ML19209D109
| ML19209D109 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/31/1979 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19209D102 | List: |
| References | |
| BAW-1560, NUDOCS 7910190414 | |
| Download: ML19209D109 (62) | |
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BAW-1560 August 1979 1
RANCHO SECO NUCLEAR GENERATING STATION, UNIT 1
- CYCLE 4 RELOAD REPORT -
PO[jR Oft!8l!'liS ii 3
243 Babcock & Wilcox 7910190 4 14
BAW-1560 August 1979 RANCHO SECO NUCLEAR GENERATING STATION, UNIT 1
- CYCLE 4 RELOAD REPORT --
BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division[} }f4 P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox
ie
I I CONTENT _S-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 g 4.2. Fuel Rod Design 4-1 5 4.2.1. Cladding Collapse 4-1 4.2.2. Cladding Stress 4-2 E 4.2.3. Cladding Strain 4-2 g 4.3. Thermal Design. 4-2 4.4. Material Compatibility. 4-2 4.5. Operating Experience. 4-3 5. NUCLEAR DESIGN. 5-1 5.1. Physics Characteristics 5-1 5.2. Imalytical Input 5-1 5.3. Changes in Nuclear Design 5-2 6. THERMAL-HYDRAULIC DESIGN. 6-1 7. ACCIDEliT AND TRANSIENT ANALYSIS 7-1 7.1. General Safety Analysis 7-1 7.2. Accident Evaluation 7-1 8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS. 8-1 9. STARTUP PROGRAM - PHYSICS TESTING 9-1 E 9.1. Precritical Tests 9-1 5 9.1.1. Control Rod Trip Test 9-1 9.1.2. RC Flow 9-1 g 9.1.3. RC Flow Coastdown 9-1 l 9.2. Zero Power Physics Tests. 9-2 9.2.1. Critical Boron Concentration. 9-2 9.'.2. Temperature Reactivity Coefficient 9-2 9.2.3. Control Rod Group Reactivity Worth. 9-2 9.2.4. Ejected Control Rod Reactivity Worth. 9-3 1173 245 l Babcock & Wilcox
CONTENTS (Cont'd) Page 9.3. Power Escalation Tests. 9-3 9.3.1. Core Power Distribution Verification at N40, 75, and 100% FP With Nominal Control Rod Position 9-3 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at %40% FP 9-5 9.3.3. Temperature Reactivity Coefficient at N100% 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 9-6 REFERENCES............................ A-1 List of Tables Table 4-1. Fuel Design Parameters and Dimensions 4-4 4-2. Fuel Thermal Analysis Parameters - Rencho Seco, Cycle 4 4-5 5-1. Physics Parameters, Rancho Seco Cycles 3 and 4.. 5-3 5-2. Shutdown Margin Calculation - Rancho Seco, Cycle 4. 5-4 6-1. Cycles 3 and 4 Thermal Hydraulic Design Conditic. - Rancho Seco 6-3 7-1. Comparison of Key Parameters for Accident Analysis -- Rancho Seco, Cycle 4 7-3 7-2. Allowable LOCA Peak Linear Heat Rate - Rancho Seco, Cycle 4 7-3 List of Figures Figure 3-1. Core Loading Diagram - Rancho Seco, Cycle 4.......... 3-2 3-2. Rancho Seco BOC 4 Enrichment and Burnup Distribution 3-3 3-3. Centrol Rod Locations - Rancho Seco, Cycle 4 3-4 3-4. Lb> Enrichment and Distribu* ion - Rancho Seco, Cycle 4 3-5 5-1. BOC 4 Two-Dimensional Relative Power Distribution Full Power, Equilibrium Xenon, Normal Rod Positions 5-5 8-1. Core Protection Safety Limits, Reactor Power Imba? 7ce 8-6 8-2. Protective System Maximum Allowable Setpoints, Reactor Fawer Imbalance. 8-7 8-3. Rod Index Vs Power Level for Four-Pump Operation, O to 160 EFPD. 8-8 1173 246 - 111 - Babcock & Wilcox
I Figures (Ccnt'd) Figure Page 8-4. Rod Index Vs Power Level for Four-Pump Operation, 140 to 310 EFFD.. 8-9 8-5. Rod Index Vs Power Level for Four-Pump Operation, 290 to 345 EFPD 8-10 8-6. Rod Index Vs Power Level for Three-Pump Operation, O to 160 EFFD. 8-11 5 8-7. Rod Index Vs Power Level for Three-Pump Operation, W 140 to 310 EFPD. 8-12 8-8. Rod Index Vs Power Level for Three-Pump Operation, 290 to 345 EFPD. 8-13 8-9. APSR Withdrawal Vs Power Level, O to 160 EFPD. 8-14 8-10. APSR Withdrawal Vs Power Level,140 to 310 EFPD. 8-15 8-11. Core Imbalance Va Power Level, O to 160 EFPD 8-16 8-12. Core Imbalance Vo Power Level, 140 to 310 EFPD 8-17 8-13. Core Imbalance Vs Power Level, 290 to 345 EFPD 8-18 E I I I I I I I I 1173 247 I - iv - Babcock & Wilcox I
l. INTRODUCTION AND
SUMMARY
This report justifies operation of the Rancho Seco Nuclear Generating Station, Unit 1, cycle 4, 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 analy-tical techniques and design bases documented in several reports that have beer submitted to the USNRC and approved by that agency. Cycle 4 reactor and fuel parameters related to power capability are summarized in this report and compared to cycle 3. All accidents analyzed in the Rancho Seco FSAR have been re.iewed for cycle 4 operation and, in cases where cycle 4 characteristics were conservative compared to cycle 3, no new analyses were per-fo rmed. The Technical Specifications have been reviewed and modified where required for cycle 4 operation. Based on the analyses performed, taking into account the ECCS Final Acceptance Criteria and postulated fuel deneification effects, it is concluded that Rancho Seco cycle 4 can be operated safely at its licensed core power level of 2772 MWt. Retainersl will be installed on all fuel assemblies containing bornable poison red assemblies (BPRAs) 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 (ORAs) and the addition of the BPRA retainers have been accounted for in the analysis performed for cycle 4. 1173 248 1-1 Babcock & \\Nilcox
I I I I I I I E I I I I E I I I I I i173 2n I
2. OPERATING HISTORY Cycle 3, the current Rancho Seco Unit 1 operating cycle, is the refarence fuel cycle for the nuclear and thermal-hydraulic analyses performed for cycle 4 operation. Cycle 3 achieved initial criticality on December 19, 1978, and power escalation began on December 21, 1978. The 100% power level, 2772 MWt, was reached on December 24, 1978. No operating anomalies occurred during cycle 3 operation that would adversely affect fuel performance during cycle 4. m 1173 250 2-1 Babcock & Wilcox
