ML20028F617
| ML20028F617 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1982 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20028F609 | List: |
| References | |
| BAW-1758, TAC-49576, NUDOCS 8302020316 | |
| Download: ML20028F617 (37) | |
Text
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RANCHO SECO UNIT 1 CYCLE 6 RELOAD REPORT l
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BAW-1758 October 1982 i
L RANCHO SECO UNIT 1 CYCLE 6 RELOAD REPORT 1
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BABCOCK & WILCOX Nuclear Power Group
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Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 j
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Babcock & Wilcox l
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CONTENTS Page 1.
INTRODUCTION AND
SUMMARY
1-1 2.
OPERATING HISTORY 2-1 3.
GENERAL DESCRIPTION 3-1 4.
FUEL SYSTEM DESIGN.
4-1 4.1.
Fuel Assembly Mechanical Design 4-1 4.2.
Fuel Rod Design 4-2 4.2.1.
Cladding Collapse 4-2 4.2.2.
Cladding Stress 4-2 4.2.3.
Cladding Strain 4-3 4.3.
Thermal Design..
4-3 4.4.
Material Design 4-3 4-4 4.5.
Operating Experience.
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 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 9.3.1.
Core Power Distribution Verification at 440, 75, and 100% FP With Nominal Control Rod Position.
9-3
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Bahk &Wilcox
CONTENTS (Cont'd)
Page 9.3.2.
Incore.Vs Excore Detector Imbalance Correlation Verification at 440% FP.
9-5 9.3.3.
Temperature Reactivity Coefficient at N100% FP... -
9-5
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9.3.4.
Power Doppler Reactivity Coefficient at 4100%-FP.
9-5 9.3.5.
Core Symmetry Test 9-5
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9.4.
Procedure for Use When Acceptance Criteria
{1
'Are Not Met
'9-6 REFERENCES A-1 List of Tables Table 4-1.
Fuel Design Parameters and Dimensions, Rancho Seco Cycle 6 4-5 4-2.
Fuel Performance Design Parameters.
4-6 5-1.
Physics Parameters, Rancho Seco Cycles 5 and 6..
5-4 5-2.
Shutdown Margin Calculation, Rancho Seco Cycle 6.......
5-5 6-1.
Maximum Design Conditions, Cycles 5 and 6 6-2 7-1.
Comparison of USAR and Cycle 6 Accident Doses 7-3 7-2.
Comparison of Key Parameters for Accident Analysis --
' Rancho Seco Cycle 6 7-4 7-3.
Bounding Values for Allowable LOCA Peak Linear Heat Rates 7-4 List of Figures Figure 3-1.
Core Loading Diagram, Rancho Seco Cycle 6 3-2 3-2.
Enrichment and BOC Burnup Distribution, Rancho Seco Cycle 6 Off a 295-EFPD Cycle 5 3-3 3-3.
Control Rod Locations and Group Designations, Rancho Seco Cycle 6.........................
3-4 3-4.
BPRA Concentration and Distribution, Rancho Seco Cycle 6...
3-5 4-1.
Mark B5 Upper End Fitting, Side View..
4-7 4-2.
Holddown Spring Retainer..
4-8 4-3.
Mark B5 Fixed Control Component Spider.
4-9 4-4.
Mark B5 Fixed Control Component Spider, Side View Section 4-10 4-5.
Mark B5 Fixed Control Component Spider / Upper End Fitting / Reactor Internals Interaction 4-11 4-6.
Gray Axial Power Shaping Rod..........
4-12 5-1.
BOC 6 Two-Dimensional Relative Power Distribution --
Full Power, Equilibrium Xenon, Group 8 Inserted 5-6 5-2.
LTA Axial Power at 31 EFPD, Measured Vs FLAME-Calculated...
5-7
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Babcock &Wilcox
l Figures (Cont'd)
Figure Page 5-3.
Cora Measured Radial Relative Power Distribution, Rancho Seco Cycle 5, 31 EFPD.
5-8 8-1.
Core Protection Safety Limits, Reactor Power Imbalance, Cycle 6............,.......
8-2 8-2.
-Protective System Maximum Allowable Setpoints, Reactor Power Imbalance, Cycle 6.........
8-3 8-3.
Rod Index Vs Power Level for Four-Pump Operation, 0 to 60 EFPD, Cycle 6 8-4 8-4.
Rod Index Vs Power Level for Four-Pump Operation, 50 to 305 EFPD, Cycle 6 8-5 8-5.
Rod Index Vs Power Level for Four-Pump Operation Af ter 305 EFPD, Cycle 6 8-6 8-6.
Rod Index Vs Power Level for Three-Pump Operation, O to 60 EFPD, Cycle 6 8-7 8-7.
Rod Index Vs Power Level for Three-Pump Operation, 150 to 305 EFPD, Cycle 6 8-8 8-8.
Rod Index Vs Power Level for Three-Pump Operation After 305 EFPD, Cycle 6 8-9 8-9.
APSR Withdra'wal Vs Power Level, O EFPD to End of Cycle 6 8-10 8-10.
Core Imbalance Vs Power Level, O to 60 EFPD, Cycle 6.....
8-11 8-11.
Core Imbalance Vs Power Level, 50 to 305 EFPD, Cycle 6 8-12 8-12.
Core Imbalance Vs Power Level Af ter 305 EFPD, Cycle 6 8-13 8-13.
(Tech Spec Figure 3.5.2-7, Deleted) 8-14.
(Tech Spec Figure 3.5.2-8, Deleted) l
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- 1. -INTRODUCTION AND
SUMMARY
This; report justifies operation of the Rancho Seco Nuclear Generating Station, Unit 1, cycle 6, at a rated core power of 2772 MRt.
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 6 reactor and fuel parameters related to power capability are summarized in this report and compared to those of cycle 5.
All accidents analyzed in the 3
Rancho Seco USAR have been reviewed for cycle 6 operation and, in cases where cycle 6 characteristics were conservative compared to those of cycle 5, no new analyses were performed.
The Technical Specifications have been reviewed and modified where required for cycle 6 operation. Based on the analyses performed and taking into account the ECCS Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Rancho Seco cycle 6 can be operated safely at its licensed core power level of 2772 MWt.
Retainers will be installed on 24 fuel assemblies containing burnable poison rod assemblies (BPRAs) and on the two fuel assemblies containing regenerative neutron sources.1 The retainers will provide r ositive retention during reactor operation. The effects of continued operatior, without orifice rod assemblies (0 ras) and the addition of the BPRA retainers have been accounted for in the analysis performed for cycle 6.
In addition to changes in cycles 1 through 5, the design for cycle 6 also in-corporated the following changes with respect to previous cycles:
- The design includes 40 axial blanket assemblies, which are identical in mechanical design to the lead test assemblies (LTAs) inserted in cycle 5.
The LTA design is described in reference 2.
1-1 Babcock & Wilcox
. The lunped burnable poison stack height has been shortened from 126 to 117 inches. This design change is discussed in section 5.
The forty batch 8B fuel assemblies will utilize the Mark B5 end fitting, which eliminates the.need' for the retainers that are required for fixed control components in the Mark B4 design.
+ Gray axial power shaping rods (APSRs) will be used for the first time in cycle 6.
The poison sections of the gray APSRs are longer and contain a weaker Inconel absorber than the silver-indium-cadmium APSRs used for previous cycles.
The margin to centerline fuel melt (CFM) for batches 7 and 8 was determined with the TACO 2 code. The CFM margin for the balance of the core was calculated with IAFY3.
Babcock &Wilcox 1-2
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2.
OPERATING HISTORY Cycle 5, the current Rancho Seco Unit 1 operating cycle, is the reference fuel
-cycle for the nuclear and thermal-hydraulic analyses performed for cycle 6 op-eration. Cycle 5 achieved initial criticality on May 3, 1981, and power esca-lation bega.1 on May 9, 1981.
No operating anomalies occurred during cycle 5 operation that would adversely affect fuel performance during cycle 6.
Babcock & Wilcox 2-1
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3.
CENERAL DESCRIPTION The Rancho Seco reactor core is described in detail in Chapter 3 of the USAR.8 The core consists of 177 fuel assemblies, 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 6, including the 40 axial blanket and four lead test assemblies, maintain a con-stant nominal fuel loading of 463.6 kg of uranium.2 The undensified nominal active fuel lengths, theoretical densities, fuel and fu.1 rod dimensions, and other related fuel parameters may be found in Tables 4-1 and 4-2 of this report.
Figure 3-1 is the core loading diagram for Rancho Seco cycle 6.
Thirteen once-2ss burned batch I assemblies with an initial enrichment of 2.01 wt %
U will be reloaded into the core.
