ML15222A068

Cycle 7 Reload Rept.

ML15222A068
Person / Time
Site:	Oconee
Issue date:	04/30/1982
From:	DUKE POWER CO.
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ML15222A069	List: ML15222A068 ML16134A673 ML20052F240
References
DPC-RD-2001, NUDOCS 8205120278
Download: ML15222A068 (57)

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OCONEE UNIT 3, CYCLE 7

- Reload Report DPC - RD - 2001 April , 1982 Duke Power Company Steam Production Department P. 0. Box 33189 Charlotte, North Carolina 28242

CONTENTS Page

1. INTRODUCTION AND

SUMMARY

. . ......... . . . ..1-1

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

3. GENERAL DESCRIPTION....... ...... . . . ..3-1

4. FUEL SYSTEM DESIGN. . ............ . . . ..4-1 4.1. Fuel Assembly Mechanical Design . . . . . . . . . 4-1 4.1.1. Mark B5 Fuel Assembly 4-1 4.2. Fuel Rod Design . . . . . . . . . . . . . . . . . 4-2 4.2.1. Cladding Collapse . . . . . . . . . . . . 4-2 4.2.2. Cladding Stress . . . . . . . . . . . . . 4-3 4.2.3. Cladding Strain . . . . . . . . . . . . . 4-3 4.3. Thermal Design . . . . . . . . . . . . . . . . . . 4-3 4.4. Material Design . . . . . . . . . . . . . . . . . 4-4

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

6. THERMAL-HYDRAULIC DESIGN. .......... . . . . ..6-1

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

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . 8-1 REFERENCES. . .............. . . . . .. A-1

- 11 -

List of Tables Table Page 4-1. Fuel Design Parameters and Dimensions . ........ 4-5 4-2. Linear Heat Rate to Melt Analysis. . . . . . . . . . . 4-6 5-1. Oconee 3 Physics Parameters . . . . . . . . . . . . . . 5-3 5-2. Shutdown Margin Calculation for Oconee 3, Cycle 7 . . . 5-5 6-1. Thermal-Hydraulic Design Conditions . ......... 6-2 7-1. Comparison of Key Parameters for Accident Analysis ....................... 7-3 7-2. LOCA Limits, Oconee 3, Cycle 7, After 50 EFPD . . . . . 7-4 7-3. LOCA Limits, Oconee 3, Cycle 7, 0-50 EFPD..... . .. 7-4 List of Figures Figure 3-1. Core Loading Diagram for Oconee 3, Cycle 7......3-2 3-2. Enrichment and Burnup Distribution for Oconee 3, Cycle 7...... ............... 3-3 3-3. Control Rod Locations for Oconee 3, Cycle 7......3-4 3-4. BPRA Enrichment and Distribution for Oconee.3, Cycle 7 ...... ............... 3-5 4-1. Mark B5 Upper End Fitting. .... ............ 4-7 4-2. Mark B5 Holddown Spring Retainer. .. .............. 4-8 4-3. Mark B5 Fixed Control Component Spider .. .. ....... 4-9 4-4. Mark B5 Coupling - Spider Assembly .. ... ........ 4-10 4-5. Mark B5 Fixed Control Component Spider/

Upper End Fitting Interaction. .. ............ 4-11 5-1. BOC Cycle 7 Two-Dimensional Relative Power Distribution Full Power, Equilibrium Xenon, Normal Rod Positions . . 5-6 8-1. Core Protection Safety Power-Imbalance Limits.....8-2 8-2. Core Protection Safety Pressure-Temperature Limits. . 8-3 8-3. Core Protection Pressure-Temperature Limits......8-4 8-4. Maximum Allowable Power-Imbalance Setpoints......8-5 8-5. Operational Power-Imbalance Limits 0-50 +/- 1 0 8-6. Ypiaf al Power-Imbalance Limits 50 +/- 10.........8-7 8-7. Operational Power-Imbalance Limits .i After 200 +/- 10 EFPD 8-8 Control Rod Position Limits, 4 Pumps, 0-50 +/- '8 EFPD.

MarkB5 oldownSprng etaner 8-8. 8-9 8-9. Control Rod Position Limits, 4 Pumps, 50 +/- 18-200

+/-10 EFPD,.. .................. ..... 8-10 8-10. Control Rod Position Limits, 4 Pumps, After 200 +/- 10 EFPD 8-11 8-11. Control Rod Position Limits, 3 Pumps, 0-50 18 EFPD. 8-12 8-12. Control Rod Position Limits, 3 Pumps, 50 +/- '8-200

+/-o10 EFPDi . . . . .C 7..........

....... 8-13 8-13. Control Rod Position Limits, 3 Pumps, After 200 10 EFPD .E .. ... i.. . . ............. 8-14

List of Figures (cont)

Figure Page 8-14. Control Rod Position Limits, 2 Pumps, 0-50 18 EFPD. 8-15 8-15. Control Rod Position Limits, 2 Pumps, 50 +/- 8-200

+/- 10 EFPD ....................... 8-16 8-16. Control Rod Position Limits, 2 Pumps, After 200

+/- 10 EFPD... . . . . . . . . . . . . . . . - .. 8-17 8-17. APSR Position Limits, 0-200 +/- 10 EFPD . . . . . . . . .8-18 8-18. APSR Position Limits, After 200 +/- 10 EFPD . . . . . . . 8-19

- iv -

1. INTRODUCTION AND

SUMMARY

This report justifies the operation of the seventh cycle of Oconee Nuclear Station, Unit 3, at the rated core power of 2568 MWt. Included are the required analyses as outlined in the USNRC document "Guidance for Proposed License Amendments Relating to Refueling," June 1975.