3. GENERAL DESC9 u' TION The Rancho Seco reactor core is described in detail in Chapter 3 of the FSAR.2 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-The fuel consists of dished-end, cylindrical pellets of ment guide tube. uranium dioxide, clad in cold-worked Zircaloy-4. All fuel assemblies in cycle 4 maintain a constant nominal fuel loading of 463.6 kg of uranium. The undensi-fled nominal active fuel lt-gths, theoretical densities, fuel and fuel rod di-mensions, 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 4. Thirteen 235U once-burned batch 1 assemblies with an initial enrichment of 2.01 wt % 1_ will be reloaded into the core. Batches 4 and 5, with initial enrichments of 3.19 and 3.04 wt % 235U, respectively, will be shuffled to new locations. 235 Batch 6, with an initial enrichment of 3.21 wt % U, will be loaded in a checkerboard pattern. Figure 3-2 is an eignth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 4. 52 LBP clus-Reactivity is controlled by 61 full-length Ag-kn-Cd control rods, ters, and solu' ole boron shim. In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The cycle 4 locations 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 4 are identical to those of the referen< c cycle described in the Rancho Seco cycle 3 reload report.3 The group designations, however, differ between cycle 4 and the reference cycle in order to minimize power peaking. The cycle 4 locations and enrichments of the LBP clusters are shown in Figure 3-4. The nominal system pressure is 2200 psia, and the core average densified nominal heat rate is 6.27 kW/ft at the rated core power of 2772 MWt. i173 251 3-1 Babcock & Wilcox
I Figure 3-1. Core Loading Diagram - Rancho Seco, Cycle 4 Fuel Transfer Canal x 5 5 5 5 5 A R10 R09 R08 R07 R06 B 5 5 5 6 5 6 5 5 5 P12 Pll LO2 N08 Ll4 P05 PO4 l 5 6 4 6 4 6 4 6 4 6 5 013 NO3 M04 M12 N13 003 W IB 1 5 6 4 6 4 4 4 6 4 6 5 (OIk) (ON) Lil Gil NO2 B08 N14 E7 LOS 1B 1B E 5 4 6 004 6 4 6 4 6 D12 6 4 5 M14 C12 (Cv1) K03 K13 revii c04 uo2 F 5 5 6 4 6 5 4 4 4 5 6 4 6 5 5 L15 B10 F10 rog LO4 E9 L12 K07 E06 P06 101 m 1B 1B G 5 6 4 F04 4 4 4 6 4 4 4 F12 4 6 5 g K15 D11 (Cv11 C09 D10 F06 F10 D06 C07 (Cyl) DOS K01 IB H 5 5 6 4 6 4 6 6 4 6 4 6 5 5 g -~Y 9 Kll H14 H04 H01 H15 H12 H02 G5 ccy1) 5 6 4 4 4 4 6 4 4 4 IB 4 6 5 IB G15 N11 009 N10 LO6 L10 N06 007 N05 G01 r 3 ( ) L 5 5 6 4 6 5 4 4 4 5 6 4 6 5 5 FIS P10 M10 G09 F04 M7 F12 G07 MC6 P06 F01 5 4 6 6 4 6 4 6 6 4 5 3 N04 N12 004 E02 E14 012 G3 G13 (Cyll frv13 lb 10 5 6 4 6 4 4 4 6 4 6 5 ( ff) Fil M9 D02 D14 K5 F05 (Cyli P08 5 6 4 6 4 6 4 6 4 6 5 0 C13 D03 E04 E12 D13 C03 5 5 5 6 5 6 5 5 5 p m B12 Bil F02 D08 F14 B05 B04 5 5 5 5 5 R A10 A09 A08 A07 A06 l z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Satch x xxx Previous Core Location I i 11732523 Babcock & Wilcox 3-2
Figure 3-2. Rancho Seco BOC 4 Enrichment and Burnup Distribution 8 9 10 11 12 13 14 15 2.01 3.21 3.19 3.21 3.19 3.21 3.04 3.04 j 18,058 0 23,783 0 20,109 0 13,559 9,022 3.19 3.19 3.19 2.01 3.19 3.21 3.04 g 21,121 23,319 21,365 14,587 19,608 0 8,699 3.04 3.21 3.19 3.21 3.04 3.04 L 12,690 0 21,765 0 11.810 7,335 2.01 3.21 3.19 3.C4 Si 14,768 0 17,639 10,064 3.19 3.21 3.04 3; 23,781 0 7,185 0 3.04 7,935 P R 3 Initial Enrichment, wt % 235U x.xx N xxxxx BOC Burnup, Mb'd/mtU 1173 -253 3-3 Babcock & Wilcox
imm
I Figure 3-3. Control Rod Locatione - Rancho Seco, Cycle 4 5 x ^ 3 B 4 7 4 C 1 6 6 1 D 7 8 5 8 7 E 1 5 2 2 5 1 F 4 8 3 7 3 8 4 G 6 2 a 4 2 6 H - 7 5 7 3 7 5 7 K 6 2 4 4 2 6 L 4 8 3 7 3 8 4 M 1 5 2 2 5 1 N 7 8 5 8 7 0 1 6 6 1 P 4 7 4 R Z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 x Group No. Group Number of Rods Function 1 8 Safety 2 8 Safety 3 5 Safety 4 12 Safety 5 8 Control 6 8 Control 7 12 Control 8 8 APSRs Total 69 3 1173 254 g 3-4 8 bcock & Wilcox
Figure 3-4. LBP Enrichment and Distribution - Rancho Seco, Cycle 4 8 9 10 11 12 13 14 15 H 0.8 0.5 1.10 K 0.8 0.8 L 1.10 0.8 M 0.5 1.10 1.10 N 1.10 0.08 0 1.10 0.8 0.08 P 0.8 I R x.xx LBP concentration, wt % BqC 1173.255 3-5 Sabcock & Wilcox ---mm----s-----
4. FUEL SYSTEM DESIGN J 4.1. Fuel Assembly Mechanical Design The cycle 4 core consists of the normal resident and reload Mark-B fuel assem-blies. The pertinent fuel design parameters and dimensions are listed in Table 4-1. Fifty-two BPRA retainer assemblies will be used on fuel assemblies con-taining BPRAs to provide positive retention during reactor operation. Similar retainer assemblies will be used on the two fuel assemblies containing the regen-erative neutron sources (RNS). The justification for the design and use of the BPRA retainers is described in reference 1, which is applicable to the retainer assemblies of cycle 4 of Rancho Seco. The two RNS retainers which were in-stalled during the last refueling will be replaced with new or relicensed re-tainers. All fuel assemblies are identical in concept and are mechanically interchangeable. 4.2. Fuel Rod Design The fuel pellet end configuration has changed from a spherical dish for batches 1 through 5 to a truncated cone dish for batch 6. This facilitates manufactur-ing while maintaining the same end void volume. The mechanical evaluation of the fuel rod is discussed below. 4.2.1. Cladding Collapse Creep collapse analyses were performed for three-cycle assembly power histories for Rancho Seco. The batch 4 fuel is more limiting than other batches due to its previous incore exposure time. The batch 4 power histories were analyzed and the most limiting assembly history was used for creep collapse analysis. The assembly power history for the most limiting assembly was used to calculate the fast neutron flux level for the energy range >l MeV. The collapse time for the most limiting assembly was conservatively determined to be >30,000 EFPH (effective full-power hours), which is greater than the maximum three-cycle de-sign life (Tabic 4-1). This analysis was performed based on the conditions set forth in reference 4. 1173 256 4-1 Babcock & \\Vilcox
E 4.2.2. Cladding Stre.ss The Rancho Seco stress parameters are enveloped by a conservative fuel rod stress analysis. For design evaluation, the primary membrane stress must be less than 2/3 of the minimum specified unirradiated yield strength and all g stresses must be less than the maximum specified unirradiated yield strength. W In all cases, the margin is in excess of 30%. The following conservatisms with respect to Rancho Seco fuel were used in the analysis: 1. Lower post-densification internal pressure. 2. Lower initial pellet density. 3. Higher system pressure. 4. Higher thermal gradient across the cladding. 4.2.3. Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding tensile circumfer-ential plastic strain. The pellet design is such that the tensile clastic clad-ding strain is le,s than 1% at 55,000 mwd /mtU. The following cladding strain conservatisms are applicable with respect to the Rancho Seco fuel: 1. The maximum specification value for the fuel pellet diameter was used. 2. The maximum specification value for the fuel pellet density was used. 3. The cladding ID used was the lowest permitted specification tolerance. 4. The maximum expected three-cycle local pellet burnup is less thun 55,000 mwd /mtU. 4.3. Thermal Design All fuel assemblies in this core are thermally similar. The design linear heat rate (LHR) capability and the average fuel temperature for each batch in cycle 4 are shown in Table 4-2. LHR capabilities are based on centerline fuel melt and were established usi..g the TAFY-3 codes with fuel densification to 96.5% of theoretical density. The pellet resinter test data from the batch 6 fuel demonstrate that the fuel capability is greater than the design LHR requirement. 4.4. Material Compatibility The chemical compatibility of all possible fuel-cladding-coolant assembly in-teractions for the batch 6 fuel assemblies is identical to that of the present 1173 257 4-2 Babcock & Wilcox
