 25 P8 All 810 87 re,ys r,, s 5
5 5
5 AS 6
43 5
5 6
5 0
23 F3 E2 46 27 28
- 9 RIO E14 F13 113 I
I 6
5 5
7 5
7 5
6 6
l 12 75 74 P6 812
- l 114
,.3 5
5 6
6 6
4 23 E4 E1Z 313 l
l 2
l 1
2 3
4 5
6 7
5 9
to 11 12 13 14 15
!at:n 4 9V c'J$ C0re *.0C45109 i
l (L*4 4mial 31ac et Less 'es assemoly) 1 I
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Figure 3-2.
Enrichment and 30C Surnup Distribution Rancho Seco Cycle 5 Off a 220-EFPD Cycle 4 8
9 10 11 12 13 14 15 2.01 3.04 3.19 3.21 3.04 3.21 3.04 3.21 4
H 18.058 14,220 27,451 9,590 21,501 9,489 18.703 9.308 3.19 3.14 3.19 3.14 3.19 3.14 3.21 K
28,531 0
25,141 0
27,265 0
9,310 3.19 3.04 2.01 3.04 3.04 3.21 L
28,266 13,710 14,625 11,151 11,237 8,306 3.04 3.14 3.04 3.21 M
20,882 0
15,469 9.157 3.04 3.21 3.21 N
18,700 9,148 8,936 3.04 0
12,791 P
R
" 2'8 X.XX Laitial enrich =ent, wt U
XX XXX 30C burnup,.%'d/stU 3-3 Babcock & Wilcox
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Figure 3-3.
Control Rod Locations and Group Designations for Rancho Seco Cycle 5 X
l A
4l B
4
/
C l1 l6 6
1 0
l 8l 5l 8l l7 7
2l 5l l1I l5 g
1 2
F 4 I l8 l 3l l8 3
7 4
G 6
2 4
4 2
l6 l
7l H
'4 7l l 7l l5 7
y 5
2 l2 K
6 2
4 4
6 3l 4
8 3
7 8
4 L
!j l
1l 5
2 2
5 1
N l
7 8
5 8
7 4
0 1
6 6
i P
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l I"
4 7
4 liIII a
i 1
1 2
3 4
5 5
7 8
9 10j11 12 13 14 15 x
Group Number Group No. of Rods Function 1
8 Safety 2
9 Safety l
3 4
Safety l
4 12 Safety l
5 8
Control 6
8 Centrol 7
12 Controi 8
8 APSRs Total 69 l
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Figure 3-4 Rancho Seco Cycle 5 L3P Concentration and Distribution 8
9 10 11 12 13 la 15 H
0.2 0.2 8
X 0.8 1.1 0.5 L
0.8 l
l M
0.2 1.1 N
1.1 1.1 1
0 0.2 i
p 0.5 i
R i
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l X.X L3P Concentration.
l at ! 3sc in A1:03 I
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4.-
FUEL SYSTEM DESIGN 4.1.
Fuel Assembiv Mechanical Design The types of fuel assemblies and pertinent fuel design paran.aters for Rancho Seco cycle 5 are 1?.sted in Table 4-1.
Batch 7 includes four axial blanket lead test assemblies (LTAs). All fuel assemblies are identical in concept and are interchangeable. The analyses and justification for the LTAs are reported in reference 2.
Retainer assemblies will be used on 32 fresh batch 7 assemblies and 8 once-burned batch 6 assemblies. Some of the retainers will be inserted for a second-cycle. The justification for the design and use of the retainers is described in references 1 and 7.
4.2.
Fuel Rod Design The batch 7 fuel differs from that of batches 4B, 5, and 6 in two respects.
As outlined in Table 4-1, these include an increase in initial pellet density from 94 to 95% TD and a decrease in fuel pellet diameter from 0.3700 to 0.3686 inch. These combined changes were implemented to improve fuel performance.
The mechanical evaluation of the fuel rod is discussed below.
4.2.1.
Cladding Collapse The batch 4B fuel is more limiting in cycle 5 than batches IC, 5, 6, and 7 because of its previous incore exposure time. The batch 4B assembly power his-tories were analyzed to determine the most limiting four-cycle power history for creep-collapse. The worst-case power history was then compared against a generic analysis to ensure that creep-ovalization will not affect fuel per-formance during cycle 5.
The generic analysis was performed based on refer-ence 8 and is applicable to the batch 4B design.
The creep-col 2 apse analysis predicts a collapse time of = ore chan 35,000 ef-factive full power hours (EFPH), which is longer than the maximum expected residence time of 29,280 EFPH (Table 4-1).
4-1 Babcock & Wilcox
4.J.2.
Cladding Stress The cycle 5 strese parameters are et.veloped by a conservative fuel rod stress analyses.
The same method was used for the analysis of this cycle and cycle 4.
4.2.3.
Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding plastic tensile circumferential strain. The pellet is designed to ensure that cladding plastic strain is less than 1% at the design local pellet burnup and heat generation rate. The design burnup and heat generation rate are higher than the worst-case values Rancho Seco fuel is expected to see. The strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density and the lower tolerance value for the cladding ID.
4.3.
Thermal Design All fuel in the cycle 5 core is thermally similar. The design of the four 2
becch 7 axial blanket LTAs is such that the thermal performance of this fuel is equivalent to the standard Mark B design used in the remainder of the core.
The thermal design evalt.stion results for the cycle 5 core are summarized in Table 4-2.
Linear heat rate capabilities are based on centerline fuel melt l
with coco orotection limits based on a 20.4 kW/ft LHR to centerline fuel melt l
as determined by the TAFY-3 code.'
The nine batch IC assemblies have fuel l
l melt limits from 19.2 to 20.4 kW/f t and are selectively loaded. l*
A design i
peaking margin of more than 6% is maintained in cycle 5 for these selectively loaded assemblies. The maximum fuel rod burnup at EOC 5 is predicted to be 42,000 mwd /mtU. Fuel rod internal pressure has been evaluated with TAF1-3 for the highest-burnup fuel rod and is predicted to be less chan the nominal reac-tor coolant system pressure of 2200 psia.
4.4.
Material Design t
l The chemical compatibility of all possible fuel-cladding-coolant-assembly in-l teractions for tacch 7 fuel assemblies is identical to that of the present fuel.
