ML19322B863

Cycle 5 Reload Rept, Revision 1

ML19322B863
Person / Time
Site:	Oconee
Issue date:	07/31/1978
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1493, NUDOCS 7912060688
Download: ML19322B863 (50)

Text

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July 1978 I 4MA con Y RA COPj I

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OCONEE UNIT 1. CYCLE 5

- Reload Report -

Revision 1 I.

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a 4 OCONEE L* NIT 1. CYCLE 5

- Reload Report -

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. Power Generation Group j Nuclear Power Generation Divistan P. O. Box 1260  !

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CONTEh"TS Page

1. INTRODUCTION . . . . . . . . . . . . .. . . . . . . . . . .... 1-1

2. OPERATING HISTORY . . . . . . . .. . . . . . . . . . . . .... 2-1

3. CENER\L DESCRIPTION . . . . . . . . .. . . . . . . . . . .... 3-1

4. JUEL SYSTEM DESIGN . . .. . . . .. .. . . .. . . . . . .... 4-1 4.1. Fuel Assembly Mechani al 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 . . . . . . . . . . . . . .. ... 4-2 4.3. Thermal Design . . . . . . . . .. . . . . . . . . . .... 4-3 Material Design . . . . . . . .. . . . . . . . . . ....

4.4.

4.5. Operating Experience . . . .... .. . .. . . . . .... 4-3

5. NUCLEAR DESIGN . . . . . . . . . . . .. . . . . . . . . . .... 5-1 5.1. Physics Characteristics . .. .. . . . . . . . . . .... 5-1 5.2. Analytical Input . . . . . . . . . . . . . . . . . . .... 5-2 5.3. Changes in Nuclear Design . . . . . . . . . . . . . . ... 5-2

6. 11tERMAL-HYDRAULIC DESIGN . . . . . . . . . ... . . . . . .... 6-1

7. ACCIDENT AND TRANSIENT ANALYSIS . . . . . . . . . . . . . .... 7-1 7.1. General Safety Analysis . .. . .. . . . . . . . . .... 7-1 7.2. Accident Evaluation . . . . . . . . ... . . . . . .... 7-1

8. PROPOSED HDDIFICATIONS TO TECllNICAL SPECIFICATIONS . .. . .... 8-1

9. STARTUP PROGRAM - PHYSICS TESTING . .. . . .. . . . . . .... 9-1 REFERENCES . . . . . . .. . . . .. ... . . . . .. . . .... A-1 l

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Revision 1 (7/31/78) t List of Tables Table Page 4-1. Fuel Design Parameters and Dimensions 4-2. Fuel Thersal Analysis Parameters ... .... .....

. ..... ..... 4-4 5-1. .....

Oconee 1. Cycle $ Physics Parameters .. . .... . ..... 4-5 5-2. 5-3 l1 Shutde n Margin Calcol. tion f or Ococce 1. Cycle 5 . . . . . . . 5-4 6-1. Thermal-Hydraulic Design Conditions ... . . .. .. .....

7-1. Comparison of Key Parameters for Accident Analysis . .....

6-3 l1 7-2. LOCA Limits. Oconee 1. Cycle 5 . . . . . . ..... ..... 7-3 7-3 List of Figures Figure 3- 1. Oconee 1. Cycle 5 - Full Core Loading Diagram . ....... .

3-2. 3- 2 3-3.

Enrichment and Burnup Distribution for Oconee 1, Cycle 5 . .. 3-3 5- 1..

Cont rol Rod Locations for Oconee 1. Cycle 5 . . . . . . . . . . 3-4 BOC. Cycle 5 Two-Dimensional Relative Power Distribution -- Full Power. Equilibrium Xenon, Normal Rod Positions ........ 5-5 8-1. Core Protection Safety Limits, Oconee Unit 1 .........

S-2. 8-2 l Protective System Maximum Allowable Setpoints. Oconee Unit 1 . 6- 3 8-3.

Und Position Limits for Four-Pump Operation. Oconee Unit 1 .. 8-4 -

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Rod Position Limits for Four-Pump Operation. Oconee Unit 1 .. B-5 i 8- 5. Rod Position Limits for Two- and Three-Pump Operation.

Oconee Unit 8-6. 1......................

Rod Position Limits for Two- and Three-Pump Operation.

. .. 8-6 Oconce Unit 1. ........................ S-7 8-7.

Power Imbalance Limits. Oconee Unit 1... ..... ..... 8-8 8-8.

8-9 Power Imbalance Limits. Oconee Unit 1.... .... ..... 8-9 APSR Position Limits. Oconee Unit 1.. ... ......... 8-10 8-10. APSR Position Limits. Oconee Unit 1.............. B-ll t

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Revision 1 (7/31/78)

1. INTRODUCTION This report justifies operation of the Oconee Nuclear Station, Unit 1, cycle 5 _ _ .

at a rated core power of 2568 MWt. The required analyses are included as out-lined in the USNRC document, " Guidance for proposed License Amendments Relat-inc. to Refueling," June 1975. This report uses the analytical techniques and design bases documented in several reports that have been submitted to and approved by the USSRC.

Cycle 5 reactor and fuel parameters related to power capability are summarized in this report and compared to those of cycle 4 All accidents analyzed in the Oconee FSAR have been reviewed for cycle 5 operation; a detailed conpar-ison of cycle 5 characteristics to the FSAR analyses showed that no new anal-vses were necessary since cycle 5 parameters are conservative.