To support cycle 7 operation of Oconee Unit 3, this report employs analytical techniques and design bases established in reports that were previously sub mitted and accepted by the USNRC and its predecessor (see references).

A brief summary of cycle 6 and 7 reactor parameters related to power capabi lity is included in section 5 of this report. All of the accidents analyzed in the FSAR' have been reviewed for cycle 7 operation. In those cases where cycle 7 characteristics were conservative compared to those analyzed for pre vious cycles, no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications required for cycle 7 operation are justified in this report.

Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cooling Systems, it has been concluded that Oconee Unit 3 can be operated safely for cycle 7 at the rated power level of 2568 MWt.

1-1

2. OPERATING HISTORY The referenced fuel cycle for the nuclear and thermal-hydraulic analyses of Oconee Unit 3, cycle 7, is the currently operating cycle 6. Cycle 5 was ter minated after 309 EFPD of operation. Cycle 6 achieved initial criticality on March 12, 1981 and power escalation commenced on March 14, 1981. The fuel cycle design length for cycle 7 - 421 EFPD - is based on cycle 6 length of 376 EFPD. No operating anomalies occurred during previous cycle operations that would adversely affect fuel performance in cycle 7.

Cycle 7 will operate in a feed-and-bleed mode for its entire design length, as did cycle 6.

2-1

3. GENERAL DESCRIPTION The Oconee Unit 3 reactor core and fuel design basis are described in detail in Chapter 3, of the FSAR.' The cycle 7 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 instrument guide tube. The fuel consists of dished-end, cylindrical pellets 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 uranium. The undensified nominal active fuel 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 3, cycle 7. Nineteen of the batch 7 assemblies will be discharged at the end of cycle 6 along with batches 5B, and 6. The remaining 37 batch 7 assemblies, designated "7B,"

and the fresh batch 9 FAs - with initial enrichments of 2.80 and 3.18 wt

% 2 35 U, respectively - will be loaded into the central portion of the core.

Batch 8, with an initial enrichment of 3.07 wt % 2 3 sU, will occupy primarily the core periphery. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 7.

Cycle 7 will operate in a rods-out, feed-and-bleed mode. Core reactivity con trol is supplied mainly by soluble boron and supplemented by 61 full-length Ag-In-Cd control rods and 64 burnable poison rod assemblies (BPRAs). In addi tion to the full-length control rods, eight partial-length axial power shaping rods (APSRs) are provided for additional control of axial power distribution.

The cycle 7 locations of the 69 control rods and the group designations are indicated in Figure 3-3. The cycle 7 locations and enrichments of the BPRAs are shown in Figure 3-4.

3-1

FIGURE 3-1. CORE LOADING DIAGRAM FOR OCONEE 3 CYCLE 7 X

A ~NO3 G14 N05 G021 N13 A 8 8 88 8 8 8

-8 8

M02 FIl M04 F05 M14 B 8 9 8 9 8 9 8 9 8 K08 F13 K04 K12 F03 H07 C 8 9 8 g 8 9 8 9 8 9 8 811 H13 A07 MOB A09 008 B05 D8 9 8 9 7B 9 8 9 7B 9 8 9 8 C13 010 006 C03 A06 N02 A10 E 9 8 9 7B 9 7B 78 7B 9 7B 9 8 9 E-C12 MO6 G01 HO1 P04 A08 G15 M1O C04 F 8 8 9 7B 9 78 9 7B 9 7B 9 7B 9 8 8 D09 F01 F02 810 F15 007 P09 P07 G 8 9 8 9 7B 9 7B 9 7B 9 7B 9 8 9 8 M12 Nl H1l P12 N14 P10 002 B04 H05 005 E04 Y HW- 8 B 9 8 7B 7B 9 lB 9 7B 7B B 9 8 8 807 N09 LO1 P06 L14 L15 N07 809 K 8 9 8 9 7B 9 7B 9 7B 9 7B 9 8 9 8 K01 RO8 B12 H15 K15 E10 004 012 E06 L 8 8 9 7B 9 7B 9 7B 9 7B 9 7B 9 8 8 R06 014 RIO 013 Clo C06 003

-m M _ 9 8 9 7B 9 7B 7B 7B 9 7B 9 8 9 P11 C08 R07 EO8 . R09 H03 P05 N 8 9 8 9 78 9 8 9 7B 9 8 9 8 H09 L13 G04 G12 L03 GOB 0 B 9 8 9 8 9 B 9 8 EO2 L1I E12 L05 E14

_ _ _ _ _ _ 8 9 8 9 8 003 K14 Dl K02 013

_ _B 8 E 8 B 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 XX PREVIOUS CYCLE LOCATION X BATCH NO.