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4.5.
Operating Experience Babcock & Wilcox op(rating experience with the Mark B 15 by 15 fuel assembly has verified the adequacy of its design. As of September 30, 1980, the follow-ing experience has been accumulated for the eight operating B&W 177-fuel assem-bly plants using the Mark 3 fuel assembly:
Max FA burnup, Cumulative net d/mtU Current electrical output, Reactor cycle Incore Discharged MWh Oconee 1 6
23.300 40,000 32,457,943 Oconee 2 5
26,100 33,700 27,786,436 Oconee 3 5
30,200 29,400 28,483,452 TMI-1 4
32,400 32,200 23,840,053 AN)-1 4
28,100 33,222 25,006.003 Raacho Seco 27,900 37,100 22,625.102 Crys al River 3 3
20,530 23,194 12.113,632 e
Davis-Besse 1 1
14,884 7,654,365 i
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Table 4-1.
Fuel Design Parameters and Dimensions, Rancho Seco Cycle 5 Batch No.
1C 4B 5
6 7
Fuel assembly type Mark 33 Mark B4 Mark B4 Mark 34 Mark B4 Number of assemblies 9
28 56 52 32(a)
Fuel rod OD, in.
0.430 0.430 0.430 0.430 0.430 Fuel rod ID, in.
0.377 0.377 0.377 0.377 0.377 Flexible spacer Corrugated Spring Spring Spring Spring Rigid spacer Zr-4 Zr-4 Zr-4
~ -4 Zr-4 z
Undensified fuel length (nom), in.
141.75 142.75 142.00 141.75 141.80 Pellet OD (mean speci-fied), in.
0.3686 0.3697 0.3697 0.3700 0.3686 Fuel pellet initial density (nom), % TD 95.0 94.0 94.0 94.0 95.0 l
Initial fuel enrichment, wt % 23sU 2.01 3.19 3.04 3.21 3.14 l
Average burnup, BOC, mwd /mtU 15,007 27,009 14,995 9,085 0
Cladding collapse time, EFPH
>35,000
>35,000
>35,000
>35,000
>35,000 1
Estimated residence time, EFPH at discharge 18,362 21.523 26,841 27,240 29,280 l
(*)Batebd 7 includes four axial blanket lead test assemblies, which are de-l l
scri e in reference 2.
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Table 4-2.
Fuel Ther=al Analysis Parameters Batch No.
1C 43 5
6 7
C 32 ")
No. of assemblies 9
28 56 52 Initial density, % TD 95 94 94 94 95 Pellet diameter, in.
0.3680 0.3697 0.3697 0.3700
-0.3686 Nominal stack height, in.
141.75 142.08 142.08 141.80 141.80 Densified Fuel Parameters Pellet diameter, in.
0.3649 0.3648 0.3648 0.3651 0.3649 Fuel stack height, in.
140.7 140.2 140.2 140.0 140.7 Nominal linear heat rate, kW/ft at 2772 MWt 6.25 6.27 6.27 6.28 6.25 Average fuel temperature at nominal LHR, F 1356 1353 1353 1348 1343 LHR capability, kW/f t t C fuel melt 20.4(c) 20.4(d) 20.4 20.4 20.4 g
(a) Includes four axial blanket lead test assemblies, described in reference 2.
(IDensificaciou to 96.52 TD assumed.
(" The batch I fuel assembly have fuel melt limits frem 19.2 to 20.4 kW/f t and are selectively loaded.18
(
Except for two selectively loaded fuel assemblies, which have a linear heat capability of 20.25 kW/f t.
Core Average LHR = 6.27 kW/ft.
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5.
STCLEAR DESIGN 5.1.
Physics Charac: eristics Table 5-1 cc pares the core phys!es para =eters of design cycles.4 and 5.
he values for both cycles were generated using FDQO7. The average cycle burnup-will be lover in cycle 5 than in the desiga of cycle 4 The =axi=un'asse=bly burnup at end of cycle 5 (275 EF?D) is 37,666 Wd/=tU in batch 43 asse=blies in core locations sy==e:ric to K9.
This burnup is si=ilar to the previous high of 37,100 Wd/s:C in four batch 3 asse=blies at the e::d of cycie 3 and less than the four batch 4 assemblies that at:ained a 40,000 Wd/=tU burnup in Oconee I cycle 5.
Although cycles 4 and 5 are both feed-and-bleed cycles with an APSR pull near ECC, differences between the physics para =eters of the two cycles can be a:-
tributed to the initial L3? leading, shorter design life, and different shuffle pattern for cycle 5.
Calculated ejected red worths and their adherence to criteria are considered at all ti=es in life and at all pcVer levels in :he develop =ent of the rod position li=its presented in section 8.
The =axi=u=
stuck red verth for cycle 5 is similar to that for the design cycle 4 at both 30C and EOC. All safety criteria associated with these worths are =et.
The adequacy of the shutdova =argin with cycle 5 stuck rod worths is de= ens::ated in Table 5-2.
The following onservatis=s were applied for :he shu devn cal-culations:
1.
10 uncer:r.inty on net rod worth.
2.
Flux redistribution penal:7 Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a :vo-di=ensional =cdel.
The reference fuel cycle shutdown =ar-gin is presented in *.ne bncho Seco cycle 4 reload report.'
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~5.2.
' Analytical Input The cycle 5 incore measurement calculation constants to be used for comp 2 ting core power distributions were prepared in the same manner as those for the reference cycle.
5.3.
Changes in Nuclear Design There are no significant core design changes between the reference and reload cycles.
The same calculational methods and design information were used to obtain the important nuclear design parameters for this cycle.
The APSRs will be withdrawn from the core during the last part _ of cycle 5 (240 EFPD). The stability and control of the core in this mode have been analyzed; the calculated stability index without APSRs is -0.033 h-1, which demonstrates the axial stability of the core.
The difference in stability index between cycles 4 and 5 is due to the differences in xenon worth, Doppler coefficient, and axial burnup distribution between the two cycles.
The oper-ating limits (Technical Specification changes) for the reload cycle are shown in section 8.
5-2 Babcock & iMilcox t
Table 5-1.
