ML19312C136

Cycle 5 Reload Rept, Revision 1

ML19312C136
Person / Time
Site:	Oconee
Issue date:	09/30/1978
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1493, NUDOCS 7912060690
Download: ML19312C136 (50)

Text

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,BAW-1493, Rev 2 September 1978 t

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I OCONEE UNIT 1 CYCLE 5 l i

- Reload Report -

Revision 2 i

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e BABCOCK & WILCO 7, Power Generation Group Nuclear Power Generatior Division P. O. Box 126')

Lynchburg, Virginia 24505 Babcock a Wilcox I

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i CONTFhTS Page

1. IhTRODUCTION . .... ....... . . . . . . . . . . . . . . . 1-1

2. OPERATING HISTORY

... .... . . . . . . . . . . . . . . . . . 2-1

3. CENERAL DESCRIPTION .......... .. .

. . . . . . . . . 3-1

4. FUEL SYSTEM DESIGN . ..........

4.1.

. . . . . . . . . . . . . 4-1 Fuel Assembly Mechanica ', Design 4.2. Fiel Rod Design . . . . . . . . . . . . . . 4-1 4.2.1. Cladding Collapse

.... .. . . . . . . . . . . .. . . . . 4-2 4.7.2. Cladding Stress . . .. . . . . . . . . . . . . . 4-2 4.2.1. Cladding Str.in . . . . . . . . . . . . . . . . . . 4-2 4.3. Thermal Design . . ......... . .. . .. .. . . . . . . . . . 4-2 4.4 Material Design . . . . . . . . . . . . . 4-3 4.5. ..............

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

. .. . . . . . .. . . . . . . . . 4-3

5. hTCLEAR DESIGN .

5.1.

Physics Oiaracteristics 5-1 5.2. Analytical Input . . .. . . . . . .* . . . . . . . . . 5-1 5.3. ... .. .. . .. .. . . . . . . . . .

Changes in Nuclear Design 5-2

.. . . .. .. . . . . . . . . . 5-2 6.

IIIERMAL-HYDAAULIC DESIGN .

7.

.................... 6-1 ACCIDEhT AND TRANSIENT ANALYSIS 7.1. .. . ..... . . . . . . . . . 7-1 General Safety Analysis 7.2. Accident Evaluation .... . . .. . . . . . . . . . . 7-1

...... . . .. . . . . . . . . . . 7-1 8.

PROPOSED HODIF1 CATIONS TO TECHNICAL SPECIFICATIONS . .. . . . . . 8-1 9.

STARTUP PROGRAM - PHYSICS TESTING

... . . . . . . . . . . . . . 9-1 REFERENCES . ....

....................... A-1

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Revision 1 (7/31/78)

List of Tables Table Page 4-1. Fuel Design Parameters and Dimensions ........

4-2. Fuel Thermal Analysis Parameters ..... 4-4 5-1. Oconee 1. Cycle 5 Pnysics Parameters .........

..... 4-5 5-2. .............

5-3 6-1. Shutdown Margin Calculation for Oconee 1, Cycle 5 . . . . . . . 5-4 il 7-1.

Thermal-Ilydraulic Design Conditions .

............. 6-3 l1 7-2. Comparison of Key Parameters for Accident Analysis . ..... 7-3 LOCA Limits. Oconee 1 Cycle 5 . .

.............. 7-3 List of Figures Figure

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3-2.

Oconce 1. Cycle 5 - Full Core Loading Diagram . ........ 3-2 1-3. Enrichment and Burnup Distribution for Oconee 1. Cycle 5 .. . 3-3 3- 1. Control Rod Locations for Oconee 1 Cycle 5 . . . . . . . . . . 3-4 BOC, Cycle 5 TVo-Dimensional Relative Power Distribution - Full Power. Equilibrium Xeno" Normal Rod Positions .... ....

5-1.

Ccre Protection Safety Limits, Oconee Unit 1 ......... 5-5 6-2. 8-2 3-3. Protective System Maximum Allowable Setpoints, Oconee Unit 1 . 8-3 3-4. Rod Position Limits for Four-Pump Operation, Oconee Unit 1 . . 8-4 3-5. Rod Position Limits for Four-Pump Operation, Oc > nee Unit 1 . . 8-5 Rod Position Oconee Limits for Two- and Three-Pung Operation, Unit 1...

3-6. .................. .... 8-6 Rod Position Limits for Two- and Three-Pump Operation, Oconee Unit 8-7. 1.....................

Power imbalance Limits. Oconee Unit

.... 8-7 8-6. 1............. 8-8 8-9. Power Imbalance Limits. Oconee Unit 1 . . . . . . . . . . . . .

APSR Position Limits. Oconee Unit 1.. ............ B-9 3-10. 8-10 APSR Position Limits. Oconee Unit 1......... ..... B-ll

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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 ara included as out-lined in the USNRC document. " Guidance for Proposed License Amendnents Relat-ing 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 USNRC.

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 (SAR have been reviewed for cycle 5 operation; a detailed compar-ison of cycle 5 characteristics to the FSAR analyses showed that no new anal-yses were necessary since cycle 5 pa.ameters are conservative.

The Tcchnical Specifications have been reviewed and modified where required for cycle 5 operation. Based on the analyses performed and taking into ac-count the ECCS Final Acceptance Criteria and postulated fuel densification 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 vill 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.