3-2

FIGURE 3-2. ENRICHMENT AND BURNUP DISTRIBUTION FOR OCONEE 3, CYCLE 7 8 9 10 11 12 13 14 15 2.80 3.18 2.80 2.80 3.07 3.18 3.07 3.07 H 20894 0 12088 12085 14660 0 14715 14648 2.80 3.18 2.80 3.18 3.07 3.18 3.07 K 20890 0 12120 0 15111 0 13600 2.80 3.18 2.80 3.18 3.07 3.07 L 15488 0 15315 0 14667 11819 2.80 3.18 3.07 3.18 M

13489 0 14948 0 3.07 3.18 3.07 15146 0 10339 3.07 14287 bL P

R X.XX INITIAL ENRICHMENT, wt % 235U XXXXX BOC BURNUP, MWd/mtU 3-3

FIGURE 3-3. CONTROL ROD LOCATIONS FOR OCONEE 3, CYCLE 7 X

A B 3 7 3 C 1 6 6 1 D 7 8 4 8 7 E 1 5 2 2 5 1 F 3 8 7 5 7 8 3 G 6 2 4 4 2 6 HW- 7 4 5 3 5 4 7 - Y K 6 2 4 4 2 6 L 3 8 7 5 7 8 3 M 1 5 2 2 5 1 N 7 8 4 8 7 0 1 6 6 1 P 3 7 3 R

z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 GROUP NO. of RODS FUNCTION 1 8 SAFETY

[j IGROUP NO. 2 8 SAFETY 3 9 SAFETY 4 8 SAFETY 5 8 CONTROL 6 8 CONTROL 7 12 CONTROL 34 8 8 APSRs TOTAL69

FIGURE 3-4. BPRA ENRICHMENT AND DISTRIBUTION FOR OCONEE 3, CYCLE 7 8 9 10 11 12 13 14 15 H 1.4 1.4 1.4 1.4 0.2 K

L 1.4 0.5 M 1.0 N 0.2 0

SL P

R X.X BPRA CONCENTRATION, wt % B4 C IN A12 03 3-5

4. FUEL SYSTEM DESIGN 4.1 Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for Oconee 3, cycle 7, are listed in Table 4-1. All fuel assemblies are identical in concept and are mechanically interchangeable. Four regenerative neutron sources will be used: two will be contained in MK B5 fuel assemblies and two in MK B4 assemblies. Retainer assemblies will be used on the two MK B4 FAs containing the regenerative neutron sources. The justification for the design and use of the BPRA retainers is described in reference 3 and 21, which is also applicable to the RNS retainers of Oconee 3, cycle 7.

The batch 9 Mark B5 fuel assemblies have redesigned upper end fittings which provide a positive holddown of BPRAs. Section 4.1.1 describes the design features of this end fitting. All 64 BPRAs will be inserted into batch 9 fuel assemblies.

Other results presented in the FSARI fuel assembly mechanical discussions and in previous reload reports are applicable to the reload fuel assemblies. Duke has performed generic mechanical analyses, as described below, which envelope the cycle 7 design. All methods are consistent with the approved methodolo gies of Reference 16 except where specifically stated.

4.1.1 Mark B5 Fuel Assembly Batch 9 fuel assemblies are Babcock & Wilcox Mark B5 fuel assemblies (FA's).

The Mark B5 assembly is identical to the Mark B4 except that its upper end fitting has been developed to provide a positive holdown of fixed control components such as burnable poison rod assemblies, neutron source rod assemblies, and orifice rod assemblies (should reinsertion of orifice rod assemblies be desirable to minimize core bypass flow). The B4 and B5 FA's function identically with existing handling equipment and movable control components, such as control rod assemblies and axial power shaping rod assemblies.

A spring loaded retainer assembly, references 3 and 21, is used with the Mark B4 FA design to insure positive holddown of the fixed control components at all design flow conditions. A locking-ball coupling attaches the control components to the FA.

4-1

The Mark B5 upper end fitting, Figure 4-1, provides four open slots that align and allow designed movement of the holddown spring retainer, Figure 4-2, and the B5 fixed control component spider, Figure 4-3 and 4-4. The holddown spring used in the B5 FA will provide positive holddown capability, with or without a fixed control component installed, for all design flow conditions.

The holddown spring is preloaded through a stop pin, welded to an ear on each side of the upper end fitting. In core,, the spider feet are captured between the holddown spring retainer and the upper grid pads on the reactor internals as shown in Figure 4-5. This arrangement retains the B5 fixed control com ponents at all design flow conditions.

Mark B5 fixed control component assemblies are not compatible with B4 FA's for in core operation and vice versa. Cycle 7 has been designed to preclude mixing of control component designs and this will be verified by video prior to plenum installation.

It has been determined that no special treatment of the B5 assembly is required for core reload design analyses. The upper end fitting form loss coefficient remains significantly unchanged, and the fuel rod design remains unchanged.

Therefore, the thermal-hydraulic and fuel rod mechanical analyses are un affected.

4.2 Fuel Rod Design The mechanical evaluation of the fuel rod is discussed below.

4.2.1 Cladding Collapse The fuel of batch 7B is more limiting than other batches due to its longer previous incore exposure time. The batch 7B assembly power histories were analyzed, and the most limiting assembly was used to perform the creep collapse analysis using the CROV computer code and procedures described in topical report BAW-10084, Rev. 3.2 The TACO 4 code was used to calculate internal pin pressures and clad temperatures used as input to CROV. The col lapse time for the most limiting assembly was conservatively determined to be more than 35,000 EFPH, which is greater than the maximum projected residence time of cycle 7 fuel (Table 4-1).

4-2

4.2.2 Cladding Stress Duke has performed a generic and conservative fuel rod cladding stress analy sis. This analysis is consistent with the methodology described in Reference 16 with the following exception: the fuel rod total stress (primary plus secondary) was permitted to exceed the unirradiated yield strength. Two times the minimum unirradiated yield strength (2.0 Sy) has been used as a criterion for the total stress calculation, as permitted by Section III, Article NB-3000 of the ASME Boiler and Pressure Vessel Code. Approximately 0.55 Sy margin remains in this total stress calculation.