Phy31cs Parameters, Rancho Seco Cycles'4 and 5 0}
Cvele 4 Cvele 5 Cycle lenath. ETPD 335 275 Cycle burnup. mwd /mte 11.316 9.289 Average core burnup - EOC.ICI mwd /stU 22.119 21.738 Initial core loadina mtU 82.1 82.1 Critical boron - BOC. ppa (no Xe)
EP(d). group S inserted 1372 1252 HFP, aroup S inserted 1154 1026 Critical boren - ECC. ppa (eq Xe) f4 f65 proupe 1-0 1002 wd 3
Control rod wortLa - HFP(, SCC. t ak/k Group 7 1.56 1.52 Group S 0.38 0.36 Control rod wroths - HFP. 240 ETPD croup 7 1.52 1.59-Group 8 0.46 0.44 Max ejected rod worth - EP
- ak/k 30C. groups 5-8 inserted 0.76 0.57 240 EFPD. Aroups 5-8 inserted 0.63 0.59 Max stuck rod worth - hip. 3 ak/k BOC 1.89 1.83 240 ETPD 1.56 1.83 Power deficit. HZP to HTP. % ak/k BOC 1.60 1.77 EOC 2.37 2.41 Doppler coeff stFP, 10~5 (ak/ k/ *F)
BOC (0 Xe. 1026 ppa, group 8 ins.)
-1.52
-1.53 EOC (eq Xe. 17 pps. group 8 100* vd)
-1.73
-1.70 Moderator coeff - HTP 10** (ik/k/*F)
BOC (0 Xe. 1026 ppe, group 8 ins.)
-1.05
-1.23-E0C (eq Xe. 17 ppm. group 8 100% vd)
-2.89
-2.92 Boron worth - HFP. ppm /*. Ik/k SOC. 1100 ppm 117 118 ECC. 17 ppa 103 104 Xenon worth - HFP. *. ak/k 4 !TPD 2.66 2.63 ECC. equilibrium 2.77 4.71 Effective Jelayed neutron fraction - HTP BOC. group 8 inserted 0.0060 0.0059 ECC.,troups 1-8 100* wJ 0.0052 0.0051 I* Cycle 5 data are for the conditions stated in this report.
The cycle 4 core conditions are identified in reference 6.
(b)3ased on 335 ETPD at 2772 MWt cycle 3.
(c)335 ETPD in cycle 4; 275 EFFD in cycle 5 unless otherwise stated.
(d)HZP:
zero power (532F T,
); HTP' hot f ull power (582F T
).
hot 5-3 Babcock s.Wilcox
~
Table 5-2.
Shutdown Margin Calculation - Rancho Seco Cycle 5'
% ik/'t
- ak/ k._
Available r p worth Total rod worth, HZP *)
8.60 8.98 I
Maximum stuck rod, HZP
-1.83
-1.83 Net worth 6.77 7.15 Less 10% uncertainty
-0.68_
-0.72 Total available worth 6.09 6.43 Recuired red worth Power deficit, HFP to HZP 1.77 2.43 Max allowable inserted rod worth 0.25 0.50 Flux redistribution 0.57 0.91 Total required worth 2.59 3.84 Shutdown margin (available worth 3.50 2.59 minus required worth)
(# HZP: hot zero power, HFP: hot full power.
Note:
Required shutdown nargin is 1.00 % ak/k.
5-4 Babcock & Wilcox
J
~
Figure 5-1.
BOC 5 (4 EFPD) Two-Dimensional Relative Power Distribution, Full Power, Equilibrium Xenon.
Normal Rod Positions (Group 8 Inserted) 8 9
10 11 12 13 14 15 H
0.80 1.05 i.00 1.20 1.11 1.23 1.09 0.83 K
0.95 1.23 1.03 1.21 1.03 1.24 0.81 8
L 1.00 1.09 0.81 1.15 1.03 0.61 I
M 1.04 1.19 1.06 0.86 i
l N
1.06 1.03 0.58 0
0.58 l
P R
X Inserted Rod Group No.
X. r(
Relative P wer Density l
l 5-5 Babecek & Wilcox
~
6.
THEREU.-HYDRAULIC DESIGN The fresh batch 7 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The four 'oacch 7 axial blanket LTAs have been analyzed to ensure _that they are not the limiting assemblies during cycle 5 operation.2 The thermal-hydraulic evaluation of cycle 5 utilized the methods and models described in references 5, 6, and 11.
The reference cycle for' thermal-hydraulic evaluation is cycle 3.
The differ-ence between cycles 5 and 3 is in the number of unplugged guide tube assem-blies (open assemblies).
Cycle 3 contained 108 open assemblies compared to cycle 5, which was designed with 40 BPRAs and 68 open assemblies. A decrease in the number of open assemblies decreases the core bypass flow, which in-creases the effective core inlet flow and results in a higher calculated min-tsum DNBR.1 Therefore, the thermal-hydraulic analysis performed for cycle 3 is applicable to and conservative for cycle 5.
Cycle 3 was also the reference cycle for cycle 4 Table 6-1 summarizes the maximun design conditions for cycles 4 and 5.
The magnitude of the rod bow penalty applied to cycle 5 is 12.
This necessary l
l burnup-dependent DNBR rod bow penalty applicable to cycle 5, minus a credit t
of 1 for the flow area reduction factor used in the hot channel analysis, re-sults in a net DNBR penalty of zero. All plant operating limits are now based i
I on an original method of calculating rod bow penalties that is more conser-l l
vative than those that would be obtained with new approved procedures.12 For cycle 5 operation, utilization of the new procedure results in a 107. DNBR mar-gin relative to the design DNBR limit value.
r l
l l
r 6-1 Babcock & Wilcox t
p.ble 6-1.
Maxi =t=x Oesign Conditions, Cycles 4 and 5 Cycle 4 Cycle 5 Design power level,.Ta' 2772 2772 Syste= pressure, psia 2200 2200 Reactor coolant flow, % design 104.9 104.9 Vessel inlet / outlet c;1 ant te=p, 100% pcwer.'F 557.7/606.3 557.7/606.3 Ref desim radial-local pcwer peaking factor 1.71 1.71 Ref design axial flux shape 1.5 ces w/ tails 1.5 cos w/ tails Hot channel factors Enthalpy rise (F )
1.011 1.011 q
Heat flux (F")
1.014 1.014 Flew area 0.98 0.98 Active fuel length (a)
(a)
Avg heat-flux, 100 power, Stu/h-f t 1.9E05(b)
- 1. 9E05 (b)
- 4. 941.05 (b) 4.94E05(D) 2 Max heat flux,100% pcwer, Stu/h-ft CEF correlation 3G-2 3&-2 MDN3R 1,74 (112%)
1.74 (112%)
(a)See Table 4-1 of reference 6 the cycle 4 reload report for Rancho Seco.