Thase assemblies will not adversely affect cycle 5 operation.

Because of performance anomalies observed at other B&W plants, orifice rod azaemblies will not be used in Oconee 1, cycle 5. This change from normal practice has been accounted for in the analyses performed for cycle 5. In 1 rddition, retainer assemblies will be installed on two fresh batch 7 fuel as-gemblies containing regenerative neutron sources. l Th2 criginal cycle 5 design was based on a planned cycle 4 length of 235 EFPD.

Dua to an extension of cycle 4 to 250 EFFD, the cycle 5 analyses have been re- 2 vitwed and modified where required. )

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2. OPEP,ATING HISTORY The ref erence cycle for the nuclear and ther:nal-hydraulic analyses of Oconee

1. cycle 5 is the currently operating cycle 4. This cycle 5 design is based on o planned cycle 4 length of 250 EFPD rather than the design length of 292 l2 EFPD.

Cycle 5 will operate in a feed-and-bleed mode for its entire design length of 330 EFPD. Initial cycle 4 operati was in a rodded mode. However, a quad- l2 rant power tilt was detected dur'r cycle 4 power escalation l , and the r Je of operation was converted to feed tr4d-bleed to provide a larger margin for cy-cle 4 operation.2 The shuffle pattern for cycle 5 was designed to minimize the effects of any power tilts present in cycle 4. No control rod interchange is planned during cycle 5.

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3. GENERAL DESCRIPTION 1

) The 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 l

by 15 array containing 208 fuel rods, 16 control rod Fulde tubes, and one in- l core instrument guide tube. The fuel consists of dished-end, cylindrical pel- l lets of uranitra dioxide clad in cold-worked Zircaloy-4. The fuel assechlies l in all batches have an average nominal fuel loading of 46%.6 kg of uranium.  !

, The undensified nominal active fuel lengths, theoretical densities, fuel and f uel rod dimensions, and other rwlated 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 wt % 235U. The remaining batches 4D. 5, and 6 were initially enriched to 3.20, 2.75, and 2.795 wt % 235 U, re-spectively. All the batch 4A and all but five batch 4B assemblies will be discharged at the end of cycle 4. n e five remainfag batch 4B assemblies will be retained in cycle 5 and are redesignated as batch 4D. The batch 4D, 5, and 6 assemblies will be shuffled to new locations at the beginning of cycle 5.

The fresh 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 control rods, eight axial l power shaping rods are provided for additional control of the axial power dis- 1

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tribution. The cycle 5 locations of the 69 control rods and the groep desig- '

nat hns are indicated in Figure 3-3. n e core locations of the total pattern (6t< 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. l however, dif fer between cycle 5 and the reference cycle in order to minimize power peaking. Neither control rod interchange nor burnable poison rods are i I

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Revision 2 (9/1/78) necessary for cycle 5.

Tne nominal system pressure is 2200 psia, and the core l average densified noctinal heat rate is 5.80 kW/ft at the rated core power of 2 2568 MWt.

3-2 Babcock & Wilcox

d. Revision 1 (7/31/78) f Figure 3-1. Oconee 1, Cycle 5 - Full Core Leading Diagra:D i

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7 7 7 7 7 E7 24 29 1

, , , , 6 , , , 7 M2 C6 FF L2 MS L14 F9 C10 M16 7 6 S 6 6 S 5 S S 6 7 311 et) L1 m3 DS Dil 513 L1S 03 SS D 6 7 6 6 6 S 7 5 6 6 6 6 7 F3 A10 09 El s2 OS* m14 KIS E3 A6 F13 7 5 6 5 6 6 40 6 6 S 6 S 7 C6 C12 A9 E!) D6 MS D10 07 A7 C4 C10 7 7 S 6 6 S S S S S 6 6 5 7 7 GS 810 E4 312 F4 18 T12 34 E12 86 C11 7 5 6 S 6 S 7 5 7 5 6 5 6 5 7 k1S M11 M13* 312 39 E14e 37 ue E3a El N1 7 6 5 7 40 S S 4D $ 40

$ 7 5 6 7 KS F10 Me F12 L4 C4 L12 Fe M12 F6 E!!

7 $ 6 S 6 S 7

• 5 7 5 6 5 6 7 E6 C12 39 C9 at DS 510 C3 B7 06 K10 7 7 5 6 6 S S S S S 6 6 S 7 7 q L3 110 C.3 C1 D2 C11* Die CIS C7 36 L13 7 5 6 , 6 6 .. 6 6 , 6 , 7 F11 C13 F1 D3 55 511 D13 FIS C3 FS

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• 1matism of thrtee-bormed batch 4D assemblies.