Batches 5B, 6, and 7, with initial enrichments of 3.04, 2ss, respectively, will be shuffled to new locations.
3.21, and 3.14 wt %
U Four of the batch 7 assemblies are axial blanket LTAs loaded in cycle 5.
The 16 batch 5B assemblies are being carried over for a fourth burn. Feed batch 8 consists of 64 assemblies to be located in a checkerboard pattern. The 24 as-semblies of batch 8A are spare cycle 5 assemblies with an initial enrichment of 3.14 wt % 2ss. Batch 8B comprises 40 axial blanket assemblies at 3.43 U
2ss wt %
U (Figure 3-2).
Reacti'ity is controlled by 61 full-length Ag-In-Cd control rods, 64 burnable poison rod assemblies (BPRAs), and soluble boron shim.
In addition to the full-length control rods, eight gray axial power shaping rods (gray APSRs) are pro-vided for additional control of the axial power distribution. The cycle 6 lo-cation of the 69 control rods and the group designations are indicated in Fig-ure 3-3.
The core locations of the total pattern (69 control rods) for cycle 6 are identical to those of the reference cycle described in the Rancho Seco cycle 5 reload report." The group designations, however, differ between cycle 6 and the reference cycle. The cycle 6 locations and LBP concentrations are shown in Figure 3-4.
LBP clusters are used in all 64 batch 8 assemblies.
3-1 Babcock s.Wilcox
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Figure 3-1.
Core Loading Diagram, Rancho Seco Cycle 6 FUEL TRANSFER CANAL I
x I
7 7
6 7
7 A
K6 M4 R6 M12 K10 l 7 7
6 88 6
88 6
7 I
8 K4 89 P12 E8 P4 K12 g
6 88 6
8A 6
8A 6
8A 6
b5 6
C RID M2 K1 K15 P14 L1 7
88 58 88 58 8A 6
8A 58 88 58 A8 7
0 09 813 86 88 810 83 07 7
6 88 10 88 10 BA 10 88 10 88 6
y E
K2 811 E5 C7 C9 E11 85 gle LTA (CY1)
(CY1)
(CY1)
ICV 1) 7 6
RA 58 88 6
88 58 38 6
88 58 OA 6
7 F
F9 N14 F2 812 K8 N3 F14 N2 F7 7
88 6
8A 10 88 6
6 6
88 10 8A 6
88 7
g D11 A9 G3 N13 R8 84 G13 A7 D5 (CY1)
(CY1) 6 6
8A 6
8A 58 6
10 6
58 8A 6
8A 6
6 Y
W.H L15 M5 H13 M9 M15 g H1 H7 H3 M11 F1 7
88 6
8A 13 85 6
6 6
88 10 8A 6
88 7
K N11 R9 K3 C12 A8 03 K13 R7 M5 (CY1)
(CY1) 7 6
8A 58 88 6
88 58 88 6
88 58 8A 6
7 L
L9 Die L2 013 G8 C4 L14 D2 L7 6
88 10 88 10 84 10 88 10 88 6
y 7
P11 M5 87 29 M11 P5 M
E2 (CY1)
(CY1)
(CY1)
(CY1) ffk 7
88 58 88 58 8A 6
8A 58 88 58 88 7
N M9 C13 P6 C8 P1D C3 N7 6
88 6
8A 6
8A 6
8A 6
88 6
0 F15 E2 G1 GIS E14 A6 3
7 6
88 6
88 6
I 7
p
'I G4 812 MB 84 G12 g
7 7
6 7
7 R
06 E4 A1D E12 G1D I
Z 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 ous Cycle Location Cy1 - Reinserted from Cycle 1. LTA - Axial Blanket Lead Test Assembly 3-2 Babcock & Wilcox
i Figure 3-2.
Enrichment and BOC Burnup Distribution, Rancho Seco Cycle 6 Off a 295-EFPD Cycle 5 8
9 10 11 12 13 14 15 2.01 3.21 3.04 3.14 3.21 3.14 3.21 3.21 H
18,058 17,643 25,000 0
21,945 0
22,151 14,619 3.21 3.43 2.01 3.14 3.21 3.43 3.14-19,278 0
1G,698 0
17,447 0
12.417 3.21 3.43 3.04 3.14 3.21 3.14 b
19,271 0
21,233 0
14,940 12,947 2.01 3.43' 3.21 3.14 M
16,280 0
17,714 12,132 3.04 3.43 3.14 N
18,648 0
12,701 3.21 0
14,616 P
R 235 X.XX Initial Enrichment, wt %
U XX,XXX BOC Burn.'o, mwd /mtU 3-3 Babcock & Wilcox
Figure 3-3.
Control Rod Locations and Group Designations, Rancho Seco Cycle 6 X
A 8
4 6
4 C
2 5
5 2
D 7
8 7
8 7
E 2
5 1
1 5
2 F
4 8
6 3
6 8
4 G
5 1
4 4
1 5
l H
W 6
7 3
3 3
7 6
Y K
5 1
4 4
1 5
L 4
8 6
3 6
8 4
M 2
S 1
1 5
2 N
7 8
7 8
7 0
2 5
5 2
P l
4 6
4 R
1.
I 2
3 4
5 6
7 8
9 10 11 12 13,14 15 x
Group Number Group No. of Rods Function 1
8 Safety 2
8 Safety 3
5 Safety 4
12 Safety 5
12 control 6
8 Control 7
8 Control 8
8 APSRs Total 69 3-4 Babcock & Wilcox
Figure 3-4.
BPRA Concentration and Distribution, Rancho Seco Cycle 6 j
8 9
10 11 12 13 14 15 l
H 0.80*
1.10*
K 1.10 1.10 0.50 L
1.10 1.10 0.50*
M 0.80*
1.10 1.10 N
1.10 1.10 0.20 i
0 1.10*
0.50*
0.20 0.50 R
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X.X BPRA Concentration, wt % bgg in Al 03 2
l f
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Spare Cycle 5 BRPAs Babcock &)Milcox 3-5
4.
FUEL SYSTEM DESIGN 4.1.
Fuel-Assemb'ly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for Rancho -
Seco cycle 6 are listed in Table 4-1.
Batch 7 includes four axial blanket LTAs. All fuel assemblies are mechanically interchangeabic into any core po-sition.- The analyses and justification for the LTAs are reported in reference 2.
Retainer assemblies will be used on 24 fresh batch 8A assemblies and 2 regenera-elve neutron sources. Some of the retainers will be inserted for their third cycle of irradiation. The justification for the design and use of the retainers is described in references 1 and 5.
- Eight gray APSR assemblies (APSRAs) will be.used in the Rancho Seco cycle 6 design. Gray APSRAs have not been used in previous cycles butcare identical' to previous APSRAs for all external and interface dimensions. Batch 8B utilizes the Mark B5 type of assembly. The difference between this assembly. and all other Mark B types is the method used to retain fixed control components (BPRAs, orifice rod assemblies, and regenerative neutron source components) during re-actor operation.
The Mark B3/B4 design uses a ball-lock coupling to attach the-fixed control components to the fuel assembly. A spring-loaded retainer assembly ensures-positive holddown force at all design flow conditions.
The Mark B5 upper end fitting (Figure 4-1) provides four open slots that align and allow designed movement of the holddown spring -and its retainer (Figure
~
4-2) and the new Mark B5 fixed control component spider (Figures 4-3 and 4-4).
The holddown spring is preloaded through a stop pin welded to an ear on each side of the upper end fitting.- Incore, as shown in Figure 4-5, the spider feet are captured between the holddown spring retainer and the upper grid pads on the reactor internals. This arrangement retains the fixed control components at all design flow conditions. The Mark B5 upper end fitting has been tested 4-1 Babcock &Wilcox
extensively, both in air and in over 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of aimulated reactor environ-ment, to determine analytical input and to assure good incore performance.
4.2.
Fuel Rod and Gray APSR Design The batch 8 fuel differs from that of batches 5B 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 to 0.3686 inch.
These combined changes were implemented to improve fuel performance. The mechanical evaluations of the fuel rod and gray APSR are discussed below.
4.2.1.
Cladding Collapse The batch 5 fuel is more limiting in cycle 6 than batches 1, 6, 7, or 8 be-cause of its previous incore exposure time. The batch 5 assembly power his-tories were analyzed to determine the most limiting four-cycle power history for creep collapse. The worst :ase power history was then compared against a generic analysis to ensure that creep ovalization will not affect fuel perfor-mance during cycle 6.
The generic analysis was performed based on reference 6 and is applicable to the batch 5 design.
The creep-collapse analysis predicts a collapse time of more than 35,000 ef-fective full power hours (EFPH), which is longer than the maximum expected residence time of 27,800 EFPH (Table 4-1).