(b)*'ith thernally expanded fuel rod CD of 0.43075 inch.
a 6-2 Babcock & Wilcox
7.
ACCIDENT ED TRANSIENT ANALYSIS 7.1.
General Safety Analvsis s
Each FSAR accident analysis has been examined with respect ta changes in cycle 5 para =eters to determine the effect of the cycle 5 reload and to ensure that thermal perfor=ance during hypothetical transients is not degraded. The ef-fects of fuel densificatien on the FSAR accident results have been evaluated-and are reported in 3AW-1393.ll Since the cycle 5 parameters are conservative with respect to the reference 11 report, the conclusions in that reference are still valid.
The radiological dose consequences of the accidents presented in Chapter 14 3
of the FSAR were re-evaluated for this reload raport because, even though the F!iAR dose analyses used a conservative t.tsis for the amount of plutonium fissioning in the core, i=provements in fuel management techniques have in-creased the amount of energy produced by fissioning plutonium.. Since plutonium-239 has different fission yields than uranium-235, the mixture of fission prod-uct nuclides in the core changes slightly as the Pu: 23sU fission ratio 23s L
changes, i.e., plutonium fissions procece core of some _nuclides and less of i
others.
Sisce the radiological doses associated with each accident are im-l pacted to a different extent by each nuclide and by various mitigating factors l
l and plant design features, the radiological consequences of the FSAR accidents 1
were recalculated using the specific parameters applicable to cycle 5.
The
(
bases used in the dose calculations are identical to those presented in the l
l FSAR excepc for the following two differences:
l 1.
The fission yields and half-lives used in the new calculations are based on = ore current data.
2.
The steam generator tube rupture accident evaluation considers the increased a=ount of steam ieleased to the environ =ent through the main stean re'.lef and atmospheric dump ve'.ves because of the slower depressuri a';1on due to the reduced hest transfer rate caused by tripping of :he reactor coolant pu=ps upon actuation of high-pressure inj ection (a post-TMI-2 modification).
l l
7-1 Babcock & Wilcox
A comparison of the radiological doses presented in the FSAR to those calcu-laced,specifically for cycle 5 shows that some doses are slightly higher and some are slightly lower than the FSAR valt:es.
However, all doses are well below the 10 CFR 100 limits of 300 Rec to the thyroid and 25 Rem to the whole body.
The small increases in some doses are eraentially of fset by reductions in other doses. Thus, the radiological impacts of accidents during cycle 5 are not significantly different than those described in Chapter 14 of the FSAR.
7.2.
Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: core thermal, thermal-hydraulic, and kinetics parameters, including tha reactivity feedback coef ficients and control rod worths.
Core thermal properties used in the FSAR accident analysis were design operat-s ing values based on calculational values plus uncertainties. First-core values of core thermal parameters and subsequent fuel batches are compared to those used in cycle 5 analyses in Ta? le 4-2.
The cycle 5 thermal-hydraulic maximon design conditions are compared o cycle 4 values in Table 6-1.
These param-etern are common to all the acc. dents considered in this report.
A comparison of the key kineries parameters f rom the FSAR and cycle 5 is provided in Table 7-2.
Cycle 5 parameters include the effects of removing the orifice rod as-semblies.
A generic LOCA analysis for a B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103).1" This analysis is generic since the limiting values of key param-eters for all plants in this category were used.
Furthermore, the combination of average fuel temperatures as a function of LHR and lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to shose calculated for this reload. Thus, the analysis and the LOCA limits re-ported in BAW-10103 and substantiated by reference 15 provide conservative re-suits for the operation of the reload cycle. Table 7-1 shows the bounding values for allowable LOCA peak LHRs for Rancho Seco cycle 5 fuel.
The basis for two sets of LOCA limits is provided in reference 16.
7-2 Babcock & VVilcox
It is concluded from the examination of cycle 5 core thermal and kinetics properties, with respect to acceptable previous cycle values, that this core-reload will not adversely affect the ability of the Rancho Seco plant to op-erste safely during cycle 5.
Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 5 is considered to be bounded by previously accepted analyses. The initial condi-s tions for the transients in cycle 5 are bounded by the FSAR, the fuel densi-fication reporell, and/or subsequent cycle analyses.
Table 7-1.
Bounding Values for Allowable LOCA Peak Linear Heat Rates Allowable Allowable Core peak LER, peak LHR, elevation, first 50 EF?D,
balance of ft kW/ f t evele kW/ft 2
14.5 15.5 4
16.1 16.6 6
17.5 18.0 8
17.0 17.0 10 16.0 16.0 f
l 7-3 Babcock &iWilcox
Table 7-2.
Comparison of Key Parameters for Accident Analysis - Rancho Seco Cvele 5 FSAR and Predicted densification cycle 5 report value value BOC Doppler coeff, 10-5 (aW k)/*F
-1.22
-1.53 EOC Doppler coeff, 10-5 (ak/k)/
- F
-1.37
-1.70 B0" moderator coeff,10-" (AWk)/'F
+0.9
-1.23 E00 moderator coeff,10-" (ak/k)/*F
-3.0
-2.92 All rod group worth. (HZP), % Ak/k 11.1 8.61 Initial boron conc.(HFP), ppm 1425 1026 Inverse baron reactivity worth (HFP), pps/1% aWk 100 118 Max ejected rod worth (HFP), % aWk 0.65 0.37 Dropped rod worth (HFP), % ak/k 0.65 0.20 t
I l
I l
l l
l l
i l
l l
7-4 Babcock & Wilcox
8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 5 operation *
-ccount for changes in power peaking and control rod worths inherent with ansi-tion to the 15-month L3P fuel cycle.
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. The following pages contain the revisions _to previca Technical Specifications.
i i
i l
I i
l l
l 8-l Babcock & Wilcox
Figure 8-1.