P imat am et tegererattwo snatena source assemblies end rotateer assembtles, [

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Revision 2 (9/1/78)

Figure 3-2. Enrich:nent and 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 H

28,923 20,985 16,578 31,581 0 16,428 6,283 l 0

3.02 2.75 2.79 2.75 2.79 2.75 3.02 K '

O 14,746 5.477 I 19,695 9,030 16,831 0 2.75 2.79 2.79 2.75 3.02 3.02 17,841 6.247 8,787 16,404 0 0 M

2.75 2.79 2.75 3.02 17,846 5,346 18,853 0 2.79 2.79 3.02 N

6.227 7,549 0 0

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Figure 3-3. Contrc.1 Rod Locatiens for oconee 1 Cycle 5 A

a 3 5 3 I

C 1 1 7 7 1 D 6 8 4 8 6 1 1 5 2 2 5 1 F 3 0 7 6 7 8 3 G 7 2 4 4 2 7 It u- 5 4 6 3 6 4 5 ~T E 7 2 4 4 2 1 L -

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P g $ 2 9 $ 1

% 6 8 4 P 6 0 1 7 7 1 P 3 5 3 a

l l 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 I i 1

a +- Group 1 ember Croup me, et mode Pection 1 8 Safety 2 8 Safety 3 9 Safety

& S Safety 5 8 Centrol 6 8 Castrol 3 12 Cactrol s e arsa.

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4 TUEL SYSTEM DESIGN

_4.1. Fuel Assechtv Mechanical Design I The types of fuel assemblies and pertinent fuel design parameters and dimen-sions for ocoree 1, cycle 5 are listed in Table 4-1. All fuel asse=blies ara identical in concept and are cechanically interchangeable. All results, raf-crences, and identified conservatisms presented in section 4.1 of the Oconee 1, cycle 4 reload report4 are applicable to the cycle 3 re* ad core Five hatch 4D Mark-B3 assemblies are remaining in the core for their fourth cycle of irradiation and will experienca burnups up to approximately 41,000 mwd /mtU as part of a joint Duke Power /B&W/ Dept. of Energy program to demon-strate extended bernup feas*bility in LWRs. The Mark-B fuel assembly nechan-ical design will maintain its structural integrity with these burnups. Anal-yses of post-irradiation examination (PIE) data from two cycles of operation in tle Oconee 1 reactor show that all parameters measured ina. are ' hat ex-tended operation is quite ~ sible. The parameters investigated inc4uda fuel rod and assembly growth, fuel swelling, and holddown spring force. The in-tended peak burnups of batch 4D ruel are within the original mechanical peak design limits reported in the Oconee FSAR.3 Design parameters can be affected by burnup, ef fective full power time, or calendar residence time. Those param-

• ters af fected most by the amount of irradiation are fuel rod and anaembly growth and fuel swelling. Since burnup is w! thin conservative design limits, growth will be acceptable. Sectlen 4.2.3 discusses fuel swelling as it relates to cladding strain. The holddown spring force is affected by residence time as well as burnup. Evalustion of the PIE data indicates that the holddown spring will meet performance requirements through the fourth cycle of irradia-tion.

Retainer assemblies will be used on the two frech batch 7 fuel assemalies that ontain the regenerative neutron source (RNS) assemblies. The justification y

for the design and use of the retainera is described in reference 13, which 1.3 applicable to the retainers used in Oconee 1, cycle 5.

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l Revision 2 (9/1/78)

A.2. Fuel Rod Design

_4.2.1. Cladding Collapse _

Creep collapse analyses were performed for three-cycle assembly power histo-ries as well as for batch 4D's four-cycle assembly pcwer histories. For cy-cle 5, the batch 5 fuel is more limiting than all other batches c;teept for 4D because of its previous incore exposure time. The batch 5 and 4D assembly 1

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power histories were ar alyzed, and the most limiting assembly fron each batch (

was determined. {

The power histories for the ex>st limiting assemblies were used to calculate  !

I the fast neutron flux level for the energy range above 1 MeV. I The collapse time for the most limiting assembly f rom each batch was conservatively deter- {

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mined to be more than 30,000 effective full-power hours (EFPH), which is longer than the maximum projected batch 5 residence time of 21,336 EfPH (three 2

cycles) and the maximum projected batch 4D residence time of 28,349 EFPH (four cycles).

The creep collapse analyses were performed based on the conditions set forth in references 4 and 5.

4.2.2. Cladding Stress The Oconee 1 stress paraneters are enveloped by a conservati se fuel rod stress analysis.

Since worst-case stress conditions are at BOL, tt.e batch 4D fuel 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 1 j

less than the minir.um specified unirradiated yield strength. The margin is in l excess of 30% in all cases. L'ith respect to Oconee 1 fuel, the following con-servatisms were used in the analysis:

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1. Low post-densification internal pressure,

2. j Low initial pellet density.

3. $

High system pressure.

4 High thermal gradient across the cladding.

The stresses reported in reference 6 for core 1 fuel represent conservative values with respect to the cycle 5 core.

4.2.3. Cladding Strain

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The f uel design criteria specify a limit of 1.0% on cladding circumferential l plastic strain.

The pellet design is established for plastic cladding strain -

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of less than 12 at maximum design local pellet burnup (55,000 mwd /mtu) and heat generation rate (20.15 kW/f t) 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 maximus Specification value for the fuel pellet diameter and density and the lowest permitted Specification tolerance for the cladding ID.

4.3. Thernal Design All f uel asse=blies in this core are thermally similar. The fresh batch 7 f uel inserted for cycle 5 operation introduces no significant dif ferences in fuel thermal performance relative to the other fuel remaining in the core.

The design nininua linear heat rate (LHR) capacity and the average fuel temp-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 IAFY-3 code 7 with fuel densification to 96.52 of theoretical density. The five batch 4D fuel assenblies have an EOC burnup of about 41,000 mwd /mtU. The EOL maximum pin pressure for these assemblies is well below the systen pressure of 2200 psia.