The gray APSRs that are to be used in cycle 6 were designed to improve creep life. Cladding thickness and rod ovality control, which are the primary fac-tors controlling the creep life of a stainless steel material, have been im-proved to extend the creep life of the gray APSR. Minimum design cladding thickness of the Mark B APSR is 18 mils, while that of the gray APSR is 24 mils. The ovality in the gap area will also be controlled to tighter toler-ances. The gray APSR is shown in Figure 4-6.
4.2.2.
Cladding Stress The cycle 6 stress parameters are enveloped by a conservative-fuel rod stress analysis. The same method was used for the analysis of cycles 5 and 6.
The gray APSR design was analyzed to determine whether it meets specified de-sign requirements. The APSR was analyzed for cladding stress due to pressure, temperature, and ovality.
It was found that the gray APSR has sufficient cladding and weld stress margins.
4-2 Babcock & Wilcox
4.2.3.
Cladding Strain The fuel design criteria specify a Itmit of 1.0% on cladding plastic tensile circumferential strain.
The pellet is designed to ensure that cladding plas-tic strain is less than 1% at the design local pellet burnup and heat genera-tion rate.
The design' burnup and heat generation rate are higher than the J
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 den-sity and the lower tolerance value for the cladding ID.
The gray APSR was analyzed for cladding strain due to thermal and irradiation swelling. The results of this analysis showed that no cladding strain is in-l duced due to thermal expansion or irradiation swelling of the Inconel absorber.
4.3.
Thermal Design 4
All fuel in the cycle 6 core is thermally similar. The design of the batch 8B axial blanket assemblies (and that of the batch 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 cycle 6 thermal analyses represent a change in analytical method in that analyses for the incoming 7
batch 8A and 8B fuel have been performed with the TACO 2 code, using the analy-sis methodology described in reference 8.
This methodology uses nominal, un-densified input parameters as provided in Table 4-2.
Densification effects are accounted for in the TACO 2 densification model. The TACO 2 analyses apply equally to the batch 7 fuel since this fuel is identical in design to that of batch 8A.
Reinserted fuel (batches ID, 5B, 6) continues to be supported by TAFY3 analyses performed for prior cycles.'
The thermal design evaluation results for the cycle 6 core are summarized in Table 4-2.
Linear heat rate capabilities are based on centerline fuel melt (CFM) with core protection limits based on a 20.4 kW/f t LHR to CFM. The TACO 2 analyses performed for batches 7, 8A, and 8B demonstrate that 20.4 kW/ft is a conservative CFM limit for this fuel. Twelve of the thirteen reinserted batch 1D assemblies have CFM ratings less than 20.4 kW/ft to CFM.18 These assemblies 4
will be selectively loaded to minimize peaking. This will achieve a design minimum of 25.0% margin in peaking compared to the limiting assembly. The maximum fuel rod burnup at EOC-6 is predicted to be less than 40,000 mwd /mtU for a batch 5B fuel rod. The fuel rod internal pressure has been evaluated i,
4-3 Babcock &Wilcox
with TAFY3' for the highest-burnup fuel rod and is predicted to be less than the nominal reactor coolant system pressure of 2200 psia.
4.4.
Material Design The chemical compatibility of all possible fuel-cladding-coolant-assembly in-teractions for batch 8 fuel assemblies is identical to that of the present fuel.
4.5.
Operating Experience Babcock & Wilcox operating experience with the Mark B 15 by 15 fuel assembly has verified the adequacy of its design. As of May 31, 1982, the.following experience has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly:
Max FA burnup(")
Cumulative net MWdM electrical output, g Current Reactor cycle Incore Discharged MWh Oconee 1 7
41,600 40,000 37,367,569 Oconee 2 6
20,565 36,800 34,229,828 Oconee 3 7
19,650 35,072 35,984,594 TMI-1 5
25,000 32,400 23,840,053 ANO-1 5
32,484 33,220 31,843,210 Rancho Seco 5
33,280-37,730 27,752,205 Crystal River 3 4
20,700 29,900 18,359,090 Davis-Besse 1 3
24,617 25,326 11,473,226
(*)As of May 31, 1982.
(b) As of January 31, 1982.
4-4 Babcock & Wilcox
Table 4-1.
Fuel Design Parameters and Dimensions, Rancho Seco Cycle 6 Batch No.
ID SB 6
7 8A 8B
}
Fuel assembly type Mark B3 Mark B4 Mark B4 Mark B4 Mark B4 Mark Is5 I
32 '}
24 40 Number of assemblies 13 16 52 Fuel rod OD, in.
0.430 0.430 0.430 0.430 0.430 0.430 Fuel rod ID, in.
C.377 0.377 0.377 0.377 0.377 0.377 Flexible spacer Corrugated Spring Spring Spring Spring Spring Rigid spacer Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Undensified fuel length 141.75 142.08 141.75 141.80 141.80 141.80 (nom), in.
Pellet OD (mean speci-
'0.3680 0.3697 0.3700 0.3686 0.3686 0.3686 fled), in.
Fuel pellet initial 95.0 94.0 94.0 95.0 95.0 95.0 density (nom), % TD Initial fuel enrichment, 2.01 3.04 3.21 3.14 1.14 3.43 wt % assU Average burnup, BOC, 16,674 21,528 17,671 12,548 0
0 mwd /mtU Cladding collapse time,
>35,000
>35,000
$35,000
>35,000
>35,000
>35,000 EFPil Estimated residence 19,560 27.800 20,900 23,500 24,400 24,400 time, %EFPil at discharge Blanket material NA NA NA 0.2 wt I NA Natural
U uranium
(*} Batch 7 contains four axial blanket lead test assemblies.
)With axial blanket.
4-5 Babcock &Wilcox
Table 4-2.
Fuel Performance Design Parameters Batch No.
1D SB 6
7 8A 8B("}
Number of assemblies 13 16 52 32(b) 24 40 Initial density, % TD 95 94 94 95 95 95 Pellet diameter, in.
0.3680 0.3697 0.3700 0.3686 0.3686 0.3686 Nominal stack height, in.
141.75 142.08 141.80 141.80 141.80 141.80 Enrichment, wt % 23sU 2.01 3.04 3.21 3.14 3.14 3.43 Nominal linear heat rate, 6.20 6.19 6.20 6.20 6.20 6.20 kW/ft at 2772 MWt LHR capability, kW/f t to 20.4(C) 20.4 20.4 20.4 20.4 20.4 CFM Densified Fuel Parameters (
(TAFY3 code analysis only)
Pellet diameter, in.
0.3649 0.3648 0.3651 Fuel stack height, in.
140.7 140.2 140.0 Nominal linear heat rate, 6.25 6.27 6.28 kW/ft at 2772 MWt Average fuel tenperature 1356 1353 1348 at nominal LHR, F (BOL)
Core average LHR = 6.20 kW/ft
(*) Batch 8B is a full-batch implementation of axial blanket assemblies.
( ) Includes four axial blanket lead test assemblies.
(*)12 of the 13 batch ID assemblies, which are rated less than 20.4 kW/ft to CFM, have been selectively loaded.
(d)Densification to 96.5% TD assumed.
4-6 Babcock & Wilcox
Figure 4-1.
Mark B5 Upper End Fitting (Side View)
SLOT STOP E
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Figure 4-2.
Iloiddown Spring Retainer FOOT ARM STOP PIN LEDGE
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A A
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4-8 Babcock & Wilcox
Figure 4-3.
Mark B5 Fixed Control Component Spider 4
FOOT C
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r "Y" ARM LEG h
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STRAIGHT O
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Figure 4-4.
Mark B5 Fixed Control Component Spider Side View Section w
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N N
N N
N N
N N
N SPIER k
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I'igure 4-5.
Mark B5 Fixed Control Component Spider / Upper End Fitting / Reactor Internals Interaction q
w l
REACTOR INTERNALS OPPER GRID PAD STOP ASSEMELY PIN
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1 l
f~n f
r i [
HOLDOWN SPRING RETAINER
(
f L
,) --l H(LD00VD4 SPRING j
7;
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4-11 Babcock & Wilcox
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5.
NUCLEAR DESIGN t'
5.1.
Physics Characteristics-Table 5-1 compares the core physics parameters of design cycles'5 and.6.-
The
_ values for both cycles were generated'using PDQ07. The average cycle burnup will be higher _in cycle 6 than in the design of cycle 5.
The maximum assembly burnup at the end of cycle 6 (345 EFPD) is 36,373 mwd /stU in batch 5B assem-blies in core locations symmetric to H10. This burnt? is less than the prev-ious high of 38,653 mwd /mtU in four batch 4B assemblies at the end of a 295-EFPD cycle 5.