The Technical Specifications have been reviewed and modified where required for cycle 5 operation. Based on the analyses perforned and taking into ac-count t.ie ECCS Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Oconee 1, cycle 5 can be safely operated at its licensed core power level of 2568 MWt.

Five fuel assemblies from batch 4 will be irradiated for a fourth cycle as part of a joint Duke Power /B&W/ Dept. of Energy program to demonstrate reliable fuel performance at extended burnups and to obtain post-irradiation data.

These assemblies will not adversely affect cycle 5 operation.

Because of performance anomalies observed at other B&W plants, orifice rod assemblies will not be used in Oconee 1, cycle 5. This change from normal t

practice has been accounted for in the analyses performed for cycle 5. In 1 addition, retainer assemblies will be installed on two f resh batch 7 fuel as-semblies containing regenerative neutron sources.

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2. OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Oconee 1 cycle 5 is the currently operating cycle 4. This cycle 5 design is based on a planned cycle 4 length of 235 EFPD rather than the design length of 292 EFPD.

Cycle 5 will operate in a feed-and-bleed mode for its entire design length of 330 EFPD. Initial cycle 4 operation was in a redded mode.

Hewever, a quad-rant power tilt was detected during cycle 4 power escalation I and the mode of operat ion was converted to feed-and-bleed to provide a larger margin f or cy-cle 4 operation.2 The shuffle pattern for cycle 5 wac. designed to minimize I

the effects of any power tilts present in cycle 4. No control rod interchange in planned during cycle 5.

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3. GENERAL DESCX1PTION fhe Oconee Unit i reactor core and fuel design basis are described in detail in section 3 of the Final Safety Analysis Report 3 for Oconee Nuclear Station, Unit 1. The cycle 5 core contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one in-core instrument guide tube. The fuel consists of dished-end, cylindrical pel-lets of uranium dioxide clad in cold-worked Zircaloy-4 The fuel assemblies in all batches have an average nominal fuel loading of 463.6 kg of uranius.

The undensified nominal active fuct lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for Oconee 1, cycle 5. The initial en-richment of the fresh batch 7 fuel is 3.02 vt % 235U. The remaining batches 4u, 5, and 6 were initially enriched to 3.20, 7.75, a.-d 2.795 wt % 2 3 U, re-spectively. All the batch 4A and all but five batch 43 assemblies will be discharged at the end of cycle 4. The five remaining batch 4B assemblies will be retained in cycle 5 and are redesignated as batch 4D. The batch 4D. 5, and 6 assemblies will be shuf fled to new locations at the beginning of cycle 5.

The f resh batch 7 assemblies will occupy the periphery of the core and eight interior locations. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 5.

Reactivity is controlled by 61 full-length Ag-In-Cd control rods and by solu-ble boron shin. In addition to the full-length contrcl rods, eight axial power shaping rods are provided for additional control of the axial power dis-scabution.

The cycle 5 locations of the 69 control rods and the group desig-nations are indicated in Figure 3-3. The core locations of the total pattern (69 control rods) for cycle 5 are identical to those of the reference cycle

{ indicated in the Oconee 1, cycle 4 reload report." The group designations, however, differ between cycle 5 and the reference cycle in order to minimize power peaking. Neither control rod interchange nor burnable poison rods are l necessary for cycle .%.

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Figure 3-2. Enrichment ar.d Burnup Distribution for oconee 1, Cycle 5 8 9 10 11 12 13 14 15 3.20 2.75 2.75 3.20 3.02 2.75 2.79 3.02 2d,479 20,488 16,053 31,135 0 15,903 5.889 0 3.02 2.75 2.79 2.75 2.79 2.75 3.02 K

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17,336 5,853 8,262 15,846 0 0 2.75 2.79 2.75 3.02 M

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Revision 1 (7/31/58)

4. FUEL SYSTEM DESIGN i

..!. Fuel Assemb!v Mechanical Design The types of f uel assemblies and pertinent fuel design paraseters ana disen-

t. ions for Oconee 1. cycle 5 are listed in Table 4-1. All fuel assemblies are identical in concept and are mechanically interchangeabic. All results, ref-crences, and identif ied conservatisms presented in section 4.1 of the Oconee

1. rvete 4 reioid report" are applicabic to the cycle 5 reload core.

Five batch 4D Ma* -B3 assemblies ere remaining in the core for their fourth cycle at irradiation and will experience burnups up to approximately 41.0^9 P4d/mtU as part of a joint Duke Power /86W/ Dept. of Energy program to demon-st rate extended burnup feasibility in 1.WRs. The Mark-B fuel assembly mechan-ical design will saintain its structural integrity with these burnups. Anal-v<es of post-irradiatica examinatien (PIE) data from two cycles of cperation j in the Oconce ! reactor shew that all parameters measured indicate that ex-tendsd operation in quite feasible. The parascters investigated include fuel rod and assembly growth, fuel swelling, and holddown spring force. The in- I tended peak burnups of batch 4D fuel are within the original mechanical peak design limits retorto) in the Oconee FSAR.3 Design parameters can be affected j

by burnup. effective full power time, or c.alendar residenc? time. Those param-  !

eters af fected most by the amount of irradiation are fuel rod and assembly growth and fuel swelling. Since burnup is within conservative design limits, growth will be acceptable. Section 4.2.3 discusses fuel swelling as it relates to cladding strein.

The holddown spring force is affected by residence time as well as burnup. Evaluation of the PIE data indicates that the holddown spring will meet performance requirements through the fourth cycle of irradia-tion.