Core Protection Safety Limits, Reactor Power Imbalance (Cycle 5)
Tech. Spec. Figure 2.1-2 l
Thermal Power Level, ",
- 120 t
(-34,112)
(35*112)
ACCEPTA8LE
- 110 WRVE 1 4 PUMP OPERATION
- 100 l
0
(- 47,90 )
(-34.86.85)
(35,86.85) l ACCEPTABLE CURVE 2 3,4 PUMP 80 OPERATION 70 j
(-34,58.88) 60 (35,58.88)
ACCEPTA8LE CURVE 3 2,314 PUMP
_ 50 OPERAT10N
- 40
(-47,36.88)
(46,36.88)
- 30 j
l l
20 UNACCEPTABLE UNACCEPTABLE l
OPERATION OPERATION l
_ 10 i
i I
I f
f I
I I
I l
1
-60
-50
- 40
-30
-20
-10 0
10 20 30 40 50 60 Reactor Power Imualance, ",
CURVE REACTOR COOLANT DESIGN FLOW. Epm 1
387,60P 2
288,3h 3
!87,986 8-2 Babcock & Wilcox
Figure 8-2.
Protective System Maximum Allowable Setpoints, Reactor Power Imbalance (Cycle 5)
Tech. Spec. Figure 2.3-2 Thermal Power Level, 5
- 120
(-16,108) 10 8
- 110 (16,108)
CURVE i l
Mj = 1.0
_ 10 0 l
M2 = -1.0
!55 l
$gj
(
(- 32,92) l 5?s l
l
_ 90 I
l l
l' (-16,80 35) 80 (16,80.35) l CURVE 2 1
1 I
l
- 70 l
l "I
=
333 I
(32,64.35)
(- 32,64. 35)
_ 60 l
l l $75 l
i I(-16,52,38)*~
(16,52.38) l l
50 l
l 1
CURVE 3 I l
l l
_ 40 I
(32, 36. 38)
(-32, 36. 38)
I l
l
,=
l l
E l
~
l UN ACCEPT A8L E l Ea@
l
^
^
l OPERATION l
l * ", *
- 20 l
l OPERATION l
4l il l
l 10 i.:
I?
i nl ln in n
v i
i " li
[l I"~t
~
i t
t i
i l
-60
-50
-40
-30
-20
-10 0
10 20 30 40 50 60 Reactor Power Imaalance,5 CURVE RC DESIGN FLOW. m m I
387,600 2
288,374 3
187,986 8-3 Babcock & Wilcox
Figure 8-3.
Rod Index Vs Power Level for Four-Pu=p' Operation, 0-60 EFPD (Cycle 5)
Tech. Spec. Figure 3.5.2-1 110-l l
l l l i l l l l
l 105 (197.102) /
(283.102) 95
{
l (280,92) 90 1/
I
/
I 55 l
QPERATION
'/
75 NOT ALL0wCD
!l i
I
/
l.
/
[
RESTRICTED
[
65 60 I
/
/!
I 55 50 (122,50) l I
i/I j 9
)*'* '
45 l !
40 35
/
I 30 f
j l
M OPERATIOR i
r ALLOWED j
15
, [
(60,85)
I l
S
,o
( 0. 6. g) 0 l
l l
l 1Il-1 I
1 i
I l l
0 20 40 60 30 100 120 140 150 183 200 220 240 260 280 300 Roo inden i
0 25 50 75 i
100 0
25 50 75 000
)
8 Ant 5 Sant 7 0
25 0
75 100 I
l 8-4 Babcock & Wilcox P
y
Figure 8-4.
Rod Index Vs Power Level for Four-Pump Operation From 50 to 250 EFFD (Cycle 5)
Tech. Spec. Figure 3.5.2-2 110 l i i l i i I i I l ! '
I i i l I i j l 1 ! i i j j I j I I II l I i i l i Iiiiiiil I (25s. 02)
I
( 2 so. i o'2 )
!IFi !!! I i i i i Iii/
l :
I I i i i i i
i i I
i i I I I I i !Ii1 i ! i i i I
/
l ;
30 II I I ! III l' l.3 ! ! I 'I I!!'! /!
I / (2.e.s2)
J a5 ll I l!! !
M 8 / (270.a0)
GeEnation i l iiii!
1 T ALL "t3
- !'If' l) 75 I I l ii l l i l I i
t t '/ I y
/I i '
i
! I I i I l 1 l l l ! j i i
f!/
/: 1 I i l l
I i 1 i i l l l l l Iil l l ! /i i I/ l I I i i i iilil i i l !
i i i > >, i if !i/iii l
l hI l 1 l l
(116.50 )
[
I l [l (22.50) i/t Mi.;I ll
'j l! I I
i ;
2 l I l i l I I
II IfI If l !I I iI !1
~
l i l i l l I l
1 1/ i M ! I i i lI i l 1 i
' ' i 1
- i [
l ! I i j i I i,XiiIiI i i i I
N'"0" 7
i i
!Tfii l
1 -
25 ALLOWED l l
I i
//T l l l 1 i i
i o is.is) fl i Ii i I j i l i j !
l IM i
l I l l l l I I I I 1 !
(0.6.6)', M{lFI l
! ' I i l l i l i i l i I j i I ! l l l 1 l i i I I i l i i i i i
I i i i i i i i i 1 ;
I I i 1 l 1 i i i l i l i I I i l 1 i i O
2C 40 60 80 100 120 140 160 ISO 200 220 250 260 250 300 I
toe feces o
25 50 75 100 0
2'5 50 75 100 S Au f 5 SAnt 6 SANr 7 D
2'5 50 75 800
/
i i
l l
l 1
i I
i l
l l
t 8-5 Babcock & Wilcox
Figure 8-5.
Rod Index Vs Power Level for Four-Pu p Operation After 230 E7PD With APSRs Out (Cycle 5)
^-Tech. Spec. Figure 3.5.2-3 110 It iliiI I i I ! ' ; i i ! !
i I I t i !
i I i i ii!I I I i i ! ! i ! I i (2**.502)I['
I (272.sy2)
I l i I i liI l l t i i i l i i I e i t
i l i t i I i i l'l i I ( !J l i liI '/
I
- 267.321 Il l l Il l l I I Ll I l ! ! lIII If i !! IIiI GeftAriON Ii ii f i t/
(2 0,,03 i
l l } } I l l l NOT ALL0 SED j{ l l } l /} /
,, l j 7,
1 1 IIIIi i i.,,,
i ! l i i 1/i /i i i ! I i !