Figure 5-1 shows a representative relative power distribution-for BOC 6 at full power with equilibrium xenon and nominal rod posiffons.
Although cycles 5 and 6 are both feed-and-bleed cycles with an APSR pull near
-EOC, differences between the physics parameters of the two cycles can be at-4 tributed to the higher initial LBP loading, longer design life, and different shuffle pattern for cycle 6.
Calculated ejected rod worths and their adher-ence to criteria are considered at all times in life and at all power levels-in the development of the rod position limits presented in section 8.
The maximum atuck rod worth for cycle 6 is similar to that for the design cycle 5 at both BOC and EOC. All safety criteria associated with these' worths are met.
~
The. adequacy of the shutdown margin with cycle 6 stuck rod worths is demon-strated in Table 5-2.
The following conservatisms were applied for the shut-i down calculations:
1.
10% uncertainty on net rod worth.
2.
Flux redistribution penalty.
Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown margin is presented in the Rancho Seco cycle 5 reload report."
i I
5-1 Babcock &Wilcox
5.2.
Analytical Input The cycle 6 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the reference cycle.
5.3.
Changes in Nuc1 car Design Core design changes for cycle 6 are the use of gray APSRs, "short-stack" LBPs, and the introduction of 40 axial blanket assemblies in addition to the four LTAs loaded in cycle 5.
Gray APSRs, which are longer and use a weaker inconel absorber, replace the silver-indium-cadmium APSRs used in all previous cyclesi Calculations with the standard three-dimensional model verified that these APSRs provide adequate axial power distribution control.
The gray APSRs will be withdrawn from the core during the last part of cycle 6 (320 EFPD). The stability and control of the core in this mode have been ana-lyzed; the calculated stability index without APSRs is -0.033 h-1, which dem-onstrates the axial stability of the core. The stability index for cycle 6 is the same as for cycle 5.
The LBP used with the axial blanket fuel assembly has a 9-inch shorter poison stack than that used with the standard Mark B design, i.e., 117 versu's 126 0 -BSC.
The top 9 inches of the poison stack are replaced by a inches of Al2 3
Zircaloy tubular spacer. This LBP design results in only slight mass reduc-tion versus the standard design, and does not change the dynamic characteris-tics of the LBP. The "short-stack" design asymmetrically positions the burn-able poison stack relative to the fuel column and alters the core axial power shape to create increased " effective maneuvering room" at the beginning of the cycle.
The Rancho Seco cycle 6 core is to contain 44 assemblies incorporating an ax-ial blanket fuel rod design that replaces the top and bottom 6 inches of the enriched fuel column with natural uranium in forty batch 8B assemblies and 23s 0.2 wt %
U in the four batch 7 assemblies. These configurations decrease the neutron axial leakage and increase the power in the central portion of the fuel rod, which in turn increases the worth of the enriched fuel. The net re-sult is more efficient use of the separative work and uranium ore than that associated with a uniformly enriched fuel column design. Because the axial blankets are positioned in regions of little importance and have a low fissile 5-2 Babcock & )Milcox
content, the nuclear characteristics of a core containing axial blanket assem-blies are very similar to those of a 12-inch-shorter standard fuel core. The only significant impact is 6 to 7% higher axial peaking. The radial power peaking distribution and other nuclear parameters are not changed.
Since the only significant change in nuclear characteristics associated with implementation of the axial blanket design is axial power peaking, one of the axial blanket LTAs was monitored during cycle 5 by a movable incore detector to verify the adequacy of current analytical models. Figure J-2 compares two axi-al power profiles for an axial blanket LTA at 31 EFPD in cycle 5.
The solid line is derived from a movable incore detector trace, while the dotted profile is produced using the FLAME computer code. A three-dimensional model of the core with one radial and 24 axial nodes per fuel assembly in a quarter-core geometry is used in the i. LAME calculations. Since the FLAME axial node spac-ing is 6 inches, the explicit modeling of the Inconel intermediate spacer grids is precluded and the grid depressions that appear in the measured axial profile are not shown in the calculated curve.
FLAME predicted the peak axial power in the LTA to within 1.4% and the assembly relative power within 2.5%.
A quarter-core power distribution comparison between FLAME and the fixed detectors at the same time in cycle 5 (see Figure 5-3) also exhibits excellent agreement. The agreement between the measured and predicted axial power shapes and assembly powers is well within the uncertainty assumed for the standard non-blanketed fuel and therefore verifies the accuracy of the FLAME model of the axial blan-ket fuel. Consequently, the same calculational methods and design information were used to obtain the important nuclear design parameters for this cycle.
The axial blanket development program and design features are described more extensively in the BAW-1643 series of reports.ll As shown in Table 5-1, neither the gray APSRs nor the axial blanket assemblies significantly affect the nuclear characteristics of the core.
Babcock & Wilcox 5-3
Table 5-1.
Physics Parameters, Rancho Seco Cycles 5 and 6 *)
I Cycle 5 Cycle 6
~ Cycle length EFPD 275 345 Cycle burnup, mwd /mtU Average core burnup - EOC(c).
9,289 11,654
, mwd /atU 21,738 22,285 Initial core loading, atU 82.1 82.1 Critical boron - BOC, ppm (no Xe)
HZP(d), group 8. inserted 1252 1531 11FP, group 8 inserted 1026 1292 Critical boron - EOC, ppm (eq Xe) groups 1-0 100% wd 55 8
Control rod worths - HFP
, BOC, % ak/ k Group 7 1.52 1.02 Group 8 0.36 0.10 Control rod worths - HFP, 320 ErTD Group 7 1.59 1.06 Crcup 8 0.44 0.10 Max ejected rod worth - HZP, % ak/k BOC, groups 5-8 inserted 0.57 0.52 320 EFPD, groups 5-8 inserted 0.59 0.44 Max stuck rod worth - HZP % ak/k BOC 1.83 1.43 320 EFPD 1.83 1.54 Power deficit, HZP to HFP, % ak/k EOC 1.77 1.72 EOC 2.41 2.49 Doppler coef f - HFP,10 s (ak/k/*F)
BOC (0 Xe, 1300 ppm group 8 ins.)
-1.53
-1.52 EOC (eq Xe, 17 ppm, group 8 100% wd)
-1.70
-1.78 Moderator coeff - HFP,10 " (ak/k/'F)
BOC (0 Xe, 1300 ppm, group 8 ins.)
-1.23
-0.68 EOC (eq Xe, 17 ppm group 8 100% wd)
-2.92
-2.85 Boron worth - HFP, ppm /% ak/k BOC, 1300 ppm 118 126 EOC, 17 ppm 104 109 Xenon worth - HFP, % ak/k 4 EFPD 2.63 2.61 EOC, equilibrium 2.71 2.73 Effective delayed neutron fraction - HFP BOC, group 8 inserted 0.0059 0.0062 EOC, groups 1-8 100% wd 0.0051 0.0053
(*) Cycle 6 data are for the conditions stated in this report.
The cycle 5 core conditions are identified in reference 4.
( ) Based on 220 EFPD at 2772 MWt, cycle 4.
(c)275 EFPD in cycle 5; 345 EFPD in cycle 6 unless otherwise stated.
(d)HZP: hot zero power (532F T,yg, HFP: hot full power (582F T, ).
5-4
Table 5-2.
Shutdown Margin Calculation, Rancho Seco Cycle 6
% Ak/k
% Ak/k
% Ak/k Available rod worth Total rod worth, HZP(a) 8.39 8.89 8.90 Maximum stuck rod, HZP
-1.43
-1.54
-1.55 Net worth 6.96 7.35 7.35 Less 10% uncertainty
-0.70
-0.74
-0.74 Total available worth 6.26 6.61 5.61 Required rod worth Power deficit, HFP to HZP 1.72 2.42 2.49 Max allowable inserted rod worth 0.20 0.35 0.48 Flux redistribution 0.71 1.19 1.20 Total required worth 2.63 3.96 4.17 Shutdown margin (avai]sble worth 3.63 2.65 2.44 minus required worth)
(a)HZP: hot zero power, HFP: hot full power.
Note: Required shutdown margin is 1.00% Ak/k.
Babcock &Wilcox 5-5
l l
Figure 5-1.
BOC 6 (4 EFPD) Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Group 8 Inserted 8
9 10 11 12 13 14 15 H
0.74 1.00 0.95 1.23 1.17 1.27 0.97 0.57 l
K 1.06 1.27 0.93 1.27 1.23 1.17 0.58 8
L 1.18 1.31 1.08 1.29 0.91 0.43 M
0.95 1.30 1.05 0.70 f
1.04 1.06 0.48 N
0 0.57 P
R Inserted Rod Group No.