• . 4. M'terial Design The batch 7 fuel assemblies are not new in concept, nor do they utilize dif-f e ren t component materials. Therefore, the chemical compatibility of all pos-sible fuel-cladding-coolant-assembly interactions for the batch 7 fuel as-

c. b!ies are identical to those of the present fuel.

4.5. Operating Erperience

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& Wilcox operating experience with the Mark-B,15 by 15 fuel assembly has verified the adequacy of its design. As af February 28, 1978, the exper-lence 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 cycle Incore Disch. output, mWh 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 'N"

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Max assembiv Cumulative

"#""#* net elect.

Current 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 Rancho 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 Table 4-1. Fuel Design Parameters and Dic.ensions Thrice- Twice- Once-burned burned burned Fresh FAs, FAs, FAs, FAs, Batch 4D Batch 5 Batch 6 Batch 7 FA type Mark-B3 Mark-B4 Mark-B4 Mark-B4 No. of 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 i

Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 Undensif active fuel 142.0 142.6 142.25 142.25 length (nom), 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 0.3695 i specif), in.

Initial fuel enrich. , 3.20 2.75 2.79 3.02 wt : 235g l BOC burnup (avg), 31,049 17,524 6,965 0 l2 mwd /mtU Cladding collapse >30,000 >30,000 >30,000 >30,000 time, EFPH Estimated residence 28,349 21,336 22,320 26,256 l2 l time (max), EFFH I

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Table 4-2. Fuel Thermal Analvsis Parameters Batch 4DI ") I S *} 6I *) 7 No. of assemblies 5 60 56 56 Nominal pellet density, % TD 95.5 93.5 94.0 94.0 Pellet diameter, in. 0.3685 0.3700 0.3695 0.3695 Stack height, in.

141.05 ) 142.6 142.25 142.25 Densified Fuel Parameteg (*}

Pellet diameter, in. O.3640 0.3645 0.3646 0.3646 Fuel stack height, in. 140.30 140.46 140.47 140.47 Nominal UIR at 2568 MWt, kW/f t 5.80 5.80 5.80 5.80 Avg fuel temp at nominal LHR, F 1320 1320 1320 1320 LH R ' o Q f uc t mel t , kW/ft 20.15 20.15 20.15 20.15 I" Data from reference 4.

I ) Conservative calculational parameter.

(* Densification to 96.5% TD assumed.

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5. NUCLEAR DESIGN 5.1. Physics Characte ristics Table 5-1 conpares the core physics parameters of design cycle 5 with those of reference cycle 4 The values for both cycles were generated using PDQO7.

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 relat ive 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 between cy-cles due to changes in group designations as well as changes in radial flux distributions and isotopics. The ejected rod worths in Table 5-1 are the max-tuum calculated values within the allowable rod insertion limits. Calculated ejected rod worths and their adherence to criteria are considered at all times in l life and at all power levels in the development of the rod position limits presented in section 8 1 The maximum stuck rod worth for cycle 5 is greater than that it r the design cycle 4 at BOC and approx eately the same at EOC.

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

The following conservatisms were applied for the shutdown calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on net rod wurth. '

3. Flux redistribution penalty.

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

Th2 .

cycle 5 power deficits f rom hot zero pnver to hot full power dif fer from those for the design cycle 4 because of the longer cycle 5 design length.

Th2 dif ferential boron worths and total xenon worths for cycle 5 are greater 5-1 Babcock s.Wilcox .

Revision 2 (9/1/78) than those for the design cycle 4 because of fuel depletion and the associated buildup of fission products. l2

' Ef fective delayed neutron fractions for both cy-cles 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 i

here were no relevant changes in core design between the reference and reload cycles.

We same calculational methods and design information were used to obtain the important nuclear design parameters, ne 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.

l Removing the ORAs does not significantly af fect the nuclear character-1stics of the core. -y J

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Table 5-1. Oconee 1. Cycle 5 Physics Parameters Cycle 4 IbI Cycle SI *I Cycle length. EFFD 292 320 Cycle burnup. NWd/stu 9.136 10.014 Average core burnup. EOC. Wd/stU 19.034 19.055 Initial core loading. stu 82.1 82.1 Critical boron. 80C (.ao Xe). pre RIF. group 8 37.5% void) 1415 1426 EIF. groups 7 and 8 inserted 1335 1293 RFF. group 8 inserted 1165 1242 Critical boron. EOC (eq Xe), pp. 2 HZF. grwp 8 37.5% wd 373 3)$

ETF. group 8 37.51 vd 88 43 Control rod wortha. EFF. BOC, I ak/k Croup 6 1.07 1.19 Croup 7 0.93 1.44 Croup 8 37.5% wd 0.50 0.42 Control rod worths. MFF EOC. I ak/k Croup 7 1.16 1.52 Croup 8 37.51 ud 0.47 0.48 Mas ejected rod worth R2F.1 Ak/k *I bOC (N-12) 0.68 0.57 EOC (N-12) 0.61 0.70 Mas stuck rod worth RZF. I Ak/k BOC (N-12) 1.74 2.17 EDC (N-12) 2.02 2.01 Power deficit. HIP to HFF. 1 Ak/k 80C 1.49 1.31 EDC 2.07 2.11 Doppler coef f.10-5(Ak/k *F)

BOC. 1001 power no T.4 -1.45 -1.45 EOC. 1003 power. eq ... -1.55 -1.61 Moderator coef f, urF.10**(ak.*k *F)

BOC (0 Xa. crit ppe, gp 8 ina) -1.00 -0.44 EDC (eg Xe.17 ppe, e 8 ina) -2.55 -2.63 Borce worth. ETF, ppe/1 Ak/k 80C (1150 ppm) 109 los EOC (17 ppm) 2 101 97 Kenne worth. HFF. I 4k/k 80C (4 Ef7D) 2.60 2.62 EOC (equilibrium) 2.61 2.74 Eff delayed neutron fraction. HFF 80C 0.00593 0.00595 EDC 0.00530 0.00520 I*I Cycle 5 dacc are for tt.: conditions Jtated in this report.