Primary membrane plus primary bending stresses are limited to 1.0 Sy, and primary membrane stress is limited to 2/3 Sy. Substantial margin exists in both of these evaluations.

The following conservatisms exist in the generic cladding stress calculation:

• a low internal pressure (HZP);

• a high external pressure (110 percent of design pressure);

• a large through wall cladding temperature gradient (fuel melt conditions), and

• BOL grid loads for worst grid cell type 4.2.3 Cladding Strain Duke has performed a cladding strain calculation using TACO in accordance with the approved methodology.'6 This analysis demonstrated that the uniform, cir cumferential strain of the cladding was within 1.0%.

4.3. Thermal Design All fuel in the cycle 7 core is thermally similar. The fresh batch 9 fuel inserted for cycle 7 operation introduces no significant differences in fuel thermal performance relative to the other fuel remaining in the core. The linear heat rate to melt capability based on centerline fuel melt was deter mined for each batch of fuel using the TACO computer code. The fuel para meters used to determine the fuel melt limits are shown in Table 4-2. With respect to Oconee 3 Cycle 7 fuel, the input shown includes the following conservatisms:

4-3

1. A lower initial density.

2. A smaller initial pellet diameter.

The design minimum linear heat rate (LHR) capability and the average fuel temperature for each batch of fuel in cycle 7 are shown in Table 4-2.

Reference 16, Section 4.6, states that "no credit is taken for fuel relocation in LHRTM analyses". This is an error. Fuel relocation is assumed in these analyses in that relocation is an integral part of the TACO model. However, credit for restructuring is not assumed in these analyses, in accordance with Reference 4.

Fuel rod internal pressure has been evaluated using TACO with a conservative pin power history, and the maximum pressure is less than the nominal reactor coolant (RC) system pressure of 2200 psia.

4.4. Material Design The batch 9 fuel assemblies are not unique in concept (excluding the upper end fitting design modification of the Mark B5 fuel assembly), nor do they utilize different component materials. Therefore, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 9 fuel assemblies is identical to those of the present fuel.

4-4

Table 4-1. Fuel Design Parameters and Dimensions Batch No.

7B 8 9 FA type Mark B4 Mark B4 Mark B5 No. of FAs 37 68 72 Fuel rod OD, in. 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 Flex spacers, type Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Undensif active fuel 142.2 141.8 141.8 length, in.

Fuel pellet OD (mean 0.3695 0.3686 0.3686 spec), in.

Fuel pellet initial 94.0 95.0 95.0 density (mean spec),

%TD Initial fuel enrich- 2.80 3.07 3.18 ment, wt % 2 3 5u Est residence 26,544 19,128 10,104 time, EOC 7, EFPH Cladding collapse >35,000 >35,000 >35,000 time, EFPH 4-5

Table 4-2. Linear Heat Rate to Melt Analysis Batch No.

7B 8 9 Initial density, % TD 93.5 94.0 94.0 Max. In-reactor densification, % TD 2.7 2.2 2.2 Burnup corresponding to 3900 2300 2300 max. densification, MWd/mtU Initial pellet diameter, in. 0.3694 0.3680 0.3680 Average linear heat rate @ 5.73 5.74 5.74 100% of 2568 MW, kW/ft Linear heat rate capability >20.15 >20.15 >20.15 (centerline fuel melt), kW/ft Average fuel temp. @ nominal 1250(a) 1240 1240 LHR, OF (a) TACO, 96.5 TD @ 4000 MWd/mtU 4-6

FIGURE 4-1 MARK B5 UPPER END FITTING (SIDE VIEW)

SLOT OP EAR 4-7

FIGURE 4-2 MARK B5 HOLDDOWN SPRING RETAINER FOOT ARM STOP PIN LEDGE RING 4-8

FIGURE 4-3 MARK B5 FIXED CONTROL COMPONENT SPIDER (TOP VIEW)

FOOT LEG Y ARM O O TRAIGHT ARM HUB 4-9

FIGURE 4-4 MARK B5 COUPLING - SPIDER ASSY - SIDE VIEW (SECTION)

SPIDER 4-10

FIGURE 4-5 MARK B5 FIXED CONTROL COMPONENT SPIDER/UPPER END FITTING INTERACTION UPPER GRID PAD COUPLING -SPIDER ASSEMBLY STOP PIN T- HOLODOWN SPRING RETAINER HOLODOWN SPRING SPRI UEF EAR 4-11

5. NUCLEAR DESIGN 5.1 Physics Characteristics Table 5-1 compares the core physics parameters of design cycles 6 and 7; the values for cycle 6 were generated by B&W 6, 7, 8, 13, 15 using PDQ07 while the values for cycle 7 were generated by Duke Power Company using methods described in Reference 16. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. The longer cycle 7 will produce a higher cycle burnup than that for the design cycle 6. Figure 5-1 illustrates a representative relative power distribution for the beginning of the seventh cycle at full power with equili brium xenon and normal rod positions.

The initial BPRA loading, longer design life, different shuffle pattern, and different control rod pattern for cycle 7 make it difficult to compare the physics parameters with those of cycle 6. The BOC critical boron concentrations for cycle 7 are higher because the additional reactivity necessary for the longer cycle is not completely offset by burnable poison. The control rod worths differ between cycles primarily due to changes in control rod patterns.