X.XX Relative Power Density 5-6 Babcock & Wilcox
i l
l Figure 5-2.
LTA Axial Power at 31 EFPD, Measured Vs FLAME-Calculated 150 -
1 L ___,
l l
E 4
.c 8
100 -
s 0
5 s
oo m
w
,\\
ue l
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~
e.
o l
60 -
\\\\
0
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8 2
m
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l a
_____ I O
i i
i 0
0.5 i
1.5 Relative Power Density LEGEND I
MIDS Power, LTA FLAME Power, LTA l
l l
i i
l l
l 5-7 Babcock & Wilcox
Figure 5-3.
Core Measured Radial Relative Power Distribution.
Rancho Seco Cycle 5, 31 EFPD 8
9 10 11 12 13 14 15 0.84 1.06 1.04 1.21 1.10 1.22 1.06 0.78 H
0.85 1.05 1.06 1.25 1.16 1.22 1.08 0.78 1.06 0.99 1.29 1.08 1.23 1.04 1.23 0.77 K
1.05 1.00 1.28 1.09 1.26 1.05 1.25 0.77 1.04 1.29 1.02 1.10 0.84 1.10 1.01 0.57 L
1.05 1.28 1.02 1.10 0.90 1.13 1.04 0.59 i
l 1.21 1.08 1.10 1.08 1.24 1.05 0.81 1.24 1.08 1.09 1.08 1.22 1.05 0.81 1.10 1.23 0.84 1.24 1.05 0.97 0.55 N
1.14 1.25 0.89 1.21 1.04 0.96 0.56 1.22 1.04 1.16 1.05 0.97 0.57 0
1.20 1.03 1.11 1.04 0.96 0.55 1.06 1.20 1.01 0.81 0.55 P
1.04 1.17 1.01 0.80 0.55 0.78 0.77 0.57 R
0.76 0.74 0.57 x.xx Measured (Fixed Rh) x.xx FLAME-Calculated 5-8 Babcock & Wilcox
l l
6.
THERMAL-HYDRAULIC DESIGN l
i The fresh fuel of batches 8A and 8B is hydraulically and geometrically similar to the fuel remaining in the core froc previous cycles. The thermal-hydraulic evaluation supporting cycle 6 operation utilized the methods and models de-scribed in references 3, 4, and 12.
The four batch 7 axial blanket LTAs and all the fuel assemblies of batch 8B are axial blanket assemblies.
The incorporation of the blanket changes the design axial flux shape from a l
1.5 cosine shape to a polynomial-generated shape with a peak of 1.59.
To off-set the decrease in DNBR caused by the change in design axial flux shape, the design radial x local peaking factor (FAH) has been reduced from 1.71 to 1.66.
i The reduction in F is fully supported by the cycle 6 nuclear design, for AH which the maximum predicted radial x local peaking factor is 1.442.
The axial blanket design was analyzed assuming a full axial blanket core and a maximum core bypass flow of 8.2%.
Cycle 6 contains 64 BPRAs and 69 control rod assem-l blies, leaving 42 open guide tube assemblies and an actual bypass flow of 7.8%.
l The reactor core safety limits have been re-evaluated based on the reduced Fg and the 8.2% bypass flow. The cycle 5 limits are applicable for. cycle 6 with a partial axial blanket core and a 7.8% bypass flow. Table 6-1 summarizes the l
maximum design conditions for cycles 5 and 6.
i The magnitude of the rod bow penalty applied to cycle 6 is 1.1%.
This neces-sary burnup-dependent DNBR rod bow penalty applicable to cycle 6, minus a credit of 1% for the flow area reduction factor used in the hot channel analy-l sis, results in a net DNBR penalty of 0.1%.
The rod bow penalty is determined l
l using the value for the highest-burnup assembly in the batch containing the l
limiting assembly in the core. This method for calculating rod bow penalty was approved by the NRC in reference 13.
The limiting assembly is the one l
j that would produce the lowest DNBR. The fresh axial blanket fuel, which con-tains the largest radial x local peak in the core, is the limiting fuel for cycle 6.
i l
6-1 Babcock & Wilcox b
Table 6-1.
Maximum Design Conditions, Cycles 5 and 6 Cycle 5 Cycle 6 Design power level, MWt 2772 2772 System pressure, psia 2200 2200 Reactor coolant flow, % design 104.9 104.9 Vessel inlet / outlet coolant 557.7/606.3 557.7/606.3 temperature @ 100% power, F Ref design radial-local 1.71 1.66 power peaking factor Ref design axial flux shape 1.5 cosine 1.59 polynomial-with tails axial blanket Hot channel factors:
Enthalpy rise (F )
1.011 1.011 q
Heat flux (F")
1.014 1.014 Flow area 0.98 0.98 Active fuct length (a)
(a)
Avg heat flux @ 100% power, 1.9E05
'1.9E05 2
Btu /h-ft Max heat flux @ 100% power, 4.94E05 5.08E05 2
Btu /h-ft CHF correlation BAW-2 BAW-2 Minimum DNBR 1.74 (112%)
1.77 (112%)
(a)See Table 4-2.
6-2 Babcock &Wilcox
7.
ACCIDENT AND TRANSIENT ANALYSIS 7.1.
General Safety Analysis 3
Each USAR accident analysis has been examined with respect to changes in cycle 5 parameters to determine the effect of the cycle 6 reload and to ensure that thermal performance during hypothetical transients is not degraded. The ef-fects of fuel densification on the USAR accident results have been evaluated and are reported in BAW-1393.12 Since the cycle 6 parameters are conservative with respect to the reference 12 report, the conclusions in that reference are still valid.
The radiological dose consequences of the accidents presented in Chapter l'e of the USAR were re-evaluated for this relaod report because, even though the USAR dose analyses used a conservative basis for the amount or plutonium fissioning in the core, improvements 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-239 23s uct nuclides in the core changes slightly as the Pu:
U fission ratio changes, i.e., plutonium fissions produce more of some nuclides and less of others.
Since the radiological doses associated with each accident are im-pacted to a different extent by each nuclide and by various mitigating f actors i
j and plant design features, the radiological consequences of the USAR accidents I
were recalculated using the specific parameters applicable to cycle 6.
The l
bases used in the dose calculations are identical to those presented in the USAR except for the following two differences:
1.
The fission yields and half-lives used in the new calculations are l
based on more current data.
2.
The steam generator tube rupture accident evaluation considers the increased amount of steam released to the environment through the main steam relief and atmospheric dump valves because of the slower i
depressurization due to the reduced heat transfer rate caused by tripping of the reactor coolant pumps upon actuation of high-pressure injection (a post-TMI-2 procedural change).
7-1 Babcock & Wilcox
A comparison of the radiological doses presented in the USAR to those calcu-lated specifically for cycle 6 (Table 7-1) shows that some doses are slightly higher and some are slightly lower than the USAR values. However, all doses are either bounded by the values presented in the USAR or are a small fraction of the 10 CFR 100 limits, i.e., below 30 Rem to the thyroid or 2.5 Rem to the whole body. The small increases in some doses are essentially offset by reduc-tions in other doses. Thus, the radiological impacts of accidents during cycle 6 are not significantly different than those described in Chapter 14 of the USAR.
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 the reactivity feedback coefficients and control rod worths.
Core thermal properties used in the USAR accident analysis were design operat-8 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 the cycle 6 analyses in Table 4-2.
The,;ycle 6 thermal-hydraulic max-imum design conditions are compared to cycle 5 values in Table 6-1.
These parameters are common to all the accidents considered in this report. A com-parison of the key kinetics parameters from the USAR and cycle 6 is provided in Table-7-2.
Cycle 6 parameters include the effects of removing the orifice rod assemblies.
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 ").
This analysis is generic since the limiting values of key param-1 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 those calculated for this reload. Thus, the analysis and the LOCA limits re-ported in BAW-10103 and substantiated by reference 15 provide conservative re-sults for the operation of the reload cycle. Table 7-3 shows the bounding values for allowable LOCA peak LHRs for Rancho Seco cycle 6 fuel. The basis for two sets of LOCA limits is provided in reference 16.
7-2 Babcock 4,Wilcox
lt is concluded from the examination of cycle 6 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 oper-ate safely during cycle 6.
Considering the previously accepted design basis used in the USAR and subsequent cycles, the transient evaluation of cycle 6 is considered to be bounded by previously accepted analyses. The initial condi-tions for the transients in cycle 6 are bounded by the USAR, the fuel densifi-cation report, and/or subsequent cycle analyses.
Table 7-1.