The cycle 4 core conditions are identified in reference 4.

Based on 292 EFFD at 2568 wt. cycle 3.

(* Cyc?e 5 data are based on a " planned" cycle 4 length of 250 CFFD; c.he cycle 4 "destga" lifetime is 292 EFFD. l2 HIP denotes hot aero power (532F Tavg). RTP denotes bot full power (579F T. ,g).

Ejected rod worth for groups 5 through 8 inserted.

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Table 5-2. Shutdown Margin Calculation for Oconee 1. Cycle 5 BOC. % Ak/k EOC. : Ak/k Available rod worth Total rod worth, HZP 8.85 8.75 2

Worth reduction due to burnup -0.36 -0.41 of poison. material Maximum stuck rod, HZP -2.17 -2.01 Nec worth 6.32 6.34 Less 10% uncertainty -0.63 -0.63 Total available worth 5.69 5.71 Required rod worth Power deficit. HFP to HZP 1.31 2.11

.m x allowable inserted rod 0.38 0.68 worth Flux redistributEon 0.59 1.19 Total required worth 2.28 3.98 Shutdown margin (total available 3.41 1.73 worth minus total required worth)

Note: Reovired shutdown margin is 1.00% Ak/k.

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i Revisien 2 (9/1/78)

{ Figure 5-1.

BOC (4 EFPD). Cycle 5 Two-Dimensional Relative Power Distributir,n - Full Power, Equilibrius Xenon k Normal Rod Positioras (Group 8 Inserted) .

8 9 10 11 12 13 14 15 H 0.82 0.93 0.95 0.89 1.37 1.02 1.10 0.89 K 7 1.35 1.06 1.20 0.98 1.09 0.94 0.85 L

1.03 1.23 1.02 0.94 1.17 0.69 2

M 1.08 1.22 0.89

. 0.93 N

1.21 0.94 0.62 0

0.71 P

l R

L Inserted Rod Group No.

x.xx Relative Power Density

)

5-5 Babcock & Wilcox

1 l

Revision 1 (7/31/78)

6. THERMAL-HiDRAULIC DESIGN Ec thermal-hydraulic design evaluation supporting cycle 5 operation utilized the methods and models described in references 3, 4, and 6 except for the core bypast. flow and the radial = local poaking. We fresh batch 7 fuel is hydraul-ically and geometrically similar to batch 6 fuel, ne cycle 4 and 5 maxisten design conditions and significant parameters are shown in Table 6-1.

De min- 1 imum DNBR shown at the design overpower for cycle 5 is based on 106.5% of RC design flow and on the Mark-B4 fuel assembly; it includes the effects of in-core f uel densification, core bypass flow, and peaking.

Fuel assen$ lies that do not contain control rods or neutron sources usually contain orifice rod assemblies (ORAs) to minimize core bypass flow. There are a total of 108 possible locations for ORAs. In cycle 4, 70 ORAs and two re-generative neutron sources were installed, leaving 36 vacant fuel sssemblies.

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

To offset the thermal-hydraulic effects of the increased bypass flow, the ref-crence design radial x local peaking factor (Fah) has been reduced from 1.78 to 1.71.

His is supported by the cycle 5 nuclear design with a predicted max-imum F h of 1.527. The reactor core safety limits take into account the in-creased bypass flow and the decreased FLh*

The potential effect of fuel rod bow on DNBR was considered by incorporating w itable margins into DNB-limited core safety limits and RPS setpoints. The maximum rod bow was calculated from the equation E

Co= 0.065 + 0.001449efti 6-1 Babcock a. Wilcox

l Revision 1 (7/31/78) l l

I where t.C = rod i,ow magnitude, mils, Co = initial gap (138 mils),

i BU = saximum assembly burnup, Wd/mtU.

The fuel cycle design esiculations show that the maximum radial-local peak during cycle 5 is always located in the batch 7 fuel assembly with the navimtra burnup. This maximum per.k (1.527) is 12% below the 1.71 reference design l1 peak. Since this fuel assembly is limiting for DNBR 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 approved l for acceptability by the USNRC.8 The Oconee 1, cycle 5 calculated rod bow pen-

)

alty is 8.0% based on the maximum burnup in batch 7 - 13,667 Wd/mtU. No I credit is claimed for the dif ference between calculated cycle 5 peaking and the reference design peaking used for the analysis. An 11.2% rod bow penalty is conservatively applied to all analyses that define plant operating limits and to design transients.