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. All safety criteria associated with these rod worths are met. The adequacy of the shutdown margin with cycle 7 stuck 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 worth.

5-1

Flux redistribution was explicitly accounted for since the shutdown analysis was calculated using a three-dimensional model. The reference fuel cycle shutdown margin is presented in the Oconee 3, cycle 6 reload report. 5 The cycle 7 power deficits, differential boron worths, and effective delayed neutron fractions differ from those of cycle 6 because of the longer cycle length and differences in core loading.

5.2 Analytical Input The cycle 7 incore measurement calculation constants to be used to compute core power distributions were obtained in the same manner for cycle 7 as for the reference cycle. CASM0 1 7 was used to verify the F-factors derived from B&W's codes.

5.3 Changes in Nuclear Design There are only two significant core design changes between the reference cycle and the reload cycle. The cycle lifetime is increased to 421 EFPD requiring an increase in the number of fresh fuel assemblies and BPRAs.

Duke. Power calculational methodsl 6 are used to obtain the important nuclear design parameters for this cycle.

5-2

Table 5-1. Oconee 3 Physics Parameters(a)

Cycle 6 (b) Cycle 7 (c)

Cycle length, EFPD 376 421 Cycle burnup, MWd/mtU 11,766 13,156 Average core burnup, EOC, MWd/mtU 20,231 21,486 Initial core loading, mtU 82.1 82.1 Critical boron - BOC (no xenon), ppm HZP, group 7 at 100% WD, 8 at 37.5% WD 1471 1572 HFP, group 7 at 87% WD, 8 at 25% WD 1282 1385 Critical boron - EOC (equil xenon), ppm HZP, group 7 at 100% WD, 8 at 37.5% WD 385 445 HFP, group 7 at 87% WD, 8 at 25% WD 78 18 Control rod worths - HFP, BOC, % Ak/k Group 6 0.98 1.21 Group 7 1.36 1.47 Group 8 (25% to 100% WD) 0.50 0.33 Control rod worths - HFP, EOC(d), % Ak/k Group 7 1.48 1.64 Group 8 (25% to 100% WD) 0.54 0.30 Max ejected rod worth - HZP, % Ak/k BOC, (N12) groups 5-8 inserted 0.38 0.72 EOC, (N12) groups 5-8 inserted 0.51 0.75 Max stuck rod worth - HZP,  % Ak/k BOC (N12) 1.39 1.51 EOC (N12) 1.52 2.03 Power deficit, HZP to HFP, % Ak/k BOC 1.39 1.80 EOC 2.22 3.06 Doppler coeff - BOC, 10- 5 (Ak/k-oF) 100% power (no xenon) -1.49 -1.34 Doppler coeff - EOC, 10- 5 (Ak/k-oF) 100% power (equil xenon) -1.62 -1.68 5-3

Table 5-1. (Cont'd)

Cycle 6 (b) Cycle 7 (c)

Moderator coeff - HFP, 10 (Ak/k-oF)

BOC (no xenon, 1325 ppm, group 8 ins.) -0.65 -0.40 EOC (equil xenon, 17 ppm, group 8 ins.) -2.82 -2.84 Boron worth - HFP, ppm/% Ak/k BOC (1070 ppm) 116 120 EOC (67 ppm) 102 108 Xenon worth - HFP, % Ak/k BOC (4 days) 2.61 2.50 EOC (equilibrium) 2.74 2.70 Eff delayed neutron fraction - HFP BOC 0.00628 0.00626 EOC 0.00526 0.00520 (a)Cycle 7 data are for the conditions stated in this report.

The cycle 6 core conditions are identified in reference 5.

(b)Based on a 299-EFPD cycle 5. (Actual cycle length 309 EFPD).

(c)Based on 376-EFPD cycle 6.

376 EFPD in cycle 6, 421 EFPD in cycle 7.

5-4

Table 5-2. Shutdown Margin Calculation for Oconee 3, Cycle 7 BOC, EOC,

% Ak/k  % Ak/k Available Rod Worth Total rod worth, HZP 8.25 9.15 Worth reduction due to poison burnup -0.42 -0.42 Maximum stuck rod, HZP -1.51 -2.03 Net worth 6.32 6.70 Less 10% uncertainty -0.63 -0.67 Total available worth 5.69 6.03 Required Rod Worth Power deficit, HFP to HZP 1.80 3.06 Max inserted rod worth, HFP 0.23 0.53 Total required worth 2.03 3.59 Shutdown Margin Total available worth minus total 3.66 2.44 required worth Note: Required shutdown margin is 1.00% Ak/k.

5-5

FIGURE 5-1 OCONEE 3 CYCLE 7 TWO DIMENSIONAL RELATIVE POWER DISTRIBUTION HFP, 004 EFPD, EQXE NOMINAL ROD POSITIONS 8 9 10 11 12 13 14 15 H 0.828 1.050 1.047 1.081 1.163 1.300 1.016 0.541 K 0.868 1.153 1.106 1.267 1.199 1.221 0.540 L 1.002 1.199 1.007 1.309 0.927 0.421 M 1.109 1.250 1.072 0.888 N 1.074 1.067 0.504 0 0.536 L

P R

5-6

6. THERMAL-HYDRAULIC DESIGN The incoming batch 9 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The thermal-hydraulic de sign analysis supporting cycle 7 operation was performed by Duke Power Company and employed the methods and models described in references 1, 5, 9 and 16.

The maximum core bypass flow for cycle 6 was 8.1% of the total system flow.