Comparison of USAR and Cycle 6 Accident Doses USAR doses, Cycle 6 doses, Accident Rem Rem Steam generator tube failure Thyroid dose at EAB 9.15 10.49 Whole body dose at EAB 0.152 0.31 Thyroid dose at LPZ 0.559 0.77 Whole body dose at LPZ 0.009 0.022 Steam line break Thyroid dose at EAB 2.54 2.92 Whole body dose at EAB 0.009 0.013 Thyroid dose at LPZ 0.155 0.234 Whle body dose at LPZ
<0.001 0.0009 Fuel handling accident Thyroid dose at EAB 4.76 7.11 Whole body dose at EAB 0.399 1.35 Thyroid dose at LPZ 0.291 0.52 Whole body dose at LPZ 0.025 0.098 Rod ejection accident Thyroid dose at EAB 3.11 2.38 Whole body dose at EAB 0.004 0.006 Thyroid dose at LPZ 0.191 0.57 Whole body dose at LPZ
<0.001 0.007 Loss-of-coolant accident Thyroid dose at EAB 3.28 2.61 Whole body dose at EAB 0.01 0.018 Thyroid dose at LPZ 0.201 0.379 Whole body dose at LPZ
<0.001 0.007 Maximum hypothetical accident Thyroid dose at EAB 137.
103.
Whole body dose at EAB 3.6 3.39 Thyroid dose at LPZ 8.4 13.8 Whole body dose at LPZ 0.229 0.54 7-3 Babcock s.Wilcox
~
.c".',
,(. _
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Table 7-2.
Comparison of Key Parameters for Accident p
Analysis - Rancho Seco Cycle 6
-'g l.;. +
USAR and Predicted
~
-s densification cycle 6 l., * -
report value value
,f 4
BOC Doppler coeff,10-s (Ak/k)/*F
-1.22
-1.52 r..
.. ~
EOC Doppler coeff, 10-5 ( Ak/k)/
- F
-1.37
-1.78
- 7.,,.
BOC moderator coeff,10~" ( Ak/k)/ *F
+0.9
-0.68
.. f EOC moderator coef f,10~" (Ak/k)/*F
-3.0
-2.85 All-rod group worth (HZP), % Ak/k 11.1 8.39
[-]
Initial boron cone (HFP), ppm 1425 1292
~
4-Inverse boron reactivity worth 100 126 (HFP), ppm /1% Ak/k
'.~
Max ejected rod worth (HFP), % Ak/k 0.65 0.27 ja
..g ' -
g<
Dropped rod worth (HFP), % Ak/k 0.65 0.20
- j.. ;') ;
+,,, ' '..( ?
3
~
._1 2 3 ' : !.. '
Table 7-3.
Bounding Values for Allowable
.. '<I. ~.
'i. ;..
[.k,([
LOCA Peak Linear Heat Rates
,g Allowable Allowable
' 4...
-r Core peak LHR, peak LHR, j, 'i.
'?
elevation, first 50 EFPD, balance of ft kW/ f t cycle, kW/ft
[.' : ;
..r.
2 14.5 15.5
.t.
4 16.1 16.6
~f' i 6
17.5 18.0
((
.kk, 8
17.0 17.0 10 16.0 16.0 J.
?.i" -
q.
5
- J '. i. x.
~ ; ~.
,.=;
4.4.,'?.
.. g ;
Y 7..
7-4 Babcock & Wilcox 1
,, ' ' o. T,.
+
i c
8.
PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 6 operation to ac-count for changes in power distribution and control rod worths.
Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria SCCS limits will not be exceeded, nor'will the thermal design criteria be violated. The following pages contain the revisions to previous Technical Specifications.
8-1 Babcock &Wilcox
Figure 8-1.
Core Protection Safety Limits, Reactor Power Imbalance, Cycle 6 (Tech. Spec. Figure 2.1-2)
THERMAL POWER LEVEL, %
- 120
( u,112)
(-35.84.112) 112 ACCEPTABLE
- 110 CURVE 1 4 PUNP OPERATION
- 100
(-47.39,92.84)
(-35.84,88.65) 88.65
- 90 (41.44,88.65)
(45.57,91.54)
ACCEPTABLE CURVE 2 4 & 3 PUMP
- 80 OPERATION
(-47.39,69.49)
- 70 (45.57,68.19)
(-35.84,61.51) 61. 51 (41.44.61.51)
- 60 CURVE 3 ACCEPTABLE 4,3 & 2 PUMP OPERAil0N
- 50
(-47.39.42.35)
- 40 (45.57,41.05)
- 30 UNACCEPTABLE UNACCEPTABLE
- 20 OPERATION OPERATION
- 10 l
I f
I I
I I
I l
i I
t 50 40 30
-20
-10 0
10 20 30 40 50 60 Reactor Power Imaalance, 5 CURVE REACTOR COOLANT DESIGN FLOW,gpm 1
387,600 2
288,374 3
187,986 Babcock & Wilcox 8-2
Figure 8-2.
Protective System Maximum Allowable Setpoints, Reactor Power Imbalance, Cycle 6 (Tech. Spec. Figure 2.3-2)
THERMAL POWER LEVEL, %
- 120
-II
(-18,106) 106 (18,106)
N2 = -1.70 Mi = 1. 452 ACCEPTABLE -100 4 PUMP
@e.
OPERATION
- 90
( -18,7 8. 8.)
78 a
- 80 (18,78.8)
(32,82.2)
ACCEPTABLE
- 70 M CCEPTA8M UNACCEPTABLE 4l OPERATION
- l
(-33.5,56.3)l l0PERATION-60
(-10,51. 4) 51.4 (18,51.4 )
(32,55) 50 1
ACCEPTABLE b
4,3&2
- 40 g
l PUMP OM RATION - 30
(.33.5,28.9) l l
(32,22.6)
,l
- 20 aI sI e
e 7
N l
- 10 l
ii ii i,
i,
-l i;
- i I
i i
I i
i i
li i L, i
-60
-50
-40
-30
-20
-10 0
10 20 30 40 50 60 Reactor Power Imaalance, ",
CURVE RC DESIGN FLOW, gpm 1
387,600 2
288,374 3
187,986 8-3 Babcock & Wilcox
Figure 8-3.
Rod Index Vs Power Level for Four-Pump Operation.
O to 60 EFPD, Cycle 6 (Tech. Spec. Figure 3.5.2-1) 110 (124,102)
(283,102) 100 (278,92) 90 OPERATION (270,80)
NOT ALLOWED 80 RESTRICTED 70 SHUTDOWN MARGIN j
LIMIT 60 u
R 1
(226,50) 50 (77,50)
O
?
40 ca OPERATION ALLOWED 30 a.
20 10 (0,7.4) 0 i.
e
=
i.
i i
=
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Roa Index l
T
=
h b
25
.50 75 150 0
25 50 75 l'00 s.
BANK 5 8ANK 7 3E
?
M
~0 25 50 75 100 BANK 6 L
Figure 8-4.
Rod Index Vs Power Level for Four-Pump Operacion, 50 to 305 EFPD, Cycle 6 (Tech. Spec. Figure 3.5.2-2) 110 100 (223,102)
(283,102) 90 (270,92) 80 PERATION NOT ALLOWED (270,80)
~a1 70 SHUTDOWN g
RESTRICTED 8
MARGIN 60 LIMIT o
50 (157,50)
(226,50) 8*
b u
40 m
30 OPERATION ALLOWED 20 10 0
0,,4. 4 }
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 g
Roa index rr O
I e
a a
a a
a b
0 25 50 75 160 0
25 50 75 160 g
BANK 5 BANK 7 E
O 25 50 75 100 l
N BANK 6
Figure 8-5.
Rod Index Vs Power Level for Four-Pump Operation After 305 EFP9, Cycle 5 (Tech. Spec. Figure 3.5.2-3) i 110 (274,102)
(231,102) f 100 PERATION (270,92) 90 NOT ALLOWED SHilTDOWN 80 y
MARGIN (270,80)
- s LIMIT 7,
N t
S 60 t
50 (158,50)
.(226,50)
[
c I
40 D
30 20 (84,15) 10 (0,,4. 4 ),
0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 F
Roa index Xg i
i x
0 25 50 75 100 0,
25 50 75 100 P
lE BANK 5 BANK 7 0
25 50 75 100 4
BANK 6
k Figure 8-6.
Rod Inder Vs Power Level for Three-Pump Operation, O to 60 EFPD, Cycle 6 (Tech Spec. Figure 3.5.2-4) 110 100 90 5
80 s
('
}
(
}
SHUTDOWN R
70 MARGIN N
LIMIT OPERATION 60 o
NOT ALLOWE0 50 g
RESTRICTED (226,50)
=
m 4
g 40 (77,38) 30 OPERATION ALLOWED 20 - (47,16) 10 (41,11.75)
(0.6.05) 0 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index as eg
(
0 25 50 75 100 0
25 50 75 1$0 BANK 5 BANK 7 h
0 25 50 75 150 N
BANK 6
Figure 8-7.