Retair.er assemblies will be used on the two fresh batch 7 fuel assemblies that contain the regenerative neutron source (RNS) assemblies. The justification y

for the design and use of the retainers is described in reference 13, which is applicabic to the retainers used in Oconee 1, cycle 5.

4-1 Babcock s. Wilcox

.h Fuel Rod Design f.z?.l. Cladding Collapse Creep collapse analyses were performed for three-cycle assembly power hi.to-ries as well as for batch 4D's four-cycle assembly power histories. For cy-cic 5, the batch 5 fuel is more lim. ting than all other batches except for 4D because of its previous incore exposure time. The batea 5 and 4D assembly power histories were analyzed, and the most limiting assembly from each batch was dete rmined.

ine power histories for the most simiting assemblies were used t> calculate the fast neutron flux level for the energy rang 1 a*oove 1 MeV.

The collapse time for the most limiting assembly from each batch was conservatively deter-mined to be more than 30,000 effective full-power hours (EFPH), which is longer than the maximum projected batch 5 residence time of 21,456 EFPH (three cycles) and the maximum projected batcl. 4D residence time of 28,469 EFPH (four cycles).

The creep collapse analyses ware performed o* ased on the conditions set forth in references 4 and 5.

j. ._2 . 2 . Cladding Stress The Oconee 1 stress parameters are enveloped by a conservative fuel rod stress analysis. Since worst-case stress conditions are at BOL, the batch 4D feel is also bounded by the fuel rod stress analysis. For design evaluation, the pri-mary membrane stress must be less than two-thirds of the minimum specified un-irradiated yield strength, and all stresses ' primary and secondary) must be less than the minimum specified unirradiated yield strength. The margin is in excess of 30% in all cases. With respect to Oconee 1 fuel, the following con-servatisms were used in the analysis:

1. Low post-dens 111 cation internal pressure.

2. Low initial pellet density.

3. High systes pressure.

4. High thermal gradient across che cladding.

The stresses reported in reference 6 for core I f iel represent conservative values with respect to the cycle 5 core.

4.2.3. Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding circumferential plastic strain.

The pellet design is esta'alished for plastic cladding strain 4-2 Babcock a nVilcox -

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of less than 1% at maximum design local pellet burnup (55,000 mwd /mtU) and heat generation rate (20.15 kW/ft) values that are higher than the values the Oconee 1 fuel is expected to see, including batch 4D. The strain analysis is also based on the maximum Specification value for the f uel pellet dia=eter and da:sity and the lowest permitted Specification tolerance for the cladding ID.

4 . 1. Ther=al Des h All f uel assemolies in this core are thermally similar. The f resh batch 7 fuel inserted for cycle 5 operation introduces no significant dif ferences in fuel thernal performance relative to the other fuel remaining in the core.

The design minimun linear heat rate (LHR) capacity and the average fuel tenp-erature for each batch in cycle 5 are shown in Table 4-2. LHR capabilities are based on centerline fuel melt and were established using the TAFY-3 code 7 with fuel densification to 96.5% of theoretical density. The five batch 4D fuel assemblies have an EOC burnup of about 41,000 mwd /mtU. The EOL maximum pin pressure for these assemblies is well below the system pressure- of 2200 psia.

4.4. Material Design The batch 7 fuel assemblies are not new in concept, nor do they utilize dif-ferent component naterials. Therefore, the chemical compatibility of all pos-sible fuel-cladding-coolant-assembly interactions for the batch 7 fuel as-ser.blies are identical to those of the present fuel.

4.5. Operating Experience Babcock & Wilcox operating experience with the Mark-B, 15 by 15 fuel asseebly has verified the adequacy of its design. As of February 28, 1978, the exper-tence described below has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark-B fuel assembly. In addition, Three Mile Island Unit 2 achieved initial criticality on March 28, 1978, and is currently in the startup testing phase that precedes commercial operation.

Max assembly Cumulative Current "#""#'

net elect.

Reactor cvele Incore Disch. output , nWh Oconee 1 4 27,200 25,300 20,385,249 Oconee 2 3 26,700 26,800 15.248.595 Oconee 3 3 27,140 27,200 16,182,813 4-3 Babcock s.Wilcox I

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"""#' Cumulative Current net elect.

Reactor cycle Incore Disch. output, cia'h TM1-1 3 31,720 25,860 18,430,506 ANO-1 2 28,290 17,650 14,575,320 kar.cho Seco 2 22,300 17,170 10,297,637 Crystal River 3 1 10,430 -

4,936,412 Davis-Besse 1 1 2,490 --

1,009,741 Tabic 4-1. Fuel Design parameters and Dimensions Thrice- Twice- Once-burned burned burned Fresh FAs, FAs, FAs, FAs, Batch 4D Batch 5 Batch 6 Batch 7 FA type Mark-B3 Mark-E4 Mark-B4 Mark-B4 No. ..f FAs 5 60 56 56 Fuel rod OD, in. 0.430 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 0.377 Flex. spacers, type Spring Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 Undensif active fuel 142.0 142.6 142.25 142.25 length (ncs), in.