7, f,,
! I Iil l I i i i l i i i i i i I f fi ! i j
i i i I i i i ! !
I I i i i i i /fi i ! t i i ! !
ii "
i i i i i liI iiliiiiff!'
i !
i i 1.!
i I i i ! I I I iil I (17. 50) fl I !
Li i i i I hl l i i k l l l l I Ie i[l l l OPERATI0n I ! $
i I I I
i !
l I I I I l ff 1 I i i ALLosE0 i i i I
i i IIiil if I i l }
i 4
Il i l i liil if i I I Iil i I j l iII I i I
25 j l I
I VI l l l i i lil i i i i l i i l I l l l
.(96,15) [
l l l [ l l l l l l l l l l l l [
',l i
l I i l !M!
I f f 1 I ! ! ? i ! ' ' ! '
f i
l i l
l i i f i
i l f l
5 M 01 I I L i i liIIillil l I II i i lI i i I i i l-1 I I I I I I l l l 1 i lIiiiliiiii! ! I 1 0
20 to 60 80 100 120 140 140 140 200 220 240 260 230 300 too inser 0
25 50 75 ICO O
$5 54 75 800 SAar 5 lANC 6 gggg 7 l
0 25 50 75 100 l
l' l
l 8-6 Babcock & Wilcox
Figure 8-6.
Rod Index Vs Power Level for Three-Pump Operation, 0 to 60 EFPD (Cycle 5)
Tech. Spec. Figure 3.5.2-4 110 105 I
I I i iI (is7,iO2)
)
1 J
(270 'o2)
I I
[
s5
/
I I f I
I/
/
85 I
/
OPER ATION
/
. l l
Il 7,
l l/
l l l l j /j l l
"," Au0",ED g
gA
.ig 7ijiig 3
l l
/l RESTRICTE,0 l l
[.(232.64) l
.5 I
I i
/
IV I
i I
f I i
/
fi i
(i22.50),f f
Il l I/
/
III I
i i$
J
/
/
ll I
/
/
I 25 OPERATION
/_ /
l 1 l Allowed l I 3,
l
/ '/
i l i i
/V I
20 (60.is)h"7 l
l (0.. ) l I
1 l
0 l l l
l I
l i!!
i i ; i 4
l 0
20 40 60 80 100 120 i40 160 180 200 220 240 260 280 300 Rod indes 0
25 50 75 Ido 2'S 50 75 100 l
sAnn 5 sAnr s.
eAmr 7 O
25 50 75 100 i
I 1
i l
l l
I i
1 i
l 8-7 Babcock & Wilcox I
Figure 8-7.
Rod Index Vs Power I.evel for Three-Pump operation From 30 to 250 ETPD (Cycle 5)
Tech. Spec. Figure 3.5.2-5 110
. l l 1 l l l l l l l l 1 Iil i i l i i i i l i i i j l
,,5 l I I II I i il l t Iil i i I
(259. 02)
(270.102)
I i l i I
l Il l } l l I e i i
~
i i i l l
l l l 1 ! l I l i l I/ /1 i '
l-II I IIf I I
[
I I !
OrtaArion i /p[/!
I!
lI l l nor Attomes
' I I i !I I
/!
I i I !
I !
I i i i i ! i ! !I i
75 f ro l l
} l 1 : I I l I l ! ! I l/[]I I j
I il i i l I
l l !l l
- 1. ! /. / (2:2. 64)
.3 5,,
iiiil l IiiI l I I / l/ l I f i i
I I II I I I II I I III I/!
I IiII I I u
I !
lI i l i i iii (is4.50) _MilI i i l i #
5.
,i I
I Il 1
1Iil I !/
i l l
l l !
5 l
I i l II l/l Ii 1
I l I I
I I !
I
! I I
I
!I 35 OrgaAT 0N
!l I
l II I
I l/ l i lIl ALLOWED l i 3,
!I I
I i
I If II I I I!
I '
25 l I I
I l/I l I I I III ITI I I 20 l
l l l
( :s,e5)
Mi i l I! Il !! !!,
I'
'[
t I l l I I l 1 l! i i# i ii l l l l I i
(0. s. 6 ),
, i t l 1l l
l l i I l
l ~l i
Iil lii l
l l l l
I i i o
i liil I i lI I i l Iiliil i 1 l
l I i 1 l
0 20 to 60 80 100 120 150 160 140 200 220 250 260 280 300 tod insen 0
25 75 100 6
25 50 75 100 l
BANK 5 5ANK 6 8ANI 7 0
2 $'
5' 0 75 800 l
l 1
l i
l a-8 Babcock & Wilcox
Figure 8-8.
Rod Index Vs Power Level for Three-Pump Operation, After 230 EFPD With APSRs Out (Cycle 5)
Tech. Spec. Figure 3.5.2-6 110 105 l
I i l {l i I
i l l
- J 25*. 8 02 )
Ii lI Ii!Iif i
I i i
I i I Y I
{
l iI I
II l I/ -! I l
l i
i ii i t!
/
i i i
I I
I I i i I I I '/ I i 70 i
ji Iiiiiiiii
/
IiiI 5
l l l
{
OPERATION l
l
[
l l l l i
! l, I
Ii l I E " ' '"' " ' *!
l!
/
I I d
50 J
(174.50)
I 45 40 OPERATION r
/
ALLOWED
[
l l l l l l l
V i i i i I
l l
/I
!ll t
(96.15)
I
~
i i
,s i
l
/
1 OI 0
1 1 Il 1 1 I i 1 i l l.
i
'I L i l l l-1ii I 0
20 40 6(L 80 100 120 140 160 140 200 220 240 260 280 300 a.e ine.,
0 25 Sh 7'5 ib0 5
25 50
/5 Ib0 l
SAN 5 sAar 7 j
b 25 50 75 io0 I
l i
8-9 Babcock & Wilcox g
---wy-p-e w
r-w r-
~
Figure 8-9.