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 March 31, 1979, the following e::perience has been accummulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly: Maximum assembly (") (b) burnup, mwd /mtU Current electrical output, Reactor cycle Incore Discharged MWh Oconee 1 5 35,300 31,100 25,423,997 Oconee 2 3 22,400 33,700 20,311,630 Oconee 3 4 24,100 29,400 22,410,960 TMI-1 4 32,400 9,200 23,880,710 ANO-1 3 33,240 28,300 19,739,776 Rancho Seco 3 31,800 29,378 15,514,225 7,643,351 Crystal River 3 1 16,300 Davis Besse 1 1 8,900 3,750,428 (")As of March 31, 1979. (b)As of February 28, 1979. 1173 258 Babcock & Wilcox 4-3
I Table 4-1. Fuel Design Parameters and Dimensions Batch IB Batch 4 Batch 5 Batch 6 Fuel assembly type Mark B3 Mark B4 Mark B4 Mark B4 No. of assemblies 13 56 56 52 Fuel rod OD, inches 0.430 0.430 0.430 0.430 Fuel rod TD, inches 0.377 0.377 0.377 0.377 Undensified active fuel 141.75 142.08 142.08 141.80 length, inches (nominal) Initial fuel enrichment, 2.01 3.19 3.04 3.21 wt % 235U Expected core residence 25,440 21,144 21,600 21,600 g time, %EFPH g Flexible spacers, type Corrugated Spring Spring Spring Fuel pellet OD (mean 0.3686 0.3697 0.3697 0.3700 specified), inches Fuel pellet initial 93.0 94.0 94.0 94.0 g density, % TD 5 Cladding collapse >30,000 >30.000 >30,000 >30,000 time, EFPH I I I I I I ~ I 1 m 259 I Babcock & Wilcox 4-4
g Table 4-2. Fuel Thermal Analysis Parameters - Rancho Seco, Cycle 4 Batch 1 Batch 4 Batch 5 Batch 6 No. of assemblies 13 56 56 52 Initial density, % TD 95 94 94 94 Pellet diameter, in. 0.3680 0.3697 0.3697 0.3700 Nominal stack height, 141.75 142.08 142.08 141.80 in. Densified Fuel Parameters
- Pellet diameter, in.
0.3649 0.364E(d) 0.3648(d) 0.3651 Fuel stack height, in. 140.7 140.2 140.2 140.0 Nominal linear heat 6.25 6.27 6.27 6.28 rate, kW/' it 2772 MWt Avg fuel temperature 1356 1353 1353 1348 at nominal LHR, F LilR capability, kW/ft 20.4(c) 20.4(b) 20.4 20.4 to centerline fuel melt (* All calculations assume densificetion from the LTL limit to 96 5% TD except for the average fuel temperature calculation, which assumed densification from the nominal density to 96.5% TD. (b)Except six selectively loaded fuel assemblies, which have a linear heat capability of 20.25 kW/ft. The batch 1 fuel assemblies have fuel melt limits from 19.2 to 20.4 kW/ft and are selectively loaded.6 The densified diameter for batches 4 and 5 had been reported as 0.3664 in the cycle 2 and 3 reload reports. This has no affect on any analytical results. 1173 260 Babcock & Wilcox 4-5
5. NUCLEAR DESIGN 5.1. Physics Characteristica Table 5-1 compares the core physics parameters of design cycles 3 and 4. The values for both cycles were generated using PDQ07. The average cycle burnup will be higher in cycle 4 than in the design cycle 3 because of the longer cycle 4 length. Figure 5-1 illustrates a representative relative power dis-tribution for the beginning of cycle 4 at full power with equilibrium xenon and normal rod positions. Although both cycles 3 and 4 are feed-and-bleed cycles with an APSR pull near EOC, differences between the physics parameter of the two cycles can be attrib-uted to the initial LBP loading, longer design life, and different shuffle pattern for cycle 4. Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits presented in section 8. The maximum stuck rod worth for cycle 4 is similar to that for the design cycle 3 at both BOC and EOC. All safety criteria associated with these worths are met. The adequacy of the shutdown margin with cycle 4 stuck rod worths is demonstrated in Table 5-2. The following conservatisms were applied for the shutdown cal-culations. 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 3 reload report.3 5.2. Analytical Input The cycle 4 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the reference cycle. 5-1 Babcock & Wilcox
I 5.3. Cl.anges in Nuclear Design There is only one significant core design change between the reference and re-locd cycles: the increase in design length to 335 EFPD and the subsequent in-corporation of LBP clusters to aid in reactivity centrol. The same calcula-g tional methods and design information were used to obtain the important nuclear W design parameters for this cycle. I B B I I I I I 8 I I I un 252 5 5-2 Babcock & Wilcox
l J J Table 5-1. Physics Parameters, Rancho Seco Cycles 3 and 4(") Cycle 3( g Cycle 4 Cycle length, EFPD 300 335 Cycle burnup, mwd /mtli 10,134 11.316 Average core burnup - EOC, Mwd /mtU 21,769 22,119 Initial core loading, mtU 82.1 82.1 Critical baron - BOC, ppm (no Xe) HZP(d), group 8 inserted 1411 1372 HFP, gr up 8 inserted 1197 1154 Critical boron - EOC, ppm (eq Xe) grou,ps 1-8 100% wd p 3 4 Control rod worths - HFP , BOC, % Ak/k Group 7 1.64 1.56 Group 8 0.41 0.38 Control rod worths - HFP, EOC % Ak/k Group 7, 335 EFPD, group 8 out 1.63 1.52 Group 8, 300 EFPD 0.46 0.46 3 Max ejected rod worta - HZP, % Ak/k BOC, groups 5-8 inserted 0.64 0.76 300 EFPD, groups 5-8 inserted 0.76 0.63 Max stuck rod worth - HZP, % Ak/k BOC 1,79 1.89 EOC 1.78 1.56 f Power deficit, HZP to HFP, % Ak/k p BOC 1.45 1.60 EOC 2.26 2.37 ] Doppler coeff - BOC, 10-5 (Ak/k/'F) j 100% power (0 Xc) -1.45 -1.52 Doppler coeff - EOC, 10-5 (Ak/k/'F) 100% power (eg Xe) -1.56 -1.73 Moderator coeff - HFP, 10 " (Ak/k/*F BOC (0 Xe, 1150 ppm, group 8 ins.) -0.81 -1.05 EOC (eq Xe, 17 ppm, group 8 100% wd) -2.81 -2.89 d Boron worth - HFP, ppm /% Ak/k J BOC, 1200 ppm 112 117 EOC, 17 ppm 100 103 Xenon worth - HFP, % Ak/k BOC, 4 days 2.68 2.66 EOC, equilibrium 2.78 2.77 Effective delayed neutron fraction - ( HFP J BOC 0.0059 0.0060 EOC 0.0052 0.0052 " Cycle 4 data are for the conditions stated in this report. The cyc-le 3 core conditions are identified in reference 3. Based on 290 EFPD at 2 772 MWt cycle 2 (c)300 EFPD in cycle 3; 335 EFPD in cycle 4 unless otherwise stat d e HZP: hot zero power (532F Tavg); HFP: hot full power (582F T avg)'1 17 3, 6 3 2 5-3 Babcock & Wilcox --in. -ee