The flux / flow trip setpoint was det.emined by analyzing an asstaned two-pump coastdown initiated from four-pump operation with an indicated power level of 102% of 2568 W t. The flux / flow trip setpoint of 1.055, previously estab-lished for cycles 3 and 4, is retained far cycle 5 and yields a minimum DNBR, during the two-ptunp coastdown, of 1.30 plus a suitable margin to offset the I

asstaned 11.2% fuel rod bow penalty.

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Table 6-1. Thermal-Hydraulic Design Conditions Cycle 4 Cycle 5 Power level. MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design flow 106.5 106.5 Vessel inlet coolant temp, 555.6 555.6 100% power, F Vessel outlet coolant temp, 602.4 602.4 100% power, F Ref design radial-local power 1.78 1.71 l1 l peaking factor Ref design axial flux shape 1.5 cos 1.5 cos Active f=nel length, in. (a) (a)

Average heat flux, 100% power, 176 176 103 Btu /h-ft2 CHF correlation BAW-2 BA*a'- 2 Hot channel factors Enthalpy rise 1.011 1.011 licat flux 1.014 1.014 Flow area 0.98 0.98 Minimum DNBR with densif'n penalty 1.91 1.98 il

<" See Table 4-2.

I Based on densified length of 140.3 inches.

L 6-3 Babcock & Wilcox l,

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7. ACCIDENT AND TRANSIENT ANALYS15 /

7.1. General Safety Analvsis Each FSAR3 accident analysis has been examined with respect to changes in cy-cle 5 parameters to determine the effect of the cycle 5 reload and to ensure that thermal perft mance during hypothetical transients is not degraded.

The ef f ects of fuel densification on the FSAR accident results have been eval-uated and are reported in reference 6. Since batch 7 reload fuel assemblies contain fuel rods whose theoretical density is higher than tnose considered in the reference 6 report, the conclusions in that reference are still valid.

7.2. Accident Eval ua tion The key parameters that have the greatest ef fect on determining the outcome of a transient can typically be classified in three major areas: core thermal pa-rameters, thermal-hydraulic parameters, and kinetics para =eters, including the reactivity feedback coef ficients and control rod worths.

Core ther=al properties used in the FSAR accident analysis were design operat-ing values based on calculetional values plus uncertainties. Fuel thermal analysis values for each batch in cycle 5 are compared in Table 4-2. The cy-cle 5 thermal-hydraulic maximum design conditions are cocpared to the previous cycle 4 values" in Table 6-1. These parameters are common to all the acci- I dents considered in this report. A comparison of the key kinetics parameters f rom the FSAR and cycle 5 is provided in Table 7-1.

I A generic LOCA analysis for the B&W 177-FA, lowered-loop NSS has been performed f

using the Final Acceptance Criteria ECCS Evaluation Model. This study is re-  !

ported in B7.* J -10103, Rev. 1.10 The analysis in BAW-10103 is generic since the limitin values of key parameters for all plants in this category were used. i Furthermore, the combination of average fuel teraperature as a function of LER and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the 7-1 Babcock & VVilcox l

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1 analysis and the LA>CA '.imits reported in BAW-10103 provide conservative results for the operation of oconee 1 cycle 5 fuel. 1 l

Table 7-2 shows the bounding values for allowable LOCA peak LHRs for Oconee 1, cycle 5 fuel.

It is concluded from the examination of cycle 5 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that this core re-Ioad will not adversely affect 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 cycle 5 is considereo to be bounded by previously accepted analyses. Ti.e initial condition.- for the 3

transients in cycle 5 are bounded by the FSAR , the fuel densification report 6, and/or subsequent cycle analyses.

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Table 7-1. Comparison of Kev Parameters for Accident Analysis  !

FSAR and Predicted densification cycle 5 Paraneter report value value A'ppler coef f, ak/k/"?

AX -1.17 x 10-5 _t,43 m 10-5 t

EOC -1.33 x 10-5 -1.61 x 10-5 i %derator coef f, ak/k/*F Eoc +0.5 x 10 -0.4S x 10-4 2 E0' - 3. 0 x 10 " -2.63 x 10-"

All-rod group worth, HZP %

ak/k 10 8.85

'nitial boron cone'n, HFP, pp:n 11.00 1242 Scron reactivity worth at 70F,

pa/11 ak/k 75 76 Lx ejected red worth, HFP, %

2i/k 0.65 0.25 Oropped rod wrth (HFP), %

2kik O.46 0.20 Table 7-2. LOCA Limits. Oconee 1, Cycle 5 Elevation, liiR limits, ft kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 7-3 Babcock 3.Wilcox h

Revision 2 (9/1/78) l D M D 'l ' D ji hrue A

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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 rod how penalties, i.e., poten-tial reduction in DNBR and increase in power peaks. A ctatistical combi-nation of the nuclear uncertainty factor, engineering hot channel factor, 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 j not used in setting the DNBR- and ECCS-dependent Technical Specification limits.

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

Sased on the Technical Specifications derived f rom the analyses presented in this report, the Final Acceptance Criteri:a ECCS limits will not be exceeded, nor will the thermal design criteria be violated. Figures 8-1 through 8-10 illustrate revisions to previous Technical Specification limits. Figure S-2 has been revised due to the extension of cycle 4 to 250 EFPD.

8-1 Babcock s. Wilcox .