Fuel pellet initial >94.5 93.5 94.0 94.0 density (nom), % TD Fuel pellet OD (mean 0.3685 0.3700 0.3695 specif), in. 0.3695 Initial fuel enrich., 3.20 2.75 2.79 3.02 wt 1 235g BOC burnup (avg), 30.604 17,011 6,539 Wd/mtU 0 Cladding collapse >30,000 >30,000 >30,000 time. EFPH >?O,000 Estimated residence 28,469 21,456 22,440 26,496 ,I time (max) EFPH 4 I

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$ Table 4-2. Fuel Thermal Analysis Parameters Batch 4D(# I#2 6(a)

S 7 No. of assemblies 5 60 56 56 Nominal pellet density, % TD 95.5 93.5 ,4.0 94.0

{ Pellet diameter, in. 0.3685 0.3700 0.3695 3.3c95 Stack height, in. 141.O(b) 142.6 142.25 f42.25 Dens i f ied Fue l Pa rame t ers #

Pellet diameter, in. 0.3640 0.3645 0.3646 0.3646 Fuel stack height, in. 140.30 140.46 140.47 140.47 Nominal LliR at 2568 MWt, kW/ft 5.80 5.80 5.80 5.80 Avg fuel temp at ocminal LliR. F 1320 -1320 1320 1320 LliR t o Q f ue l cci t , kW/ft 20.15 20.15 20.15 20.15 I

(* Data from reference 4.

(b) Conservative calculational parameter.

4 (*}Densification to 96.5% TD assumed.

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of design cycle 5 with thore of reference cycle 4. The values for both cycles were generated using PDQ07.

The average cycle burnup will be higher in cycle 5 than in the design cycle 4 because of the longer cycle 5 length. Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 5 at full power with equilibrium xenon and normal rod positions.

The critical boron concentrations for cycle 5 are comparable to those of the design cycle 4.

The control rod worths for hot full power differ betweea cy-cles due to changes in group designations us well as changes in radial flux distributions and isotopics. The ejected rod worths in Table 5-1 are the max-trum calculated values within the allowable rod insertion limits. Calculated elected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rcJ position limits presented in section 8.

The maximum stuck rod worth for cycle 5 is greater than that for the design cycle 4 at BOC and approximately the sane at EOC.

All safety criteria associated with these worths are met. The adequacy of the shutdown margin with cycle 5 atuck rod worths is demonstrated in Table 5-2.

The following conservatisms were applied for the shutdowr1 calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth.

3. Flux redistribution penalty.

Flux redistribut. ion was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown mar-gin is presented in the Oconee 1, cycle 4 reload report."

The cycle 5 power deficits from hot zero power to hot full power differ from i those for tne design cycle 4 because of the longer cycle 5 design length.

The differential boron worths and total menon worths for cycle 5 are greater 5-1 Babcock & Wilcox 1

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l than or equa. to those for the design cycle 4 because of f uel depletion and the associated buildup of fission products. Effective delayed neutron frac- ,

l tions for both cycles show a decrease with burnup.

5.2. Analytical Input The cycle 5 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the reference cycle.

_5. 3. Changes in Nuclear Design There were no relevant changes in core design between the reference and reload cyc l es .

The same calculational methods and design information were used to obtain the important nuclear design parameters. The only significant opera-tional procedure changes f rom the reference cycle are the operation in a feed-and-bleed mode and removal of the ORAs. The reference cycle began operation in the rodded mode but was subsequently modified for operation in the feed-and-bleed mode.

Therefore, since nearly the entire reference cycle 4 was op-erated in the feed-and-bleed mode, this is not actually a new mode of opera-tion.

Removing the ORAs does not significantly affect the nuclear character-istics of the core. g i

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9 Tabic 5-1. Oconee 1. Cycle 5 Physics ? ara eters(#'

t rycle 4 N cycle SI '}

I Cycle length. ETPD 292 110 Cycle burnup. .'fW4/stU 9.136 10.327 g

Averete core burnup. EOC. 'L'J/atU 19.03.

I 19.C27 Initial core loading, atu 82.1 82.1 Critical baron BOC (no Xe , ppe FZP. group 8 37.5l wdtd 1415 1458 h2P. groups 7 and 8 inserted 1315 1324 EFP. group 8 inserted 1145 1276 Critical boron. EOC (eq Xe), ppe NZP. group 8 57.5% wd 373 H7P. group 8 37.5% wd 34)

BS 44 Control od worths. HFP. BOC. 2 Ak/k Group 6 1.07 l

Croup 7 1.21 Croup 8 17.51 vd 0.93 1.45 0.50 0.43 Control rod worths. HFP EOC I ak/k Cresp 7 1.16 1.53 i Croup 8 37.5 wd 0.47 0.48 Na elected rod worth. HIP, 2 Ak/k

• SOC (N-12) 0.68 0.57 Luc (N-12) .

0.61 0.70 M as stuck rod worth. HIP.1 Ak/k 3x (N-12) 1.74 2.17 FJC (N-12) 2.02 2.01 FaerBM deficit. HZP to HFP.1 Ak/k LOC 1.49 1.31 2.07 2.12 l

e Doppler coef f.10'S(6k/k 'F)

BOC. 100% power. no Xe -1.45 EOC.1001 power. eq Xe -1.45

-1.55 -1.62 Nderator coef f. HFP. 10-6(ak/k *F)

Sc< (0 Xe. crit ppe. gp 8 ins) -1.00 -0.45 EOC (rq Xe. 17 ppe. gp 8 ina) -2.55 -2.%

Boron worth. HFP. pps/1 Ak/k BOC (1150 pra) 10S EX (17 ppa) 109 101 97 Xenon worth. HFP. Ak/k Sec (4 EFFD) 2.60 EOC (equilibriu:s) 2.62 2.61 2.73 Eff delayed neutron fraction. HFP 50C EOC 0.00593 0.00598 ,

0.00530 0.00521

(* Cycle 5 data are for the conditions stated in this report .