Rod Index Vs Power Level for Three-Pump Operation, 50 to 305 EFPD, Cycle 6 (Tech. Spec. Figure 3.5.2-5) 110 100 90 E
80 (265,77) 2 OPERATION (223,77) g 70 NOT ALLOWE0 SHUTDOWN MARGIN 60 LIMIT o
ESTRICTE 50 (226,50)
J m
a Eo 40 (157,38)
OPERATION l
30 ALLOWE0 20 (123,25) 10 - (0,3.73)
(88,11.75) 0 i
e i
i i
i i
e 0
20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 F
Rod index X
i i
i i
e i
g w
0 25 50 75 l'00 0
25 50 75 10'0 l
s=
l lE!
8ANK 5 BANK 7 i
i 0
25 50 75 100 l
0 M
BANK 6
l Figure 8-8.
Rod Index Vs Power Level for Three-Pump Operation After 305 EFPD, Cycle 6 (Tech. Spec. Figure 3.5.2-6) 110 100 90 i
E 80 (265.6,77) 2 (231.0,77) i
~
p 70 OPERATION NOT ALLOWED n
}
60 SHUTDOWN MARGIN g'
50 LIMIT (226,50) a a
f 40 (158.0,38) 30 20 OPERATION ALLOWED 10
'~
.(81,11) 0 0
20 40 60 80 100 120 140 16'O 180 200 220 240 260 280 300 F
Roa Inaex o
Xg O
25 50 75 100 0
25 50 75 100 g
BANK 5 BANK 7 5
84 0
25 50 75 100 BANK 6
'm,
Figure 8-9.
APSR Withdrawal Vs Power Level, O EFPD to End of Cycle, Cycle 6 l
11 0 (0,102)
(100,102) 100 90 80 E
70
=
N 60
~o 50 E
E 40 30 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 APSR Witnarawal, ",
Note: There are no withdrawal limits for the gray APSRs in Rancho Seco 1, Cycle 6.
Therefore the APSR withdrawal limits are being deleted from the Tech. Specs.
l 8-10 Babcock & Wilcox
Figure 8-10.
Core Imbalance Vs Power Level, O to 60 EFPD, Cycle 6 (Tech Spec. Figure 3.5.2-9) 11 0 RESTRICTE0
(-10,102) r 100 90 (19,92 )
(-22,80)
(22,80) 80 52 70 PERillSSIBLE E
OPERATING 2
60 REGION I
O 00 s
C 50
(-33,50)
(33,50)
E 2
40 30 l
20 10 l
0 I
I I
I I
I I
l 43
-30
-20
-10 0
10 20 30 40 50 Core Imoalance, ",
i 8-11 Babcock & Wilcox
Figure 8-11.
Core Imbalance Vs Power Level, 50 to 305 EFPD, Cycle 6 (Tech. Spec. Figure 3.5.2-10) 110 -
(-15,102)
(20,102) 100<
RESTRICTED REGION 90
(-27,80)
(23,80) 80 h
PERMISSIBLE 1
70 OPERATING g
N REGION o
60 e
(-41.50)
(41,50) 15 50 o.
40 30 -
20 j
10 8
I I
I 0
-50
-40
-30
-20
-10 0
10 20 30 40 50 Core Imoalance, 5
{
8-12 Babcock & Wilcox
T i
Figure 8-12.
Core Imbalance Vs Power Level After 305 EFPD, Cycle 6 (Tech. Spec. Figure 3.5.2-11) 11 0
(-15,102)
(20,102) 10 RESTRICTED REGION
(-15,92)
(20,92) gg
(-24,80)
PERNISSIBLE (23,80) 80 OPERATING s
REGION 70 n
R n
o 60 e
50
(-41,50)
(41,50) e 40 30 20 10 0
50 30
-20
-10 0
10 20 40 50 60 Core imoalance, ",
8-13 Babcock & Wilcox
r Figure 8-13 (Tech Spec. Figure 3.5.2-7) i a
DELETED 8-14 Babcock & Wilcox
Figure 8-14 l
(Tech Spec. Figure 3.5.2-8)
DELETED 8-15 Babcock & Wilcox
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2.
Control Rod Group and Power Distribution Limits Applicability This specification applies to power distribution and operation of control rods during power operation.
Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejec-tion, and to assure core subcriticality after a reactor trip.
Specification 3.5.2.1.
The available shutdown margin shall be not less than 1% Ak/k with the highest worth control rod fully withdrawn.
If the shutdown margin is less than 1% Ak/k then, within one hour, initiate and continue bora-tion until the required shutdown margin is established.
3.5.2.2.
Operation with inoperable rods:
A.
Operation with more than one inoperable rod as defined in Speci-fication 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted.
B.
If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Spec-ification paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existence of 1% Ak/k hot shutdown margin.
Boration may be initiated to increase the avail-able rod worth either to compensate for the worth of the inoper-able rod or until the regulating banks are fully withdrawn, whichever occurs first.
C.
If within one hour of determination of an inoperable rod as de-fined in Specification 4.7.1, it is not determined that a 1%
Ak/k hot shutdown margin exists combining the worth of the in-operable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is estab-lished.
D.
Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised by a movement until indication is noted but not exceeding 2 inches within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.
E.
If a control rod in the regulating or safety rod groups is de-clared inoperable per 4.7.1.2, power shall be reduc.ed to 60% of the thermal power allowable for the reactor coolant pump combination.
8-16 Babcock & Wilcox
l
}
[
^
RANCHO SECO UNIT 1' TECHNICAL SPECIFICATIONS
/
Limiting Conditions for Operation E.
Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.D above, subsequent reactor. operation is permitted for the purpose of mea-surement, testing, and corrective action provided the thermal power and.
power range high flux setpoint allowable for the reactor coolant pump-combination are restricted by a r' eduction of 2% of maximum allowable power for each 1% tilt, or fraction thereof, for the maximum tilt observed prior to shutdown.
i F.
The quadrant power tilt shall be determined to be within the limits at least once every shift during operation above 15% of rated thermal power-except when the quadrant power tilt alarm is inoperable, then the quadrant power tilt shall be calculated and evaluated at least once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
3.5.2.5.
Control Rod Positions A.
Technical Specification 3.1.3.5 (safety rod withdrawal) does not prohibit the exercising of individual safety rods as required by Table 4.1-2'or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.
B.
Operating rod group overlap shall be 25% 15% between three sequential groups, except for physics tests.
C.
Position limits are specified for regulating control rods. Except for phy-sics tests or exercising control rods, the regulating control rod insertion /
withdrawal limits are specified on Figures 3.5.2-1 through 3.5.2-6. If any of these control rod position limits are exceeded, such that control rod posi-tions are in the restricted region, an acceptable control rod position shall be obtained within two hours.
If control rod positions exceed the shutdown
' margin limit, such that control rods are in the region defined as operation not allowed then, with!' one hour, initiate and continue boration until the required shutdown margis is achie'ved.
D.
Except for physics test, power shall not be increased above the power level cutoff of 92% of the maximum allowable power level unless one of the following conditions is satisfied:
1.
Xenon reactivity is within 10% of the equilibrium value for opera-tion at the maximum allowable power level and asymptotically ap-proaching stability.
2.
Except for xenon-free startup, when 3.5.2.5.D(1) applies, the re-actor has operated within a range of 87 to 92% of the maximum allow-able power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison con-trol node.
8-17 Babcock & Wilcox
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features 5.3.
Reactor Specification 5.3.1.
Reactor Core 5.3.1.1.
The reactor core contains slightly enriched uranium dioxide pellets.
The pellets are encapsulated in zircaloy-4 tubing to form fuel rods.
The reactor core is made up of 177 fuel assemblies. Each fuel as-sembly contains 208 fuel rods.1,2 5.' 3.1. 2. The reactor core shall approximate a right circular cylinder with an equivalent diameter of 128.9 inches and a nominal active height of 144 inches.2 5.3.1.3.
The maximum enrichment of the core for Rancho Seco is a nominal 3.5 weight percent of U235, 5.3.1.4.
There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSR) distributed in the reactor core as shown in FSAR Figure 3.2-45.
The full-length CRA contain a 134 inch length of silver-indium-cadmium alloy clad with stainless steel.
The APSR contain incorrel clad with stainless steel.3 5.3.1.5.
The core may utilize burnable poison assemblies with similar dimen-sions as the full-length centrol rods.