APSR Withdrawal Vs Power Level, 0-60 EFPD (Cycle 5)
Tec. Spec. Figure 3.5.2-7 102 (6,102)
(29,102)
RESTRICTED REGION 100 1
90 (6,92)
(30,92)
(0,80)
(37,80) 80
]
70
_5 60 o
(100,50)
\\
t
=
l 5
PERMISSIBLE OPERATING REGION l
30 l
l 20 10 0
I I
I I
I I
I I
i l
0 10 20 30 40 50 60 70 80 90 100
,tPSR Witndrawal, ",
l 8-10 Babcock & Wilcox
w Figure 3-10.
APSR '41thdrawal Vs Power Level, 50-250 EFPD (Cycle 5)
Tech. Spec. Figure 3.5.2-8 110 (6,102)
(32,102)
RESTRICTE0 100 REGION (38,92)
_ j (6,92) 90 80 (0,80)
(45,80) e
]
70
_5 E 80 E
(100,50) 50 PERill SSI BLE g
OPERATING d'
40 REGION 30 l
(
20 10 O
i i
l 1
i i
i i
i 0
10 20 30 40 50 60 70 80 90 100 APSR fitnarawal, 5 S-11 Babcock 8. Wilcox l
l
Figure 8-11.
Core Icbalance Vs Pcwer Level, O to 60 EFFD (Cycle 5)
Tech. Spec. Figure 3.5.2-9 110
(- 13, !C2 )
(23,102)
RESTRICTED 100 7
REGION 90
(- 14,9 2)
(23,92) 80
(.26,80)
(25,80)
E "g
70
=
=
OPERATING li 50
--(- 5 0, 50 )
REGION (50,50) 5 a.
40 30 20 10 0
i i
i i
-50
-40
-30
-20
-10 0
to 20 30 40 Core Imaalance, 5 8-12 Babcock & Wilcox
^
~
Figure 8-12.
Core Isbalance Vs Power Level, 50 to 250 EFFD With APSRs In (Cycle 5)
Tech. Spec. Figure 3.5.2-10 110
(-18,102)
(23,102-)
100 RESTRICTE0 REGION
(- 21,92)
(23,92) 90 80
(-32,80)
(26,80) 70 a
E 60 50
(-50,50)
(50,50)
=
PERill SSI BLE J
OPERATING 40 REGION o.
30 20 t
10 O
I I
I I
i i
i I
l 1
-50
-40
-30
-20
-10 0
10 20 30 40 50 Core Imaalance, "
l l
l 8-13 Babcock & W'!cox
Figure 8-13.
Core Iabalance Vs Pever Level, Af ter 230 EFFD
' lith APSRs Cut (Cycle 5)
Tech. Spec. Figure 3.5.2-11 110
(-27,102)
( 14. 10 2) 100 r
RESTRICTED REGION
(-28.92)
(!8.32) 90
(-28,30)
(28,80) 80 70 2
60 o
(-50,50)
(32,50) a 0
PERNISSISL E OPERATING l
.E 40 REGION t
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Figure 8-14.
LOCA Limited Maxi =us Allowable Linear Heat Rate Tech. Spec. Figure 3.5.2-12 21 I
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9.
SIARTUP PROGRAM - PHYSICS TESTING The planned startup test program associateo with core performance is outlined below. These tests verify that core performance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of the unit.
9.1.
Precritical Control Rod Trip Test Preeritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptable criteria state that the rod drop time f rom f ully withdrawn to 75% inserted shall be less than 1.66 seconds a,e the conditions above.
It should be noted that saf ety analysis calculations are based on a rod drop time of 1.40 seconds from f ully withdrawn to two-thirds inserted. Since the most accurate position indication is obtained from the zone reference switch at the 75%-inserted position,. this position is used instead of the two-thirds inserted position for data gathering. The acceptance criterion of 1.40 seconds corrected to a 75%-inserted position (by rod insertion versus time correlation) is 1.66 seconds.
9.2.
Zero Power Physics Tests
- 9. 2.1.
Critical Boron Concentration l
l Criticality is obtained by deboration at a constant dilution rate. Once cri-ticality is achieved, equilibrium boron is obtained and the critical boron j
concentration determined. The critical boron concentration is calculated by l
correcting for any rod withdrawal required in achieving equilibrium boron.
The acceptance criterion placed on critical boron concentration is that the actual boron concentration must be within 1100 ppm of the predicted value.
j 9.2.2.
Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot sero power rod insertion limit. The 9-1 Babcock & Wilcox l
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b avarage coolant temperature is varied by first decreasing and then increasing temperature by 5'F.
During the change in temperature, reactivity feedbactt is compensated by discrete change in rod motion; the change le r2 activity is then calculated by the summation of reactivity (obtained from reactivity calcula-tion on a strip chart recorder) associated with the temperatur change.
Ac-certance criteria state that the measured value shall not diff6r from the predicted value by more than 20.4 x 10 " (ak/k)/*F (predicted value obtained
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from Physics Test Manual curves).
l The
- terator coefficient of reactivity is calcul ted in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain the moderator coefficient. This value muJ-not be in excess A
of the acceptance criteria limit of +0.9 x 10~" (ak/k)/*F.
- 9. 2. 3.
Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6: and 7) are measured at hot -
zero power conditions using the boron / rod swap methoc.
This method consists of establishing a deboration rate in the reactor coolant system and compen-pating for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 in incremental steps.
The reactivity changes that occur during these measurements are calculated based on Reactimeter data, and dif-ferential rod worths are obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each of the controlling groups are then sunned to obtain integral rod group worths. The acceptance criteria for the control bank group worths are as follows:
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1.
Individual bank 5, 6, 7 worth:
predicted value - m,easured value x 100 < 15 measured value 2.
Sum of groups 5, 6 and 7:
predicted value - measured value x 100 4 10 measured value 9-2 Bab:ock 3 Wilcox
9.2.4.
Ejected Control Rod Reactivity Worth After CRA groups 7, 6, and 5 - have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth obtained by adding the incremental changes in reactivity by boration.
After the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is swapped in versus the controlling rod g'aup and the worth determined by the change in the previously calibrated controlling rod group position. The boron swap and rod swap values are aver-agad and error-adjusted to determine ejected rod worth.