I E Table 5-2. Shutdown Margin Calculation - Rancho Seco, Cycle 4 BOC, % Ak/k 335 EFPD, % Ak/k 1. Available rod worth I Total rod worth, HZP ") 8.37 8.76 a. b. Maximum stuck rod, HZP -1.89 -1.56 c. Net wo"'. 6.48 7.20 d. Less 10% uncertainty -0.65 -0.72 e. Total available worth 5.83 6.48 2. Required rod' worth a. Power deficit, HFP to HZP 1.60 2.37 b. Max, allowable inserted rod worth 0.30 0.32 c. Flux redistribution 0.39 1.03 d. Total required vorch 2.29 3.72 3. Shutdown margin (available worth 3.54 2.76 minus required worth)
- HZP: hot zero power, HFP: hot full power.
Note: Required shutdown margin is 1.00 % Ak/k. 8 8 8 E I I 1173 264 5-4 Babcock & Wilcox
Figure 5-1. BOC 4 (4 EFPD) Two-Dimensional Relative Power Distribution Full Power, Equilibrium Xenon, Normal Rod Positions (Group 8 Inserted) 8 3 10 11 12 13 14 15 H 0.95 1.27 1.02 1.25 1.05 1.25 1.15 0.74 K 1.06 1.01 1.02 0.87 1.10 1.19 0.71 8 L 1.17 1.20 0.93 1.21 0.99 0.54 'M O.95 1.23 1.09 0.79 N 1.08 1.17 0.60 0 0.72 P R I \\x Inserted Rod Group No. Relative Power Density x.xx L 1173 265 5-5 Babcock & Wilcox
si
6. THERMAL-HYDRAULIC DESIGN The incoming batch 6 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The thermal-hydraulic design evaluation supporting cycle 4 operation utilized the methods and models de-scribed in references 2, 3, and 7 except for the core bypass flow and the in-clusion of retainers to provide positive holddown of burnable poison rod assem-blies (BPRAs) and neutron sources. =-- The maximum core bypass flow due to the removal of all orifice rod assemblies (ORAs) in cycle 3 increased to 10.4%. For cycle 4 operation, 52 BPRAs will be inserted, leaving 56 vacant fuel assemblies and resulting in a decrease in cal-culated maximum core bypass flow to 8.3%. The BPRA retainers introduce a small DNBR penalty, as discussed in reference 1. Reactor core safety limits have been re-evaluated based on the insertion of these BPRAs with retainers and increased core flow. The cycle 3 and 4 maximum design conditions and sig-nificant parameters are shown in Table 6-1. The increase in core flow more than compensates for the decrease in DNBR due to the BPRA retainers so that the cycle 3 analysis is conservative and applicable to cycle 4. A flux / flow trip setpoint of 1.05 has been maintained for cycle 4 operation. This setpoint and other plant operation limits based on minimum DNBR criteria contain a DNBR margin of 10.2% from the design minimum DNBR limit of 1.30 using BAW-2. In response to reference 8, B&W has committed to prepare a topical report ad-dressing the potential for and effects of fuel rod bow. In addition, B&W has submitted an literim rod bow penalty evaluation procedure 9,10 for use until the topical report is completed and reviewed. The rod bow penalty applicable to cycle 4 was calculated using the interim rod bow penalty evaluation procedure. As in the previous cycle the burnup is based on the maximum fuel assembly burnup of the batch that contains the fuel assem-bly with the maximum radial x local peak. For cycle 4 this burnup is 14,537 1173 266 6-1 Babcock & Wilcox
I mwd /mtU, a batch 6 fuel assembly. The calculated penalty using this procedure s less than 0.8%. Utilizing the 1% DNB credit for the flow area reduction factor, the actual penalty applied to the DNB calculations is zero. Er I 5 I I I I I I I I 5 I I \\\\15 261 l 5 6-2 Babcock & Wilcox
Table 6-1. Cycles 3 and 4 Thermal Hydraulic Design Conditions - Rancho Seco Cycle 3 Cycle 4 Design power level, MWt 2772 2772 = System pressure, 17fa 2200 2200 Reactor coolant flow, % design 104.9 104.9 Vessel inlet / outlet coolant 557.7/606.3 557.7/606.3 temp, 100% power, F Ref design radial-local power 1.71 1.71 peaking factor Ref design axial flux shape 1.5 cosine 1.5 cosine with tails with tails tiuc channel factors Euthalpy rise (F ) 1.011 1.011 9 Heat flux (F") 1.014 1.014 Flow area 9 0.98 0.98 Active fuel length See Table 4-1 See Table 4-1 Avg heat flux, 100% power, 1.9E05(a) 1.9E05(") Btu /h-ft2 Max heat flux, 100% power, 4.54E05(") 4.94E05(*) Btu /h-ft2 CHF correlation BAW-2 BAW-2 Minimum DNBR (% power) 1.74 (112%) 1.74 (112%) " With thermally expanded fuel rod OD of 0.43075 inch. 1173 268 Babcock & Wilcox
7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analysis Each FSAR2 accident analysis has been examined with respect to changes in cycle 4 parameters to determine the effect of the cycle 4 reload and to ensure that thermal performance dt ring hypothetical transients is not degraded. The effects of fuel densificatic,a on the ; SAR accident results have been evaluated and are reported in BAW-1393.7 Since :he cycle 4 parameters are conservative with re-spect to the reference 7 report, the conclusions in that reference are still valid. ^ Improved fuel utilization and the inherent increase in core average burnup ex-perienced in cycle 4 have resulted in a higher plutonium-to-uranium fission ratio than that used in the FSAR. A study of the major FSAR Chapter 14 acci-dents using the cycle 4 iodine and noble gas inventories concluded that the thyroid and whole body dases were well below the 10 CFR 100 limits. 7.2. Accident Evaluation The key parameters that have the greatest effect on determining the outco.2e 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 FSAR accident analysis were design operat-ing values based on calculational values plus uncertainties. First-core values 2 of core thermal parameters and subsequent fuel batches are compared to those used in cycle 4 analyses in Table 4-2. The cycle 4 thermal-hydraulic maximum design conditions are compared to cycle 3 values to 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 FSAR and cycle 4 is provided in Table 7-1. Cycle 4 parameters include the effects of removing the orifice rod assemblies. }{g 7-1 Babcock <.Wilcox
I A generic LOCA analysis for the BW 177-fuel assembly, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model.Il This analysis is generic since the limiting values of key parameters for all plants in this category were used. Furthermore, the combination of average fuel temperature as a function of the linear heat rate (LHR) and the lifetime pin pressure data used in the reference 11 LOCA limits analysis are conserva-tive compared to those calculated for this reload. Thus, the analysis and the LOCA limits reported in reference 11 and substantiated by reference 12 provfde conservative results for the operation of Rancho Seco cyclo 4. Table 7-2 shows g the bounding values for allowable LOCA peak LHRs for Ranr.ho Seco cycle 4 fuel. W It is concluded from the examination of cycla o core thermal and kinetics prop-g erties, with respect to acceptable previous cycle values, that this core reload ur will not adversely affect the plant's ability to operate safely during cycle 4. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 4 is considered to be bounded by pre-viously accepted analyses. The initial conditions for the transients in cycic 2 4 are bounded ' y the FSAR, the fuel densification report 7, and/or subsequent cycle analyses. 8 I I I I I I 1173 270 5 7-2 Babcock & Wilcox