For cycle 7 operation, 64 BPRAs will be inserted, and four assemblies contain regenerative neutron sources. The number of open assemblies is 40, and the maximum core bypass flow is reduced to 7.6.%. The cycle 6 and 7 maximum design conditions are summarized in Table 6-1.

A rod bow DNBR penalty has been calculated for cycle 7 operation according to procedures approved by reference 10. The burnup used to calculate the penalty is the highest batch 9 burnup, 16,945 MWd/mtU. The burnup/pin power relation ships of batches 7 and 8 are enveloped by that of batch 9. The net rod bow penalty1 8 is 0.0% after taking credit for the flow area reduction hot channel factor used in all DNBR calculations. For cycle 7 operation a flux to flow setpoint of 1.08 is maintained. The minimum DNBR value determined by the flux to flow setpoint analysis is above the design minimum DNBR of 1.30. However, all other plant operating limits based on DNBR criteria included a minimum of 10.2% DNBR margin from the B&W-2 correlation design limit of 1.30.

6-1

Table 6-1. Thermal Hydraulic Design Conditions Cycle 6 Cycle 7 Design power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design flow 106.5 106.5 Core bypass flow, % total flow 8.1 7.6 Vessel inlet/outlet coolant temp at 555.6/602.4 555.6/602.4 100% power, oF Ref design radial-local power 1.71 1.71 peaking factor Ref design axial flux shape 1.5 cosine 1.5 cosine Hot channel factors: Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Active fuel length, in. (a) (a)

Avg heat ljx at 100% power, 103 1 76 (b) 17 6 (b)

Btu/h- ft2 CHF correlation BAW-2 BAW-2 Min DNBR with densification penalty 2.05 >2.05 (a)See Table 4-1.

(b)Heat flux based on densified length of 140.3 in., which is a con servative minimum value.

6-2

7. ACCIDENT AND TRANSIENT ANALYSIS 7.1 General Safety Analysis Each FSAR' accident analysis has been examined with respect to changes in cycle 6 parameters to determine the effect of the cycle 7 reload and to ensure that thermal performance during hypothetical transients is not degraded. The effects of fuel densification on the FSAR accident results has been evaluated and are reported in reference 9. Since batch 9 reload fuel assemblies contain fuel rods with a theoretical density higher than those considered in reference 9, the conclusions in that reference are still valid.

No new dose calculations were performed for the reload report. The dose con siderations in reference 20 are conservative for Oconee 3 cycle 7 based upon comparisons of core average burnup for the two cycles.

7.2 Accident Evaluations The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: core thermal parameters, thermal-hydraulic parameters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.

Fuel thermal analysis parameters for each batch in cycle 7 are given in Table 4-2. Table 6-1 compares the cycle 6 and 7 thermal-hydraulic maximum design conditions. Table 7-1 compares the key kinetics parameters from the FSAR and cycle 7.

A generic LOCA analysis for the B&W 177-FA, lowered-loop NSS has been per formed using the Final Acceptance Criteria ECCS Evaluation Model. This study is reported in BAW-10103, Rev. 1" 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 7-1

function of LHR and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis 11 2 is conservative compared to those calculated for this re load. Thus, the analysis and the LOCA limits reported in BAW-10103 provide conservative results for the operation of Oconee 3, cycle 7 fuel.

Table 7-2 shows the bounding values for allowable LOCA peak LHRs for Oconee 3 cycle 7 fuel after 50 EFPD. The LOCA kW/ft limits have been reduced for the first 50 EFPDs. The reduction will ensure that conservative limits are maintained while a transition is being made in the fuel performance codes that provide input to the ECCS analysis' 9 in order to account for mechanistic fuel densification. The limits for the first 50 EFPD are shown in Table 7-3.

From the examination of cycle 7 core thermal properties and kinetics proper ties with respect to acceptable previous cycle values, it is concluded that this core reload will not adversely affect the safe operation of the Oconee 3 plant during cycle 7. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 7 is con sidered to be bounded by previously accepted analyses. The initial conditions of the transients in cycle 7 are bounded by the FSAR and/or the fuel densifi cation report. 9 7-2

Table 7.1. Comparison of Key Parameters for Accident Analysis FSAR' Predicted Parameter value cycle 7 value BOC Doppler coeff, 10-5, Ak/k/oF -1.17 -1.34 EOC Doppler coeff, 10- 5 Ak/k/oF -1.3 3 (a) -1.68 BOC moderator coeff, 10-4, Ak/koF +0.5(b) -0.40 EOC moderator coeff, 10-4, Ak/k/oF -3.0 -2.84 All rod bank worth, HZP, % Ak/k 10.0 9.15 0

Boron reactivity worth, 70 F ppm/1% Ak/k 75 83 Max. ejected rod worth, HFP, % Ak/k 0.65 0.20 Dropped rod worth, HFP, % Ak/k 0.46 0.12 Initial boron conc, HFP, ppm 1400 1385 (a)-1.2 x 10-s Ak/k/F was used for steam-line analysis.

-1.3 x 10- 5 Ak/k/F was used for cold water accident (pump start-up).

4 (b)+0.94 x 10 Ak/k/F was used for the moderator dilution accident.

7-3

Table 7-2. LOCA Limits, Oconee 3, Cycle 7, After 50 EFPD Elevation, LHR limits, ft kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 Table 7-3. LOCA Limits, Oconee 3, Cycle 7 0-50 EFPD Elevation, LHR Limits, ft kW/ft 2 14.5 4 16.1 6 17.5 8 17.0 10 16.0 7-4

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 7 operation in accordance with the methods of reference 16 to account for minor changes in power peaking and control rod worths inherent with a transition to 18-month, lumped burnable poison cycles. Cycle 6 Technical Specifications were generated in accordance with the methods described in Reference 14.