The cycle 4 core conditions are identified in reference 4.

Based on 292 EFFD at 2568 MWt. cycle 3.

(* Cycle 5 data are based on a "plar.ned" cycle 4 length of 235 EFPD; the cycle 4 " design" lifetime is 292 EFPD.

HIP denotes hot sero power (532F Tavg). RFP denotes hot I full power (579F Tayg).

• Ejected rod worth for groupe 5 through 8 inserted.

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Table 5-2. Shutdown Margis calculation for Oconee 1, Cvele 5 BOC, % Ak/k EOC, % Ak/k Avsilable rod w:rth Total rod vorth, H2P 8.91 d.79

'.' orth reduction due to burnup -0.36 -0.42 of poison raterial Maxicium stuck rod H2P -2.17 -2.01 Net worth 6.38 6.36 Less 10 uncertainty -0.64 -0.64 Total available worth 5.74 5.72 Required rod worth Power deficit, liFP to HZP 1.31 2.12 Max allowable inserted rod 0.40 0.60 worth Flux redistribution 0.59 1.20 Total required worth 2.30 3.92 Shutdown margin (total available 3.44 worth minus total required worth) 1.80 hte: Required shutdown margin is 1.00: J.k /k .

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Figure 5-1. 30C (4 EFPD), Cycle 5 'No-Dimensional Relative Power Distribution - Full Power, Equilibriu:2 Xenon, Normal Rod Positicos (Group 8 Inserted) 8 9 10 11 12 13 I I. 15 i

H 0.83 0.93 0.96 0.90 1.37 1.03 1.09 0.87 1

I ,

K 1.35 1.07 1.21 0.98 1.09 0.93 0.83 i

L X 1.05 1.25 1.03 0.95 1.15 0.67 I

M 1.09 1.23 0.89 0.91 N 1.21 0.94 0.61 i

f 0 0.70 i,

P ,

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Inserted Rod Group No.

x.xx Relative Power Density 1 il l

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6. T11ERMAl.-IlYDRAt:1.IC DESIGN I, The thermal-hydraulic design evaluation supporting cycle 5 operation itilized the acthods and models described in ref erences 3, 4, and 6 except for the core l

bypass flow and the radial = local peaking. The fresh bat:h 7 fuel is hydraul- l ically and geometrically similar to batch 6 fuel. ne cycle 4 and 5 naximum  !

design conditions and significant parameters are shown in Table 6-1. The min- 1 imura DSBR shown at the design overpower for cycle 5 is based on 106.5% of RC design flow and on the Rark-B4 fuel assembly; it includes the effects of in-core fuel densification, core bypass flow, and peaking.

Fuel .nssemblics that do not contain control rods or neutron sources usually I contain orifice rod assemblies (ORAs) to minimize core bypass flow. There are 6

a total of 108 possible locatiotas for ORAs. In cycle 4, 70 ORAs and two re-r,enerative neutron sources were installed, leaving 36 vacant fuel assemolies.

The r_aximuni core bypass for cycle 4 analysis was 8.34% based on an assumed l removal of 44 ORAs. All ORAs will be removed in cycle 5 (two RNSs with retain-ers will remain incore) leaving 106 vacant fuel assemblies and a maximum core bypass flow of 10.4%.

To offset the thermal-hydraulic effects of the increased bypass flow, the ref-crence design radial = local peaking factor (Fah) has been reduced free 1.78 to 1.71. This is supported by the cycle 5 nuclear design with a predicted man-imum F.ht of 1.527. The reactor core safety limits take into account the in-creased bypass flow and the decreased F.h* t The potential effect of f uel rod bow on DNBR was considered by incorporating suitable margins into DNS-limitcd core safety limits and RPS setpoints. The maximum rod bow was calculated from the equation EC - 0.065 + 0.001449. E o

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6-1 Babcock a.Wilcox ll l

Revision 1 (7/31/78) where LC = rod bow magnitude, mils.

Co= initial gap (138 mils),

BU = maximum assembly burnup mwd /atU.

The fuel cycle design calculations show that the maximum radial-local peak during cycle 5 is always located in the batch 7 fuel asse=bly with the maximiza burnup.

peak.

This maximum peak (1.527) is 12% below the 1,71 reference design l1 Since this fuel assembly is limiting for DNER analysis, the rod bow penalty associated with batch 7 is applied to cycle 5 operation. This method for calculating the maximum core rod bow penalty has been reviewed and e approv d for acceptability by the USNRC.8 The Oconce 1, cycle 5 calculated rod bow pen-alty is 8.0% based on the maximtsu burnup in batch 7 - 13,667 mwd /mtU ,

credit . No the is clained for the difference between calculated cycle 5 peaking and reference design peaking used for the analysis.

An 11.2% rod bow penalty is conservatively applied to all analyses that define plant operating limits

.ind to design transients.

The flux / flow trip setpoint was determined by analyzing an asstmed two-pump I coastdown initiated from four-pump operation with an indicated power level of 102: of 2 568 .'f4t.

The flux / flow trip setpoint of 1.055, previously estab- 1 11o' ed for cycles 3 and 4, is retained for cycle 5 and yields a ministan DNBR ,

during the two-ptamp coastdown, of 1.30 plus a suitable margin to of fsetthe  ;

asstased 11.2% fuel rod bow penalty.