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_oom-o. . d h. 2 Figure 8-1. Core Protection Safety Linits. 0 onee L* nit 1 Thermal Power Level. %

- 120

(-28.112) (32.112) 1

. ACCEPTABLE I

4 PtNP (50.98)

(-40.98)

OPERATION l (32.85.3) 2

(-23.85.3) 80 I p

3&4 PUMP (50.71.3)

(-40.71.3) OPERAT10M

,_ 60 (32.58.2) 3

(-28.58.2)

ACCEPTABLE

(-40.44.2) 2.314 PLMP (50.44.2)

_ _ 40 OPERATION 20 8 I 1 I t I e 60 -40 -20 0 20 40 60 80 Reactor ft, er imbalance. %

CURVE RC FLOW (GPM) This is proposed new Tech-1 374.880 3.I cal Specification r1gure 2.1-2a.

2 280.035 3 183.690 I

8-2 Babcock t. Wilcox

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[ Fevisien 2 O*.'1/ 76) bob @. ) . .J 3 Figure B-2. ' Protective Syste:s Lximum Allowable Setpoints, Oconee L' nit 1 Thermal Power. %

(-18.105.5) (20.105.5)

Mi = 0.659 . .100 M2 = -0. 75 ,

4RW (30.98)

( -40.91 ) OPERATim l i I E ( -18. 78. 8 ) _

80 (20.78.8) 5 5

n. 3&t4 PJ4P  %

I I (I.71.3) 5 C

< ,i r OPERATI0pl

• (-W.64. 3) g 5 8

--m 8

(-18.51.69) (3.51.69) 'j l

u

=

l (E.%.19) w 5 (-40.37.19) ~~W l 1 l $ i i i = '

4 I l

o UIz Io -

- 20 o p E22 la fl o

. E w ** 7 g ti n w

_g  : l o.

c, I

a uns I 11 :  : l.  : I f

-60 -40 -3 0 m 40 60 Reactor Power imnalance. "

This is proposed new Tecnnical Specification Figure 2.3-2A 8-3 Babcock a. Wilcox

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Figure 6-3. Rod Position Limits for Four-Pump dl o .fl erat on.

. $ J Od " ._,

Oconee Unit 1 (0 to 100 : 10 EFPD)

(125.1C2) (274.1.102) 100 -

O CPERAft0s NOT ALLChrtD (274.1.92) __

PCwfR '

~

, , CUT 0F 8 92% FP

$WJTDOWN MAR 6in LIMIT 7 REGI0s CO -

1- (214.7.60) 5 (70.50)

N

[, 40 -

,k PERMIS$tBLE CPE RATims 20 -(33,3$) i, REGlos l

(

(0.12.5

,0,0)

(

et , , , , , , , ,

u 50 10C 150 200 250 300 0 Rod inden. 1 WD 25 50 75 0

. , . . f o, o 25 50 75 800 Gra e 5 sem, 7 0 25 50 75 300 Group 6 This is propo$ed new Technical Specification Figure 3.5.2-1A1.

s-4 Babcock s, Wilcox

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{U Fid;ure 8-4. Rod Position !.imits for Four-Pu=p Operation, Oconee Unit 1 (After 100 2 10 EFPD)

(241.s02) (274.1.102) i0o _

n CPERaiton (274.1.92) *----

40f ALL0ut0 6 245.2.80)

• LEL e

80 -

CUT 0FF

= 321 FP EO  %

j (284.7.E0)

% (179.50)

I PEeMI55ISLC 0FCA'stG U ad'wiDWu MARGlu REGIOe d LIMIT - __

20 -

(103.15)

(0.E.8)

,10. 0) , , , , , , ,

O_ , , ,

0 50 10C 150 200 250 300 0 25 50 75 100 O 25 50 75 400 m , t e a a e I t t Group 5 Sroup 7 0 25 50 75 300 t t t i Group 6 This is proposed new Technical Specification Figure 3.5.2-1A2.

8-5 Babcock a.Wilcox

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• D D ~TV l 60 ob . 2_ J .a i Figure 8-5. Rod Position Limits for ho- and Three-Pu:np Operation, Oconee Unit 1 (0 to 100 : 10 EFPD)

(125.102) (198.102 (248.2.I'02)

~ '

OPleaTIGu WOT ALLOWED gd 80 -

/d%

S (214.7.76)

(120.Cl) gt

,SHUTD0wn putGiu gd g

j Lirlf G (70.50) j PERMIS$1BLE 40 -

OPERATING

• REGION e

f* (18.15)

,O

). tz.5 '

e

} ' ' ' ' ' ' ' ' '

06 '

200 250 300 0 50 100 150 Rod inder. % Withdrawn 29 y p i,w 9 25 5p 7,5 ipo Group 5 Grows 7 0 25 50 75 800 t f . 1 B Group 6 This is proposed new Technical Specification Figure 3.5.2-2A1.

s 8-6 Babcock a, ?.'i:cox .

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. . n Figure 8-6. Rod Position Limits for Two- and Three-Pu p Operation, Oconee L' nit 1 (After 100 : 10 EFP3)

LP ER Allon I .m Ito - NOT ALLenE3 RESTRICTED FOR 3 PUMP

• ~

g214.7.76) 2 S

"j to -

,$ (177.50) PEsisissitLE OPERATipS

% SMUT 00mm etARGin LlulT GESION

", s0 -

6 i

e

1. ~

(103.154 i

(c.s.1) 01 , .3 , , , . , , , . .