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 8-18 are revisions to previous Technical Specification limits.

8-1

Figure 8-1 Core Protection Safety Power-Imbalance Limits Thermal Power Level, %

-120 M1=0.638 (-36.5,112.0) 112.0 (33.0,112.0)

-110

(-49.5,103.70) ACCEPTABLE 4 PUMP M2=-1.864 OPERATION -100 90.65

-90 ACCEPTABLE

(-49.5,82.35) 3&4 PUMP 8 OPERATION OPERTION80 I(49.5,81.25)

-70 63.26 ACCEPTABLE -60 (49.5,59.90)

(-49.5,54.90) 2,3,&4 PUMP OPERATION - 50

- 40 UNACCEPTABLE OPERATION 30 (49.5,32.51)

I -I2 20 UNACCEPTABLE LQl OPERATION

-1 0I cOa co

-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Reactor Power Imbalance, %

8-2

Figure 8-2 Core Protection Safety Pressure-Temperature Limits 2400 ACCEPTABLE OPERATION 2200 C.

2000 0

1800 UNACC PTABLE OPERA' ION 1600 580 600 620 640 660 Reactor Coolant Core Outlet Temperature, 0 8-3

Figure 8-3 Core Protection Pressure-Temperature Limits 2400 ACCEPTABLE /

OPERAT ON /

2200

0. 4 PUMP",,

2000' 0

0 / PUMP 3PPUMP/

1800 UNACC EPTABLE OPERATION 1600 580 600 620 640 660 Reactor Coolant Core Outlet Temperature, OF PUMPS COOLANT POWER TYPE OF LIMIT OPERATING FLOW (GPM) (% FP) 4 374,880(100%) 112.0 DNBR 3 280,035(74.7%) 90.7 DNBR 2 183,690(49.0%) 63.63 DNBR/QUALITY 8-4

Figure 8-4 Maximum Allowable Power-Imbalance Setpoints Thermal Power Level, %

-120

(-16.0,108.0) 108.0 -110 (16.5,108.0)

ACCEP ABLE 4 M1=0.992 PUMP OF ERATION M2=-1.890 1- 100

(-33.5,90.64) 80.67 UNACCEPTABLE 80. U OPERATION ACCEPT ABLE 3&4 (

I PUMP OERATION (33.5,75.94)

I - 70

(-33.5,63.31) UNACCEPTABLE

- 60 OPERATION 52.92I

- 50 (354.1 ACCEPTABLE 2,3,&4N(335,.61)

PUMP OPERATION

-40

(-33.5,35.56)

- 30 I I I- 20 (33.5,20.86)

L 9 I 10 I1 I I -lC' C\JI C-I I II I II I n sI I

-50 -40 -30 -20 -10 0 10 20 30 40 50 Reactor Power Imbalance, %

8-5

Figure 8-5 Operational Power-Imbalance Limits 0-50 +/- EFPD REACTOR POWER,%FP

(-21.3,102.0) -100 (25.0,102.0)

(-25.0,92.0) ACCEP 'AB LE (30.0,92.0)

OPER ATION

(-30.0,80.0) 80

- 60 RESTRICTED OPERATION RESTRICTED OPERATION

- 40

- 20

-100 -80 -60 -40 -20 0 20 40 60 80 100 IMBALANCE,%

8-6

Figure 8-6 Operational Power-Imbalance Limits 50 +/ 200 +/- 18 EFPD REACTOR POWER,%FP

(-28.9,102.0) 100 (25.0,102.0) 7ACCEPrABLE

(-31.8,92.0) 0ACP ABLE (30.0,92.0)

OPERATION 80 60 RESTRICTED OPERATION RESTRICTED OPERATION 40 20 Ill i I Il

-100 -80 -60 -40 -20 0 20 40 60 80 100 IMBALANCE,%

8-7

Figure 8-7 Operational Power-Imbalance Limits After 200 ! 10 EFPD REACTOR POWER,%FP

(-30.0,102.0) -100 (25.0,102.0)

(-32.7,92.0) ACCEP ABLE P A (30.0,92.0)

OPERATION

- 80

- 60 RESTRICTED OPERATION RESTRICTED OPERATION

- 40

- 20

-100 -80 -60 -40 -20 0 20 40 60 80 100 IMBALANCE,%

8-8

Figure 8-8 Control Rod Position Limits, 4 Pumps, 0-50 + 1 EFPD POWER LEVEL (300,102) 100- (150,102) CUTOFF=100% FP (2 8 0

07 - s (275,92) 80- SHUTDOWN (270,80)

MARGIN LIMIT RESTRICTED OPERATION LL

- 60 LU

i. UNACCEPTABLE L OPERATION (90,50) (200,50) 0 40 ACCEPTABLE OPERATION 20 (40,15) (90,15)

(0,10)<

(0,5) 0 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 I 5 I BANK 5 0 25 50 75 100 II I BANK 6 0 25 50 75 100 I I II BANK 7 8-9

Figure 8-9 Control Rod Position Limits, 4 Pumps, 50 + 18 - 200 + 10 EFPD (150,102) POWER LEVEL (275,102) (300,102 100- 100 ~CUTOFF=100% FP _ _