Table 7-1. Comparison of Key Parameters for Accident Analysis - Rancho Seco, Cycle 4 FSAR and densification Predicted report value value OC Doppler coeff, 10-5 (Ak/k)/*F -1.22 -1.52 EOC Doppler coeff, 10-5 (Ak/k)/*F -1.37 -1.73 BOC moderator coeff, 10-4 (Ak/k)/*F +0.9 -1.05 EOC moderator coeff, 10-4 (ak/k)/*F -3.0 -2.89 All-rod group worth (HZP), % A'.,/k 11.1 8.37 Initial boron cone (HFP), ppm 1425 1154 Inverse boron reactivity worth (HFP), ppm /1% Ak/k 100 117 Max ejected rod worth (HFP), % Ak/k 0.65 0.39 Dropped rod v)rth (HFP), % Ak/k 0.65 0.20 Table 7-2. Allowable LOCA Peak Linear Heat Rate - Rancho Seco, Cycle 4 Core elevation, Linear heat rate ft limits, kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 i173 271 m Babcock & \\Vilcox 7_3 mm-- - - - --- - - ---si
I I 8. PROPOSED MODIFICATIONS TO TECilNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 4 operation. Changes were the result's of the following: 1. The required quantity and concentration of boric acid necessary to reach a cold shutdown condition have been reviewed and found to be within pre-vious Technical Specification limits. Pages 3-17 and 3-18 of the Technical Specifications have been changed so that the basis is consistent with cycle 4 requirements. 2. The boron concentration during refueling has been set to remove all re-strictions on movement of the control rod assemblies during fuel shuffling. These changes appear on Technical Specification pages 3-44 and 3-45. 3. Tecanical Specification 5.3.1.6 is revised to read "keload fuel assemblies and rods shall conform to Design and evaluations described in the FSAR." The elimination of the 3.2 wt % of 235U defined by the original specifi-cation does not affect any safety evaluations. The real allowable quadrant power tilt limit for cycle 4 remains at the cycle m 3 limit. Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated. Figures 8-1 through 8-13 provide revisions to previous Technical Specification limits. I I I l 11/3 272 ~ m 8-1 Babcock & Wilcox
I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Cond!tions for Operation 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the operational status of high pressure injection and chemical ad-dition systems. Objective To provide for adequate boration under all operating conditions to ensure ability to bring the reactor to a cold shutdown condition. Specification The reactor shall not remain critical unless the following conditions are met: 3.2.1 Two pumps capable of supplying high pressure injection are operable (also see Specification 3. 3.2). 3.2.2 The borated water storage tank and its flow path to the reactor for high pressure injection are operable. 3.2.3 A source of concentrated boric acid solution in addition to the borated water storage tank is available and operable. This requirement is ful-filled by the concentrated boric acid st.orage tank. This tank shall contain at least the equivalent of 10,000 gallons of 7,100 ppm boron. System piping and valves necessary to establish a flow path for high pressure injection shall also be operable and shall have at least the same temperature as the boric acid storage tank. One associated boric acid pump is operable. The concentrated boric acid storage tank water shall not be less than 70F, and at least one channel of heat tracing g shall be operable for this tank's associated piping. The concentrated g boric acid storage tank boron concentration shall not exceed 8,500 ppm Bases The makeup and purification system and chemical addition systems provide con-l trol of the reactor caolant system boron concentration.1 This is normally ac-w complished by using either the makeup pump or one of the two high pressure injection pumps in series with a boric acid pump associated with the concen-g trated boric acid storage tank. The alternate method of boration will be the g use of the makeup or high pressure injection pumps taking suction directly-from the borated water storage tank.2 I The quantity of boric acid in storage from either of the two above-mentioned sources is sufficient to borate the reactor coolant system to a 1 percent sub-critical margin in the cold condition (70F) at the worst time in core life with a stuck control rod assembly. The maximum required is the equivalent of 9105 gallons of 7100 ppm boron. This requirement is satisfied by requiring a min-imum volume of 10,000 gallons of 7100 ppm in the concentrated boric acid 1173 273 3-17 (8-2)
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS storage tank during critical operations. The minimum volume for the borated water storage tank (390,000 gallons of 18G0 ppm boron), as specified in sec-tion 3.3, is based on refueling volume requirements and easily satisfies the cold shutdown requirement. The specification assures that the two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon decay. The quickest method allows for the necessary boron addition in less than one hour. The primary method of adding boron to the primary system is to pump the concentrated boric acid solution (7100 ppm boron, minimum) irto the makeup tank using the 50 gpm boric acid pumps. Using only one of the two boric acid pumps, the required volume of boric acid can be injected in less than 3.5 hours. The alternate method of addition is to inject boric acid from the borated water storage tank using the high pressure injection pumps. Concentration of boron in the cot entrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions. For this reason and to ensure that a flow of boric acid is available when needed, this tank and its associated piping will be kept above 70F (30F above the crystallization temperature for the concentration present). Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures ensure boric acid solubility. The value of 70F is significantly above the crystallization temperature for a so-lution containing 12,200 ppm boron. REFERENCES 1 FSAR subsections 9.2 and 9.3. 2 FSAR Figure 6. 2 '. 3 Technical Specification 3.3. 1173 274 3-18 (8-3)