Acceptance criteria for the ejected rod worth test are as follows:
1.
predicted value - measured value x 100
< 20 measured value 2.
Measured value (error-adjusted) 51.0% ak/k The predicted ejected rod worth is given in the Physics Test Menual.
- 9. 3.
Power Escalation Tests 9.3.1.
Core Power Distribution Verification at %40, 75, and 100% FP With Nominal Control Rod Position Core power distribution tests are performed at 40, 75, and 100% full power (FP). The test at 40% FP is essentially a check on power distribution in the core to identify any abnormalities before escalating to the 75% plateau. Rod index is established at a nominal full-power rod configuration at which the core power distribution was calculated.
APSR position is established to pro-vide a core power imbalance corresponding to the imbalance at which the core power distribution calculations were performed.
The following acceptance criteria are placed on the 40% FP test:
1.
The worst-case maximun LHR must be less than the LOCA limit.
2.
The minimum DNBR must be greater than 1.30.
3.
The value obtained from extrapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30, or the extrapolated value of imbalance must.f all outside the RPS power /i= balance / flow trip envelope.
9-3 Babcock 3.Wilcox
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The value obtained from the extrapolation of the worst-case maxi-mum '3R to the next power plateau overpower trip setpoint must be less than the fuel melt limit, or the extrapolated value of im-balance must f all outside the RPS power / imbalance / flow trip enve-lope.
5.
The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.
6.
The highest measured and predicted radial peaks shall be within the following ILnits:
predicted value - measured value x 1GO s8 measured value 7.
The highest measured and predicted total peaks shall be within the followicg limits:
87redicted value - measured value f
measured value x 100 s 12 i
Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and ther-mal calculational models, thereby verifying the acceptability of data from these models for input to safety evaluations.
Items 3 and 4 establish the criteria whereby escalation to the next power pla-ceau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR and LHR.
The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test except that core equilibrium xenon is established prior to the 75 and 100% FP tests. Accordingly, the 75 and 100% FP measured peak acceptance criteria are as follows:
1.
The highest measured and predicted radial peaks shall be within the following itmits:
predicted value - measured value x 100 5
measured value 2.
The highest measured and predicted total peaks shall be within the following limits:
predicted value - measured value x 10 0 7.5 measured value 94 Babcock & Wilcox
,.; i 9.3.2.
Incore Vs Encore Detector I= balance Correlation Verification at N40% FP Imbalances are set up in the core by control rod positioning.
Imbalancas are read sinultaneously on the incore detectors and excore power range detectors for various imbalances. The excore detector offset versus incore detector of fset slope must be at least 1.15.
If this slope criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.
9.3.3.
Temperature Reactivity Coefficient at N100% FP The average reactor coolant temperature is decreased and then increased by about 5'F at constant reactor power. The reactivity associated with each temperature change is obtained from the change in the controlling rod group position. Controlling rod group worth is measured by the fast insert / withdraw method. The temperature reactivity coefficient is calculated f rom the mea-sured changes in reactivity and temperature.
The acceptance criteria state that the moderator temperature coefficient shall be negative.
9.3.4.
Power Doppler Reactivity Coefficient at %100% FP Reactor power is decreased and then increased by about 5% FP.
The reactivity change is obtained from the change in controlling rod group position.
Control rod group worth is measured using the fast insert / withdraw method. Reactivity corrections are made for changes in xenon and reactor coolant temperature that occur during the measurement. The power Doppler reactivity coefficient is calculated from the measured reactivity change, adjusted as stated above, and the measured power change.
The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual. Acceptance criteria state that the coasured value shall be more negative than -0.55 x 10-" (ak/k)/% FP.
9.4.
Procedure for Use When Acceptance Criteria Are Not Met If acceptance criteria for any test are not met, an evaluation is performed before the test program is continued.
This evaluation is performed by site i
9-5 Babcock & '#ilcox
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test personnel with participation by Babcock & Wilcox technical personnel as required.
Further specific actions depend on evaluation results. These ac-tions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for anomalies, or design pesonnel per-forming detailed analyses of potential safety problems because of parameter deviation.
Power is not escalated until evaluation shows that plant safety will not be compromised by such escalation.
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REFERENCES 1
BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.
2 Axial Blanket Lead Test Assembly - Licensing Report, 3AW-1664, Babcock &
Wilcox, March 1981.
2 J. H. Taylor (3&W Licensing) to L. S. Rubenstein (USNRC), t.etter Septem-ber 5, 1980.
J. H. Taylor (3&W Licensing) to V. J. Stello (USNRC), Letter, October 29, 1980.
5 Rancho Seco Nuclear Station. Unit 1, Final Safety Analysis Report,
Sacramento Municipal Utility District (Docket No. 50-312).
8 Rancho Seco Nuclear Generating Station, Unit 1 - Cycle 4 Reload Report, BAW-1560, Babcock & Wilcox, Lynchburg, Virginia, August 1979.
7 J. H. Taylor to S. A. Varga, Letter, "BPRA Retainer Reinsertien,"
January 14, 1980.
8 Program to Determine in-Reactor Ferformance of B&W Fuels - Cladding Creep-Collapse, BAW-10034A, Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, October 1978.
C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, 3AW-10044, Babcock & Wilcox, May 1972.
l'
" Classification and Selective Loading of Fuel for Rancho Seco," Letter Report, Babcock & Wilcox, Lynchburg, Virginia, December 1973 (Proprietary).
11 Rancho Seco Unit 1 - Fuel Densification Report, 3AW-1393, Sabcock & Wilcox, Lynchburg, Virginia, June 1973.
10-1 Babcock & Wilcox
12 D. F. Ross and D. G. Eisenhut (NRC) to D. S. Vassallo and K. R. Goeller (NRC) Memorandum, " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors,"
December 8. 1976.
13 L. S. Rubenstein (NRC), to J. H. Taylor (B&W), Letter, " Evaluation of Interim Procedure for Calculating DNBR Reduction due to Red Bow,"
October 18, 1979.
1" ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 1, Babcock & Wilcox, September 1975.
Is J. H. Taylor (B&W Licensing) to R. L. Baer (Reactor Safety Branch, USNRC).
Letter, July 8, 1977.
18 J. H. Taylor (B&W Licensing) to L. S. Rubenstein (USNRC), Letter, September 5, 1980.
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