0 50 EDO  !!O 200 250 300 foo Inces. ', mitecrawn C *S 50 75 100 0 25 50 75 tC0 t t t t t I 1 i I I Group 5 Grevo 7 0 25 50' 75 800 t t Gro.3 6 This is proposed new Technical Specification Figure 3.5.2-2A2.

, 8-7 Babcock & Wilcox

Figure 8-7. Power Imbalance Limits. Oconee Unit 1 (0 to 100 10 EFPD)

Power, % of 2568 W t

' ~ ' '

RESTRICTED REGION

(-26.9,102% .,

(15.9,102)

(-28.2,92) < '

90

(-29.2,80) <' -- 80 '

' (15.9,80)

- - 70 PERMISSIBLE - - 60 OPERATING R EGION

-- 50

- . 40

- - 30 l

- - 20 l in e t i i f f I I I

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

This is proposed new Technical Specification Figure 3.5.2-3A1.

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Figure 8-8. rever Iebalance Limito. Oconee 'Jn it 1 (After 100 : 10 EFPD)

Power. 5 of 2568 w t

--110 RESTRICTED REGION

(-25.0.102) r

_,, (16.4.102)

(-28.5.92) 8 -

-- 90

(-29.2.80)< i -- 80 d '

(16.4.80)

PERMISSIBLE -- 70 OPERATING REG!ON -- 60

-- 50

-' 40

-- 30

- - 20

_ . 10

-50 -40 -30 -20 -10 0 10 20 30 40 50 l

Axial Power labalance, %  ;

This is proposed new Technical Specification Figure 3.5.2-3A2.

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. S-9 Babcock 8. Wilcox i

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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 - ' '

RESTRICTED (14.l92) <

(45.0,92) 80 (57.9,80)

(0.0.80)

(100.60)

{ 60 -

E N

g PERMISSIBLE s, OPERATING

~

J REGION Y

c2 20 -

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

0 ' . , ,

0 20 40 60 80 100 Bank 8 Position, 7. Withdrawn 4

.I 8-10 Babcock s. Wilcox l

Figure 8-10.

APSR Positic.n Linits. Oconee Unit 1 (After 100 : 1sEFPD) 4 (9.3.102) (45.0.102)

RESTRICTED

' ~

l REGION (7.6.92) , (45.0.92) 80 (0.0.80) (57.9.80) l l

. 100.60) f 60 -

h PERMISSIBLE o OPERATING REGION .

Y, 40 -

20 -

l This is :-oposed new Technical Specification Figure 3.5.2-4A2.

O i i e 0 20 40 60 80 100 -

Bank 8 Position, 7. WitNrawn l

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8-11 Babcock 8. Wilcox l 1

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Revision 1 (7/31/78) l

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 the necessary data for continued safe opera-t tion.

P_recritical Tests

1. Control rod trip test.

Zer$ Power Physics Tests .

1. Critical boron concentration.

Tenperature reactivity coefficient.

2.

a. All rods out, group 8 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 pucips running at hot zero power, steady- l state conditions.

g

6. 1 Reactor coolant flow coastdown from tripping of the limiting pump co=bi-nation f rom four pumps running at zero power.

Power Tests

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

full power with normal control rod group configuration.

2.

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

3.

Power Doppler reactivity coefficient at approximately 100% full power.

4.

Temperature reactivity coefficient at approximately 100% full power.

91 Babcock s. Wilcox

RE FERENCES I

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

2 A. C. Thics (Duke Power Co.) to Edson G. Case (USNRC). Letter, October 29, 1977. Docket No. 50-269.

3 Oconee Nuclear Station, Units 1, 2, and 3 - Final Safety Analysis Report..

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-Reaccor Performance of B&V Fuele - Cladding Creep Collapse. BAW-10084, Rev. 1. Babcock & Wilcox, Lynchburg, Virginia, Dece -

ber 1976.

6 Oconee i Fuel Densification Report, BAW-1388, Rev. 1 Babcock & Wilcox.

Lynchburg, Virginia, July 1973.

7 C. D. Morgan and 11. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure .

Analysis. BAW-10044 Babcock & Wilcox, Lynchburg, Virgiaia, May 1972.

8 Safety Evaluatien by the Of fice of Nuclear Reactor Regulation Supporting

! Amendment 14 to Facility Operating License No. DFR-54, Rancho Seco Nuclear l Generation Station, Sacrasento Municipal Utility District, Docket No.

l 50-312.

9 Duke Power Company to E. G. Case (Acting Director, Of fice of Nuclear Reac-tor Regulation), Letter, " Revision of Oconee Nuclear Station Tech. Spac. to Modify Pump Monitor Trip Setpoint," September 14, 1977.

10 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BA*=*-10103, Rev. 1, Babcock )

& Wilcox Lynchburg, Virginia, September 1975. j i

, A-1 Babcock & \Milcox

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Revision 1 (7/31/78) 11 S. A. Varga (CSNRC) to J. H. Taylor (B&W), Letter, "Co:xaents on B&W's Sub-mittal on Combination of Peaking Factors," May 13, 1977.

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

Fuel Densification Report " December 5,1977.

I3 BPlu Retainer Design Report, BAW-1496 Babcock & Wilcox Lynchburg, Vir- .y ginia, May 1978.

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