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Revision 1 (7/31/78,)

Table 6-1. Thermal-Hydraulic Design Conditions Cycle 4" Cycle 5 Power level, .Wt 2563 2568 Systets pressure, psia 2200 2200 Reactor coolant flow, % design flow 106.5 106.5 Vessel inlet coolant teep, 555.6 535.6 k 100% power. F Vessel outlet coolant te=p, 602.4 602.4 1002 power, F i Ref design radial-local power 1. TS 1. 71 l1 peaking factor Ref design axial flu shape 1.5 cos 1.5 cos Active fuel lenr,th, in. (a) (a)

Average heat flux, 100 power, 176 176 l

i 10* Btu /h-ft2 CilF correlation B.W-2 SX4-2 Ilot channel factors Enthalpy rise 1.011 1.011 Ilea t flux 1. 0! *. 1.014 Flow area 0.98

{ 0.98 Minimu:n DNBR with densif'n penalty 1.91 1.98 il t

I"}See Table 4-2.

(b) Based on densified length of 140.3 inches.

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{ 7. ACCIDENT AND TRANSIENT ANALYSIS 7_ ._1. General Safety Analysis Fach FSAR3 accident analysis has been examined with respect to changes in cy-cle 5 parameters to determine the ef fect of the cycle 5 reload and to ensure that thermal performance during hypothetical transi .s is not degraded.

The effects of fuel densification on the FSAR accident results have been . val-uated and are reported in refs;rence 6. Since batch 7 reload fuel assenSlies contain fuel rods whose theoretical density is higher than those considered in the reference 6 report, the conclusions in that reference are e till valid.

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 thernal pa-raneters, thermal-hydraulic parameters, and kinetics parameters, including the f react ivity feedback coef ficients and control rod worths.

Core t h e rs.n l p rope r t i c.9 used in the FSAR accident analysis were design operat-ing values based on calculational values plus uncertainties. Fuel thermal analysis values for each batch in cycle 5 are cotspared in Table 4-2. The cy-l cle 5 thernal-hydraulic maximum design conditions are compared to the previous cycle 4 valu'es" in Table 6-1. These parameters are comon to all the acci-dents considered in this report. A comparison of the key kinetics parameters f rom the ISAR and cycle 5 is provided in Table 7-1.

A generic LOCA analysis for the B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model. This study is re-ported in BAW-10103, Rev. 1.10 The analysis in BAW-10103 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 L11R and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the f 7-1 Babcock 3. Wilcox

atalysis and the LOCA limits reported in RAW-!OlO3 provide conservative results for the operation of Uconee 1 cycle 5 fuel.

4 Table 7-2 shows the bounding vtlues for allowable LOCA peak LHRs for Oconce 1, cycle 5 fuel.

It is concluded f rom the examination of cycle 5 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that this core re-load will not adversely af fect the Oconee 1 plant's ability to operate safely during cycle 5. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cyclc 5 is considered to be bounded by previously accepted analyses. The initial conditiens for the transients in cycle 5 are bounded by the FSAR 3 , the fuel densification report 6, and/or subsequent cycle analyses.

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j l e 7-1. Comparison of Key Parameters for Accident Analysis FSAR and Predicted densification cycic 5 Parameter report value value Doppler coeff. Sk/k/*F SOC -1.17 a 10-5 -1.41 d 10-5 Eoc -1.33 a 10-5 -1.62 a 10-5 Moderator coeff, ak/k/*F SOC +J.5 a 10 * -0.45 a 10-'

EOC - 3. 0 a 10 * -2.64 = 10-*

All-rod group worth. I!ZP 2 Ek/k 10 8.91 Initial boron cone'n. HFP. ppm 1400 1276 Boron reactivity worth at 70F, pps/1: ak/k 75 76 i Max ejected rod worth. HFP %

Ak/k 0.65 0.25 Dropped rod worth (HFP). I ak/k 0.46 0.20 Tatte 7-2. 1.0CA I.inits. Oconee 1. Cycle 5 Elevation. LHR limits, ft r

• 'J/ft 2 15.5 4 16.6 6 18.0 8 17.0

' 10 16.0 1

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8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 5 operation. Changes were the results of the following:

1. The Technical specification limits based on DNBR and LHR criteria include appropriate allowances for projected fuel red bow penalties, i.e., poten-I tial reduction in DNBR and increase in pcwar peaks. A statistical co=hi-nation of the nuclear uncertainty factor, engineering hot channel factor, t

and rod bow peaking penalty was used in evaluating LHR criteria, as ap-proved in reference 11.

2. Per reference 12, the power spike penalty due to fuel densifIcation was not used in setting the DNBR- and ECCS-dependent Technical Specification limits.

3. The allowabic quadrant tilt limit for cycle 5 is 5.0%.

Based on the Technical Specifications derived from the aaalyses presented in this report, the Final /.cceptance Criteria ECCS limite will not be exceeded,

) nor will the thermal design criteria be violated. Figures 8-1 through S-10 illustrate revisions to previous Technical Specification limits.

8-1 Babcock & Wilcox l

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Fir,ure 8-1. Core Protection Safety Liciits. Oconee Unit i Thermal Power Level, %

- 120

(-28,112) (32,112) l ACCEPTABLE 100

(_go,gg) 4 PUMP (50.98)

OPERATION (32.85.3)

(-29,85.3) - -

80 i

ACCEPTABLE 384 PUNP (50,71.3)

(-40,71.3) OPERATION

, . 60 (32,58.2) 3

(-2a,58.2)

ACCEPTABLE

(-40,44.2) . . 40 2.384 PUNP OPERATION 20 .