I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.8 FUEL LOADING AND REFUELING App}icability Applies to fuel loading an/ mfueling operations. _ Objective To ensure that fuel loading and refueling operations are performed in a re-sponsible manner. Specification 3.8.1 Radiation levels in the reactor building refueling area shall be moni-g tored by R15026 and R15027. Radiation levels in the spent fuel storage g area shall be monitored by R15028. If any of these instruments become inoperable, portable survey instrumentation, having the appropriate ranges and sensitivity to fully protect individuals involved in refuel-ing operations, shal.1 be used until the pe:manent instrumentation is returned to service. 3.8.2 Core subcritical neutron flux ahall be continuously monitored by at least two neutron flux monitors, each with continuous indication avail-able, whenever core geometry is being changed. When core geometry is not being changed, at least one neutron flux monitor shall be in service. 3.8.3 At least one decay heat removal pump and cooler shall be operable. I 3.8.4 During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration shall be maintained at not less than 1850 ppm. 3.8.5 Direct communications between the control room and the refueling person-nel in the reactor building shall exist whenever changes in core geome-try are taking place. 3.8.6 During the handling of irradiated fuel in the reactor building at least l one door on the personnel and emergency hatches shall be closed. The equipment hatch cover shall be in place with a minimum of four bolts W securing the cover to the sealing surfaces. 3.8.7 Isolation valves in lines containing automatic containment isolation valves shall be operable, or at least one shall be in a safety features position. 3.8.8 When two irradiated fuel assemblies are being handled simultaneously withih the fuel transfer canal, a minimum of 10 feet separation shall be maintained between the assemblies at all times. Irradiated fuel as-senblies may be handled with the auxiliary bridge crane provided no other irradiated fuel assembly is being handled in the fuel transfer canal. 1 73 275 3-44 (8-4)
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS 3.8.9 If any of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core shall cease; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the re-activity of the core shall be made. 3.8.10 The reactor building purge system, including the radiation monitors, R15001A and R15001B, shall be tested and verified to be operable immed-iately prior to refueling operations. 3.8.11 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours. 3.8.12 No loads will be handled over irradiated fuel stored in the spent fuel pool, except the fuel assemblies themselves. A dead weight load test the rated load will be performed on the fuel storage building han-at dling bridge prior to each refueling. Bases Detailed written procedures will be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment, as described in suhcection 9.7 of the FSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation. Continuous non-itoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain a uni-form boron concentration.1 The refueling boron concentration indicated in Specification 3.8.4 will be maintained to ensure that the more restrictive of the following reactivity conditions is met: 1. Either a keff f 0.95 or less with all control rods renoved from the core. 2. A boron concentration of 1800 ppm. Specification 3.8.5 allows the control room operator tc inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement. The specification requiring testing reactor building purge termination is to verify that these components will function as required should a fuel handling accident occur that resulted in the release of significant fission products. Specification 3.8.11 is required as the safety analysis for the fuel handling accident was based on the assunption that the reactor had been shut down for 72 hours and all 208 fuel pins in the hottest all gap activity.3 fuel assembly fail, releasing 1173 276 3-45 (8-5)
I Figure 8-1. Core I'rotection Safety Limits, Reactor Power Imbalance (Cycle 4) THERMAL POWER LEVEL, % 120 E ( -3'6. 9.112 ) CURVE 1 (+35.8.112) W -110 ACCEPTABLE 4 PUMP -100 OPERATION (+59.5,96.5) ( 47.1,90.5) - 90 (-36.9,84.6) CURVE 2 (35.8,84.6) - 80 ACCEPTABLE 4 & 3 PUMP - 70 OPERATION (+59.5,69.1) ( 47.1.63.7) 60 (35.8,57.4) ( -3 6. 9. 5 7. 4 ) CURVE 3 ACCEPTABLE - 50 4, 3 & 2 PUMP OPERATION (+59.5,41.9) - 40 ( -4 7.1. 3 5. 9 ) - 30 - 20 - 10 t f I t t -60 -40 -20 0 20 40 60 Reactor power imoalance, '; CURVE REACTOR COOLANT DESIGN FLOW, GPM 1 387.600 2 288,374 3 187.986 TECHNICAL SPECIFICATION FIGURE 2.1 -2 I I 1173 277 8 abcod & ECOX 8-6
Figure 8-2. Protective System Maximum Allowable Setpoints, Reactor Power imbalance (Cycle 4) THERMAL POWER LEVEL, % (-26.105) CURVE 1 110 (26,105) @ l ACCEPTABLE100 ,7/ 4 PUMP (49 90) e OPERATION - 90 I (-36,84) CURVE 2 I l (-26,78.1 ) 80 l (26,78.1) l ACCEPTABLE I l 4 & 3 PUMP - 70 l OPERATION I (49,63.1) (-36,57.1) - 60 I g (-26,50. 9) CURVE 3 l(26,50.9) 50 ACCEPTABLE 4,3 & 2 l PUMP - 40 l (49,35.9) l OPERATION ( -36,2 9. 9 ) - 30 l t 20 ,,l $ yj - 10
- l z
l l i i i i -60 -40 -20 0 20 40 60 Reactor Power Imbalance, % CURVE REACTOR C00LANi OESIGN FLOW, GPM 1 387,600 2 288,374 TECHNICAL SPECIFICATION FIGURE 2.3-2 3 187,986 1173 278 Babcock & Wilcox g_7
I Figure 8-3. Rod Index Vs Power Level for Four-Pump Operation, O to 160 EFPD (Cycle 4) 110 105 135,102) (280,102) 100 95 OPERATION NOT (280,92) 90 85 ALLONEO 80 75 (265,80) 70 SHUTDOWN OPERATION 65 LIMIT RESTRICTED 60 55 E 50 (61,50) (232,50) 45 40 35 g 30 I-3 25 OPERATION 20 PERMISSIBLE 15 - (8.15) TECHNICAL SPECIFICATION FIGURE 3.5.2 1 ~ (0,0) 0 t i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 5 Rod index g i I I f I t i t i i g 0 25 50 75 100 0 25 50 75 100 5 Bank 5 t Bank 7 0 25 50 75 100 Bank 6 I I 1173 279 I Babcock & Wilcox 8-8
Figure 8-4. Rod Index Vs Powar Level for Four-Pump Operation, 140 to 310 EFPD (Cycle A) 110 ( 80.102) 105 (187.102) 100 95 OPERATION NOT ( B0.92) 90 ALLOWED 85 80 (265,80) 75 70 SHUTDOWN OPERAT10N 65 LIMIT RESTRICTED 60 t 55 a 50 (109,50) (232,50) = 45 40 35 30 OPERATION 25 PERMISSIBLE ~ (15,15) 15 10 0.10) TECHNICAL SPECIFICATION FIGURE 3.5.2-2 5 (0.0) 0 e i i i i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Roa index i i I i i l 0 25 50 75 100 0 25 50 75 100 Bank 5 Bank 7 0 25 50 75 100 Bank 6 il73 280 J 8-9 babcock & Wilcox