5.3.1.6.
Reload fuel assemblies and rods shall conform to design and evalua-tion described in the USAR.
5.3.2.
Reactor Coolant System 5.3.2.1.
The reactor coolant system shall be designed and constructed in ac-cordance with code requirements."
5.3.2.2.
The reactor coolant system and any connected auxiliary systems ex-posed to the reactor coolant conditions of temperature and pressure, if shall be designed for a pressure of 2,500 psig and a temperature 650*F.
The pressurizer and pressurizer surge line shall be designed for a temperature of 670*F.s 5.3.2.3.
The reactor coolant system volume shall be less than 12,200 cubic feet.
l 8-18 Babcock a.Wilcox
I 9.
STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is outlined below. These-tests will verify that core performance is within the assump-tions of the-sa#ety analysis'and-provide confirmation for continued safe oper-
- s. tion of the unit.
9.1.
Precritical Control Rod Trip Test
-Precritical control rod drop times are recorded for all full-length control
~ rods at hot full-flow conditions before zero power physics testing begins.
~
Acceptance criteria state that the rod drop time from fully withdrawn to 75%
inserted shall be less than 1.66 seconds at the conditions above.
It should be noted that safety analysis calculations are based on a rod drop time of 1.40 seconds from~ fully withdrawn to two-thirds inserted. Since the-nost 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 sec,nds.
-9.2.
Zero Pcwer Physics Tests 9.2.1.
Critical Boron Concentration Criticality is obtained by deboration. Once criticality is achieved, equilib -
rium boron is obtained and the critical boron concentration determined. The critical boron concentration is calculated by correcting for any rod withdrawal required in achieving equilibrium boron. The acceptance criterion placed on critical boron concentration is that the actual borcn concentration must be within 1100 ppm of the predicted value.
9.2.2.
Temperature Reactivity Coefficient
.The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limits. The 9-1 Babcock &Wilcox
ave sge coolant temperature is varied by first decreasing and then increasing the tem, Tature by 5'F.
During the change in temperature, reactivity feedback is compensated by discrete change in rod motion; the change in reactivity is then calculated by the summation of reactivity (obtained from reactivity cal- ~
culation on a strip chart recorder) associated with the temperature change.
Acceptance criteria state-that the measured'value shall not differ from the predicted value by more than 0.4 x 10-" (Ak/k)/*F (predicted value obtained i
from Physics Test Manual curves).
b The moderator coefficient of reactivity is calculated from the temperature coefficient-data. After the temperature coefficient has been measured, a pre-dicted value of. isothermal fuel Doppler coefficient of reactivity [-2.0 x 10-5 l
(Ak/k)/*F) is subtracted to obtain the moderator coefficient. 'This value must' not be in excess of the acceptance criteria limit of +0.9 x 10-" (Ak/k)/*F.
9.2.3.
Control Rod Group Reactivity Worth i
Control bank group reactivity worths.(groups 5, 6, and 7) are measured at hot zero power conditions using the boron / rod swap method. This method consists of establishing a deboration rate in the reactor coolant system and compensat-ing 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 summed to obtain integral rod group worths. The acceptance criteria for the control bank group worths are as follows:
1.
Individual bank 5, 6, and 7 worth:
predicted value - measured value 1
measured value
-< 15 x 100 2.
Sum of groups 5, 6, and 7:
predicted value - measured value measured value
-< 10 x 100 I
I 9-2 Babcock &Wilcox
9.2.4.
Ejected Control Rod Reactivity Worth After CRA groups 7 and 6 have been inserted and group 5 is between 0 and 10%
withdrawn, 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 group and the worth determined by the change in the previously calibrated controlling rod group position. The boron swap value is used to determine ejected rod worth. Acceptance criteria for the ejected rod worth test are as follows:
1.
predicted value - measured value x 100 measured value
-< 20 2.
Measured value (error-adjusted) 51.0% Ak/k The predicted ejected rod worth is given in the Physics Test Manual.
9.3.
Power Escalation Tests 9.3.1.
Core Power Distribution Verification at s40, 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 ar.y abnormalities before escalating to the~75% plateau. Rod index is established at a cominal 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 maximum 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 fall outside the RPS power / imbalance / flow trip envelope.
9-3 Babcock & Wilcox
4.
The value obtained from the extrapolation of the worst-case maxi-mum'LHR to the next power plateau overpower trip setpoint must be less than the fuel melt-limit, or the extrapolated value of im-balance must fall 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 limits:
predicted value - measured value x 100 s8 measured value 7.
The highest measured and predicted total peaks shall be within the following limits:
predicted value - measured value x 100 5 12 measured value 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-teau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR Lnd 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 fellowing limits:
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 7.5 x 100 measured value 9-4 Babcock & Wilcox
9.3.2.
Incore Vs Excore Detector Imbalance Correlation Verification at %40% FP Imbalances, set up in the core by control rod positioning, are read simultan-eously on the incore detectors and excore power range detectors for various imbalances. The excore detector offset versus incore detector offset 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 tem-perature change is obtained from the change in the controlling rod group posi-tion. The worth of the controlling rod group is measured by the fast insert /
withdraw method. The temperature reactivity coefficient is calculated from the measured 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 N100% FP Reactor power is increased and then decreased 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 cal-culated 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 measured value shall be more negative than -0.55 x 10-"(Ak/k)/% FP.
9.3.5.
Core Symmetry Test Core symmetry is measured during the initial escalation to 40% FP.
Data col-lection is initiated when the on-line computer NSS calculations become operable at about 15% FP.
The input from the incore detectors being used in the tilt Babcock & Wilcox 9-5
i calculations will first be checked and verified to be consistent with previous operation. The acceptance criteria state that the core symmetry is acceptable if the absolute value of the quadrant power tilts are less than the error ad-1 justed alarm limit.
l 9.4.
Procedure for Use When Acceptance Criteria Are Not Met l
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 test personnel with B&W technical personnel participating as required. Fur-ther specific actions depend on evaluation results. These actions can include repeating the tests, adding tests to search for anomalies, or performing de-tailed analyses because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be compromised by such escalation.
l I
Babcock &Wilcox 9-6
m.
i I
REFERENCES i
1
~
BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.
2 Axial Blanket Lead Test Assembly - Licensing Report, ~ BAW-1664, Babcock &
Wilcox, Lynchburg, Virginia, March 1981.
8 Rancho Seco Nuclear Station, Unit 1 Updated Safety Analysis Report, Sacramento Municipal Utility Distriat (Docket-No. 50-312),..
" Rancho Seco Nuclear Generating Station, Unit 1 - Cycle 5 Reload Report, BAW-1667, Babcock & Wilcox, Lynchburg, Virginia, March 1981.
8 J. H. Taylor (B&W Licensing) to S. A. Varga (USNRC) Letter, "BPRA Retainer Reinsertion," January 14, 1980.
8 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep-Collapse, BAW-10034A, Rev. 2. Babcock & Wilcox, Lynchburg, Virginia, October 1978.
7 i
Y. H. Hsii, et al., TACO Fuel Pin Performance Analysis, BAW-10141P, Babcock & Wilcox, Lynchburg, Virginia, January 1979.
a J. H. Taylor (B&W) to J. S. Berggren (USNRC), Letter, "B&W's Reciponses to l
TACO 2 Questions," April 8, 1982.
8 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure i
Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia,_ May 1972.
i
" " Classification and Selective Loading of Fuel for Rancho Seco " Letter Report, Babcock & Wilcox, Lynchburg, Virginia, December 1973 (Proprietary).
11
(
M. A. Hannah, Axial Blanket Fuel Design and Demonstration, Semi-Annual Prog-ress Reports, BAW-1643-1, BAW-1643-2, BAW-16!.3-3, November 1980, July 1981, July 1982, Babcock & Wilcox, Lynchbttrg, Virginia.
I-l A-1 Babcock & Wilcox l
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e
-<m-es-
-w--,,
wea
--co--.c em..
---.---y
- w,-
--*me
,y v
12 Rancho Seco Unit 1 - Fuel Densification Report, BAW-1393, Babcock & Wilcox, Lynchburg, Virginia, June 1973.
13 L. S. Rubenatein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of Interim Procedure for Calculating DNBR Reduction Due to Rod Bow," October 18, 1979.
1" ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 1, Babcock
& Wilcox, Lynchburg, Virginia, September 1975.
13 J. H. Taylor (B&W) to R. L. Baer (Reactor Safety Branch, USNRC), Letter, July 8, 1977.
is J. H. Taylor (B&W) to L. S. Rubenstein (USNRC), Letter, September 5,-1980.
A-2 Babcock &Wilcox
-. - _ _ _