' ' ' i i , ,

60 -40 -20 0 20 40 60 80 Reactor Power imbalance, %

CURVE RC FLOW (GPM) This is proposed new Tech-1 374,880 nical Specification Figure 2.1-2A.

2 280,035 3 183,690 l

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J Figure B-2. Frotective Systes .t ximus A*1cvable  ;

Setpoints. Oconee Unit I l l

i Thermal Power, 1.

(-18.105.5) ,

(20.105.5) h M2 = -0. 75 HI = 0.568 ~8

. .100 4 PUMP OPERATION l {3 0. 96)

(-40.93)

I e

, (-18.78.8)_

- 80 (20,78.8) z A

3

$ t-40.66.3) 384 PUMP  %

l (30.71.3)

E O

a.

" OPERATION s' w l g

- - 60 l w h (-18.51.69) (20.51.,69) y l

8 g (-40.39.19) l

- 40 b (30.44.19) b 8

l g l l 4 I l

o $Im I o - - 20 0 l o

,, a0 ,, l n a ,l ,, ,

, = < a o ,i # e ,, #. ,

-60 -40 -20 0 20 40 60 Reactor Power imbalance, %

This is proposed new Technical Specification Figure 2.3-2A.

8-3 Babcock & Wilcox i

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Figure 8-3. Rod Position Limits for Four-Pump Operation.

Oconec Unit 1 (0 to 100 3 10 EFF3)

(125.802) (274.1.sc2) 100 .

O CPitAT10s NOT ALLW'O (274.1.92) ---

PCwit 80 -

I#

(258.2.80) aE57titfE0 SMUT 00mm MARGlu LIMITS R[GION to -

(215.7.60) f 3 (70.50) 7:

.. so -

6 I Pitul5588LE meATim 20 -(ie.is) f afGION 4

(0.12.5

,0,0)

( , , ,

8 Ot , , , , , . .

9 50 100 150 200 250 300 1, Rod inces % WD 0 25 50 75

. . . . i O. O O.

25 50 75 lec Group 5 Gros, 7 0 25 50 7s 800 Group 6 This is r,roposed new Technical Specification Figure 3.5.2-1A1.

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i Rod Position Limits for Four-Puna Operation, Figure 8-4.

Oconce Unit 1 (After 100 : 10 EFPD)

(2ss.102) (273.1.102)

Ico . O r

l CPilatIDs (275.8.92) *____

m37 ALL0wtD ,

! 248.2.60) LitEL 80 -

Cut 08F

= 924 FP CO -

j (214.7.60)

% (179.50)

PitMIS$15L' Orrga!,nG

$NU1095144tG iu REGI0s

{ LIMIT - _

20 -

(103.85)

{0.6.1) 0_ i (0 01 , , , , , . , , ,

0 50 IOC 150 200 250 300 Rod IMes. % Withdrawn 0 25 50 75 100 0 25 50 75 100 t 1 a a i t t I f Group 5 Gro.3 7 0 25 50 75 100 t t t i Gro.e 6 This is proposed new Technical Specificatic.. Figure 3.5.2-1A2.

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Fir,ure 3-5. Pod Positien 1.imits for Two- and Three-Pump Operation. Oconce Unit 1 (0 to 100 : 10 EFPD)

(125.802) (198.102 (248.2.lc2) t00 .

CP(Rail 0e DOT ALLout3 gd

$0 *

//%

f

, (284.7.76)

(120.68) gi

,$MUTDows hafGen gf 7 LlWli 2

(10,50) 40 -

PitWIS$l5LE

• OPER4finG REGION (18.85) 20 h,12.5 O g

• i t i f , , ,

0 50 100 ISO 200 250 300 25 led laces. 5 Witherawn

? 5,o 75 0

'.m 2p 5p 7,5 ipo Grove 5 Group 7 0 25 50 75 000 t i a i Group 6 This is proposed new Technical Specification Figure 3.5.2-2A1.

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I Figure 8-6.

Rod Position Limits for Two- and Three-Pu=p

operation, Oconce ll nit 1 (After ICO

10 EFFD)

,,g33,,,

4 100 - (2st.ac2) (2ss.2.102) not ALL0wtD

?

f EE5talCTED FOR

~

3 P"# P g21s.7.76) 2

( to -

j (179.50)

% P(M81SSl8tE

$nUTD0*u MARGIN LIMIT CPTIAT14G

{ , =0 ggg f t a

~

(103.151 (0.6. 0 )

0 i , ,

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i ,

0 50 i

100 150 200 0 25 50 75 too laces. I mitNta-a 250 300 g e 800 0 a t i

i 25 50 75 800 1 1 I Gro e 5 i Grows 7 0 25 50 t i 75 800 t

A Crose 6 This is proposed new Technical Specification Figure 3.5.2-2A2.

s.;

Babcock & Wilcox

Figure s- 7. Power Itbalance 1.1 sits. Oconec t; nit 1 (0 to 100 ! 10 EFFD)

Power, % of 2568

• t RESTRICTED REGION

- - 110

(-26. 9.102'i r .,, (15.9.102)

(-28.2,92) '

'(15.9,92) 90 .