I Figure 8-5. Rod Index Vs Power Level for Four-Pump Operation, 290 to 345 EFPD (Cycle 4) 110 105 (195.102) (275,102) 100 95 OPERATION 90 NOT ALLOWED OPERATION ~ SHUTDOWN RESTRICTED (236.80) 80 LIMIT 75 g 70 W 65 s. 60 P. 10N g 55 PERMISSIBLE g 50 (116,50) 45 40 TECHNICAL SPECIFICATION FIGURE 3.5.2-3 35 g 30 3 25 5 NOTE: THIS FIGURE VAll0 ONLY FOR OPERATION AFTER APSR WITH0RAWAL 10 i,10 5 ~ (0.0) 0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index i i i i I i i i 1 0 25 50 75 100 0 25 50 75 100 Bank 5 Bank 7 0 25 50 75 100 Bank 6 I I un 281 I I Babcock & Wilcox 8-10
Figure 8-6. Rod Index Vs Power Level for Three-Pump Operation, O to 160 EFPD (Cycle 4) 110 - 105 (135,102) (265,102) 100 10N 95 90 NOT ALLOWED 85 OPERATION ~ 80 RESTRICTED 75 70 65 LIMIT 60 (232,64) 0 55 5 50 (61,50) 45 ~ OPERATION 40 ~ PERMISSIBLE 35 30 25 20 15 (8.15) TECHNICAL SPECIFICATION FIGURE 3.5.2-4 10 - ( 0,10 ) 5 t(0.0 0 i t i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod indet i t i i i i i 0 25 50 75 100 0 25 50 75 100 Bank 5 Bank 7 t i 0 25 50 75 100 Bank 6 1173 232 Babcock & Wilcox 8-11
I Figure 8-7. Rod Index Vs Power Level for Three-Pump Operation, 140 to 310 EFPD (Cycle 4) 110, 105 (187,102) (265,102) 100 OPERtTION 0 OT A M O NN 85 RESTRICTED 80 5 SHUToowN 70 LIMIT 65 (232,64) 60 55 (109.50) gg 5 45 40 OPERATION PERMISSIBLE 35 g 30 g 25 20 15 15,15) 10 (0.10) TECl!NICAL SPECIFICATION FIGURE 3.5.2-5 0,0) 0 i i I i i i i t i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index 1 I I I I t i i i 0 25 50 75 100 0 25 50 75 100 Bank 5 Bank 7 0 25 50 75 100 Bank 6 I 1173 283 I I Babcock & Wilcox 8-12
py _ _ Figure 8-8. Rod Index Vs Power Level for Three-Pump Operation, 290 to 345 EFPD (Cycle 4) 110 105 (195.102) (236.102) 100 95 OPERATION g 90 NOT ALLOWED 0 4 85 4 SHUTDOWN 7n LIMIT 65 60 ~ 55 PERMISSIBLE E 50 (116.50) 45 40 TECHNICAL SPECIFICATION FIGURE 2.5 2-6 35 30 25 20 22.1 NOTE: THIS FIGURE (All0 ONLY FOR OPERATION AFTER APSR 15 WITH0RAWAL (0.10 5 .04 i i i 0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index t I i i l t i 0 25 50 73 100 0 25 50 75 100 Bank 5 Bank 6 i i 0 25 50 75 100 Bank 7 1173 284 8-13 Babcock & Wilcox
m-i---um i-
I Figure 8-9. APSR Withdrawal Vs Power Level, O to 160 EFPD (Cycle 4) 110 (6,102) (27.102) 100 90 (6,92) (38,92) RESTRICTED 80 (0,80) (30,80) REGION 70 I = a b 60 50 (100,50) f PE?MISSIBLE 40 j OPERATING 30 REGION 20 TECHNICAL SPECIFICATION FIGURE 3.5.2-7 10 0 l O 10 20 30 40 50 60 70 80 90 100 APSR withdrawal ", I I I I I i173 285 I Babcock a.Wilcox 8-14
11 T M A-Figure 8-10. APSR Withdrawal Vs Power Level, 140 to 310 EFPD (Cycle 4) 110 (6,102) (27,102) 100 N 90 (6,92) (39,92) 80 -( 0. 80 ) (30,80) RESTRICTE0 REGION o ? 70 5 60 O e 50 (100,50) J PERMISSIBLE 5 40 OPERATING a. REGION 30 20 TECHNICAL SPECIFICATION FIGURE 3.5.2-8 10 0 i i i i i e i i i 0 10 20 30 4C 50 60 70 80 SO 100 APSR withdrawal, r, i173 286 8-15 Babcock & Wilcox
I Figure 8-11. Core Imbalance Vs Power Level, O to 160 EFPD (Cycle 4) ~ I RESTRICTED (-18,102) (24,102) 100 REGION ( -20,92 ) l (23,92) 80 (-20.7,80)s W-@ j 70 e 60 I = PERMISSIBLE a 50 (-50,50) (47,50) g OPERATING d W g 40 REGION m. 30 20 ,g TECHNICAL SPECIFICATION FIGURE 3.5.2-9 0 l -60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Core imbalance, % I I I I 1173 287 8 Babcock & Wilcox g_
I I Figure 8-12. Core Imbalance Vs Power Level, 140 to 310 EFPD (Cycle 4) I 110 (-15.5,102) (24,102) 100 RESTRICTED ~ 90 80 (-19,80) (31,80) m l 70 60 I 'o 50 (.50,50) PERMISSIBLE (38,50) a J 0 OPERATING g REGION h 30 20 10 - TECHNICAL :,fECIFICATION FIGURE 3.5.2-10 0- -60 -50 -40 30 -20 -10 0 10 20 30 40 50 60 I Core imbalance, % I I B l 1173 288 I I 8-17 Babcock & Wilcox
I Figure 8-13. Core Imbalance Vs Power Level, 290 to 345 EFPD (Cycle 4) 110 RESTRICTED (-24.3,102) (17,102) 100 90 ~ g 80 (-27,80) (18,80) i 70 = a 5 60 PERnllSSIBLE g OPERATlhG 50 (-50,50) REGION (35,50) g 40 8 = 30 20 THIS FIGURE VAll0 ONLY FOR OPERATION AFTER APSR WITH0RAWAL 10 TECHNICAL SPECIFICATION FIGURE 3.5.2-11 0 l -60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Core imbalance, % I I I I i17.5 289 I Babcock & Wilcox 8-18
9. STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated 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 Tests 9.1.1. Control Rod Trip Test Precritical 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 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 most accurate position indication is obtained from the zone reference switch at the 75%-inserted position, this position is used instead of the two-tht cs 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. 9.1.2. RC Flow Reactor coolant flow with four RC pumps running will be measured. Acceptance criteria require that the measured flow be within allowable limits. This test is planned for cycle 4 only to verify flow performance with the addition of the LBP rods. 9.1.3. RC Flow Coastdown The RC flow coastdown from the tripping of the most limiting RC pump combi-nation from four pumps running will be measured at hot zero power conditio~ns. The coastdown of RC flow versus time will then be compared to the, required RC flow versus time to determine whether acceptance criteria are met. This test i173 290 9-1 Babcock & Wilcox
I is planned for cycle 4 only to verify flow performance with the addition of LBP rods. 9.2. Zero Power Physics Tests 9.2.1. Critical Boron Concentration Criticality is obtained by deboration at a constant dilution rate. Once cri-ticality is achieved, equilibrium 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 boron concentration must be within 100 ppm boron of the predicted value. 9.2.2. Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all-rod-out configuration and at the hot zero power rod insertion limit. The average coolant temperature is varied by first decreasing then increasing tem-perature by 5*F. During the change in temperature, reactivity feedback is com-pensated by discrete change in rod motion, the change in reactivity is then calculated by the summation of reactivity (obtained from reactivity calculation on a strip chart recorder) associated with the temperature change. Acceptance criteria state that the measured value shall not differ from the predicted W value by more than 0.4 x 10-4 (Ak/k)/*F (predictec value obtained from Physics Test Manual curves). The moderator coef ficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is adJed to obtain moderator coefficient. This value must not be in excess of the acceptance criteria limit of +0.5 x 10-4 (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 method. The boron / rod swap method consists of establishing a deboration rate in the reactor coolant sys-tem and compensating for the reactivity changes of this deboration by_ inserting control rod groups 7, 6, and 5 incremental steps. The reactivity changes that 1173 291 I 9-2 Babcock & Wilcox
occur during these measurements are calcualted based on Reactimeter data, and differential 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 ther. summed to obtained integral rod group worths. The acceptance criteria for the control bank group worths are as fo; lows: 1. Individual bank 5, 6, 7 worth: predicted value - measured valuf x 100 < 15 measured value 2. Sum of groups 5, 6, and 7: predicted value - measured value measured value -< 10 x 100 9.2.4. Ejected Control Rod Reactivity Worth Af ter the CRA groups have been positioned near the rod insertion 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 then swapped in versus the controlling rod group and the worth determined by the change in the previously calibrated con-trolling rod group position. The boron swap and rod swap values are averaged 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) s 1.0% 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 'Jerification 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 any abnormalities before escalating to the 75% FP plateau. Rod index is established at a nominal full power rod configuration at which the oe }} 29 9-3 Babcock & VVilcox
I pover distribution was calculated. APSR position is established to provide 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. .ne minimum DNBR must be grearar than 1.30. g 3. The value obtained from the extrapc,letion of the minimum DNBR to the 5 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. 4. The value obtained from the extrapolation of the worst-case maximum LHR to the next power plateau overpower trip setpoint must be less g than the fuel melt limit or the extrapolated value of imbalance must g fall outside the RPS power / imbalance / flow trip envelope. 5. The quadrant power tilt shall not exceed the limits specified in the l Technical Specifications. W 6. The highest measured and predicted 1idial reaks shall be within the following limits: predicted value - measured value measured value -< 8. x 100 7. The highest measured and predicted total peaks shall be within the following limits: predicted value - measured value x 100 s 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 lata from these mod is for input to safety evaluations. Items 3 and 4 establish the criteria whereby escalation to the next power plateau may be accomplished without exceeding the safety limits specified by the safety analysis with re-gard tc JNBR 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. I 1173 293 I 9-4 Babcock & Wilcox
l. The highest measured and predicted radial peaks shall be within the following 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 x 100 s 7.5. measured value 9.3.2. Incore Vs Excore Detector Imbalance Correlation .ation at s40% FP Imbalances are set up in the core by control rod positioning. Imbalances are read simultaneously 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 criterion is not met, gain ampli-fiers on the excore detector signal processing equipment are adjusted to pro-vide the required gain. 9_. 3. 3. Temperature Reactivity Coefficient i t N100% FP The averizge 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 po-sition. Controlling rod group worth is measured by the fast insert / withdraw method. The temperature reactivity coefficient is calculated from the mea-sured changes in teactivity and temperature. Acceptance criteria state that the moderator temperature coefficient shall be negative. 9.3.4. Power Dcppler Reactivity Coeffic.ent at N100% 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 tud group worth is measured t.oing 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 ae ated above, and r the measured power change. 1173 294 9-5 Babcock & %Vilcox
The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual, n:r.eptance criteria state that the measured value shall be more negative than -0.55 x 10-4(ak/k)/% FP. 9.4. Procedure for Use When Acceptance Criteria Are Not Met. If the 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 iith Babcock & Wilcox technical personnel particip. -ing as re-quired. Further specific actions depend on evaluation results. These. actions can include repeating the tests with more detailed attention to test preree:1-sites, added tests to search for anomalies, or design personnel performing de-tailed analyses of potential safety problems because of parameter deviation. W Power is not escalated until evaluation shows that plant safety will not be compromised by such escalation. I I I I I I I I I I )373 295 9-6 Babcock & wiicox g
REFERENCES 1 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978. 2 Rancho Seco Nuclear Station, Unit 1 -- Final Safety Analysis Report, Sacramento Municipal Utility District (Docket No. 0-312). 3 Rancho Seco Nuclear Generating Station, Unit 1 -- Cycle 3 Reload Report, BAW-1499, Babcock & Wilcox, Lynchburg, Virginia, September 1978. 4 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, November 1976. 5 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1972. 6 " Classification and Selective Loading of Fuel for Rancho Seco," Letter Re-port, Babcock & Wilcox, Lynchburg, Virginia, December 1973 (proprietary). 7 Rancho Seco Unit 1 - Fuel Densification Report, BAW-1393, Babcock & Wilcox, Lynchburg, Virginia, June 1973. 8 D. B. Vassallo (NRC) to J. H. Taylor (B&W), Letter, " Calculation of the Effect of Fuel Rcd Bowing on the Critical Heat Flux for Pressurized Water Reactors," Dated June 12, 1978, Revised September 15, 1978. 9 J. H. Taylor (B&W) to D. B. Vassallo (NRC), Letter, " Determination of the Fuel Rod Bow DNB Penalty," December 13, 1978. 10 J. H. Taylor (B&W) to S. A. Varga (NRC), " Determination of CHF Penalty at 55% Closure," June 22, 1979. 11 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Babcock & Wilcox, Lynchburg, Virginia, June 1975. 1173 296 A-1 Babcock & \\Milcox
I 12 J. H. Taylor (B&W) to R. L. Baer (NRC), Letter, June 8,1977. (This letter confirms that the analysis in BAW-10103 is more conservative than the re-vised analysis submitted with the letter. The revised analysis incorpo-rates removal of the inlet nozzle U-baffle.) I I I I I I I I I I I I I I 1175 29 A-2 Babcock & Wilcox g .}}