(-29.2.80) < ,

-- 80 '

'(15.9.80) 70 PERMI SSI Bl.E -- 60 OPERATING REG l0N

- - 50

- - 40

- - 30

- - 20 10

• ' ' ' I r , , ,

-50 -40 -30 -20 -10 0 +10 +20 +30 '40 +50 Axial Power imbalance. %

This is proposed new Technical Sp'ecification Figure 3.5.2-3A1.

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rigure S-8.

Power Imbalance Limits, Oconee L* nit I (Af ter 1001 10 EFPD)

Power. % of 2568 Mit

- 110 RESTRICTED REGION

(-28.0.102) r

..gg '(16.4.102)

'Io.l.92)

- i

,,g ii(16.4.92)

(-29.2.80)i ,

- 60

(16.4.80)

PERMISSIBLE -

70 OPERATIKG REGION -

- 60

- 50

- 40 30

-. 20

.. 10

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, 7 This is proposed rew Technical Specification Figure -3.5.2-3A2.

g9 Babcock s. Wilcox

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Figure 8-9.

APSR Position Limits. Oconee Unit 1 (From 0 to 100 ! 10 EFPD)

(16.3.102) (45.0.102) 100 . t g RESTRICTED (14.l.92) '

(45.0.92) 80 (0,0.80) .9.80)

(100.60) k 60 -

S N

i g PERMISSIBLE e OPERATING i J REGION l m

)

20 -

This is proposed new Technical Specification Figure 3.5.2-4A1.

0 ' s e 0 20 40 60 80 100 Bank 8 Position, 7. Withdrawn l i

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Figure 8-10.

APSR Position Li::its. Ocence Unit 1 (After 100 : 10 EFFD)

(9.3.102) (45.0.102)

, , RESTRICTED

' 100 -

REGION

, (7.6,92) , (45.0.92) b 80 (0.0.80) (57.9.80) 100.60) ft

. 60 -

j k PERMISSIBLE

% 'PERATING

. REGION 2 40 -

b 20 -

This is proposed new Technical Specification Figure 3.5.2-4A2.

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0 20 40 60 80 100 Bank 8 Position. % Withdrawn i l

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Revision 1 (7/31/78) l

9. START 1'P PROGRAM - P11YSICS 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 the necessary data for continued safe opera-tion.

Precritical Tests

1. Control rod trip test.

Zero Power Physics Tests

1. Critical boron concentration.

2. Temperature reactivity coefficient.

a. All rods out, group E in.

b. Groups 5 through 8 inserted, groups 1 through 4 out.

3. Control rod group reactivity worth.

4 Ejected control rod reactivity worth.

5. Reactor coolant flow with four pumps running at hot zero power, steady ~

state conditions. l 1

6. Reactor coolant flow coastdown f rom tripping of the limiting pump combi-nation from four pumps running at zero power.

Power Tests

1. Core power distribution verification at approximately 40, 75, and 100%

f ull power with normal control rod group configuration.

2.

Incore versus out-of-core detector imbalance correlation verification at less than f ull power.

3. Power Doppler reactivity coef ficient at approximately 100% full power.

4. Temperature reactivity coefficient at approximately 100% full power.

9-1 Babcock a,Wilcox

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l REFERENCES I

Oconee 1. Cycle 4 Quadrant Flux Tilt. BAW-1477, Babcock & Wilcox, Lynchburg, V atinia, January 1978.

2 A. C. Thies (Duke Power Co.) to Edson G. Case (USNRC) Letter, October 26, 1977 Docket No. 50-269.

3 Oconee Nuclear Station, Units 1, 2, and 3 - Final Safety Analysis Reports, Docket Nos. 50-269, 50-270, and 50-287 Duke Power Company.

Oconee Unit 1, Cycle 4 Reload Report, BAW-1447, Babcock & Wilcox, Lynchburg, Virginia, March 1977.

5 Program to Determine In-Reactor Performance of B&V Fuels - Cladding Creep Collapse, BAV-10084 Rev. 1. Babcock & Wilcox, Lynchburg, Virginia, Decem-A ber 1976.

6 Oconce 1 Fuel Densification Report, BAV-l'588, Rev. 1 Babcock & Wilcox, Lynchburg, Virginia, July 1973.

7 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pres.sure Analysis, BAV-10044 Babcock & Wilcox, Lynchburg, Virginia, May 1972.

9 Safety Evaluation by the Of fice of Nuclear Reactor Regulation Supporting A:xndment 14 to Facility Operatisig License No. DPR-54 Rancho Seco Nuclear t,eneration Station, Sacramento Municipal Utility Distric t. Docket No.

50-312.

9 Duke Power Company to E. G. Case (Acting Director. Office of Nuclear Reac-tor Raeulation), Letter, " Revision of Oconee Nuclear Station Tech. Spec. to Modify Pump Monitor *. rip Setpoint," September 14, 1977.

10 ECCS Analy:;is of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 1, Babcock

& Wilcox, Lynchburg, Virginia, September 1975.

A-1 Babcock & WilCOX M P e -. --. - , .r a f5

Revision 1 (7/31/78)

II S. A. Varga (USNRC) to J. H. Taylor (B&W), Letter, "Conaments on B&W's Sub-mittal on Combination of Feaking Factors," May 13, 1977.

12 S. A. Varga (USNRC) to J. H. Taylor (B&W), letter, " Update of BAW-10055 -

ruel Densification Report," December 5,1977.

I3 BPRA Retainer Design Report, EJ-1496, Babcock & Wilcox, Lynchburg, Vir- g ginia, May 1978.

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