SHUTDOWN (260,92)

MARGIN LIMIT 80- (250,80)

RESTRICTED OPERATION UNACCEPTABLE 60- OPERATION x

0 C. (90,50) (200,50) 0 L 40 ACCEPTABLE OPERATION 20 (40,15) (90,15)

(0,10)*

- - (0,5)<

0 I I I I I 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 BANK5 0 25 50 75 100 BANK6 0 25 50 75 100 BA 6 2 5 7 0 BANK7 8-10

Figure 8-10 Control Rod Position Limits, 4 Pumps, After 200 + 10 EFPD POWER LEVEL CUTOFF=100% FP (300,102)

(220,102) _(275,102) 100-- - - -

(260,92) 80- (250,80)

UNACCEPTABLE RESTRICTED OPERATION OPERATION CL 60 a- (160,50) (200,50) 0 u 40 SHUTDOWN ACCEPTABLE MARGIN OPERATION LIMIT.

20 (100,15)

(0,5)<

0 I I I .

0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50. 75 100 Il I 5 BANK5 0 25 50 75 100 i l l I BANK 6 25 50 75 100 1IIll BANK T*

- 8-11

Figure 8-11 Control Rod Position Limits, 3 Pumps, 0-50 + 18 EFPD 100 80- (300,77)

(130,77) (263,77) (

U 60 Wj UNACCEPTABLE 0 OPERATION RESTRICTED (90,50) OPERATION (200,50) 0 S40 cr SHUTDOWN MARGIN ACCEPTABLE LIMIT OPERATION 20 (30,15) (90,15)

(0,10)<

(0,5) 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 Il I BANK5 0 25 50 75 100 I I III BANK 6 0 25 50 75 100 I Ill BANK 7 8-12

Figure 8-12 Control Rod Position Limits, 3 Pumps, 50 + 18 - 200 + 10 EFPD 100 80 -(0

,7 (130,77) (245,77)

CL

- 60 UNACCEPTABLE OPERATION LU RESTRICTED O OPERATION C.. (90,50)

(9050 (200,50) 4 040 SHUTDOWN MARGIN ACCEPTABLE L OPERATION 20 (30,15) (90,15)

(0,5)<

0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 BANK 5 0 25 50 75 100 I I I BANK 6 0 25 50 75 100 I I I I I BANK 7 8-13

Figure 8-13 Control Rod Position Limits, 3 Pumps, After 200 + 10 EFPD 100 RESTRICTED OPERATION 80 80- (210,77) (300,77),

,(245,77) 0 4U 60 (L

UNACCEPTABLE

~(160,50) (0,0 0I OPERATION o 40 SHUTDOWN ACCEPTABLE MARGIN OPERATION LIMIT 20 (100,15)

(0,5)<

0 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100

. 1 I I I BANK 5 0 25 50 75 100 BANK 6 0 25 50 75 100 I 1I BANK 7 8-14

Figure 8-14 Control Rod Position Limits, 2 Pumps, 0-50 + 1 EFPD 100 80 0L

' LL 60 (205,52) (300,52) 0 (80,52) a a- (200,50) 0 F

0 40 SHUTDOWN* RESTRICTED MA RGIN OPERATION LIMIT ACCEPTABLE OPERATION 20 - UNACCEPTABLE OPERTION (50,15) (90,15)

- - (0,10)<

(0,5)*

0 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 I I 5I I BANK 5 0 25 50 75 100 I I 11 I BANK 6 0 25 50 75 100 I I I II BANK 7 8-15

Figure 8-15 Control Rod Position Limits, 2 Pumps, 50 + 18 - 200 + 10 EFPD 100 80 LL 60

( , (203,52) (300,52)

(200,50) 0

- 40 U-i SHUTDOWN RESTRICTED MARGIN OPERATION ACCEPTABLE LIMIT OPERATION 20 - UNACCEPTABLE OPER TION (50,15) (90,15)

(0,10)<

- - (0,5)<

0 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 I A I I BANK 5 0 25 50 75 100 SI I I I BANK 6 0 25 50 75 100 I I Il l BANK 7 8-16

Figure 8-16 Control Rod Position Limits, 2 Pumps, After 200 + 10 EFPD 100 80 U RESTRICTED

- 60- OPERATION U

(10,2 (203,52) (300,52) 0 UNACCEPTABLE (200,50) o OPERATION o 40 U

SHUTDOWN ACCEPTABLE MARGIN OPERATION LIMIT 20 (110,15)

(0,5)<

0 0 50 100 150 200 250 300 ROD INDEX,%WD 0 25 50 75 100 I 5 I BANK 5 0 25 50 75 100 I 1 I BANK 6 0 25 50 75 100 BANK 7 8-17

Figure 8-17 APSR Position Limits, 0-200 +/- 10 EFPD (2.5,102) (40,102) 100 -

(2.5,92) 80 - (0,80) (40,80)

RESTRICTED OPERATION L.

U S60 60 . 60,60) o 0

0 0

S40ACCEPTABLE OPERATION (100,40)

S40-20 0- I I 0 20 40 60 80 100 APSR POSITION,%WD 8-18

Figure 8-18 APSR Position Limits, After 200 1 10 EFPD 100,02(7512 (40,102) 100 -

(2.5,92) 80 *(0,80) (40,80)

RESTRICTED OPERATION U

C 60 0) 0,0 T IC E L 40 -E AC EP AB ERPE AT ON(1 0

U ACCEPTABLE OPERATION

< 40- (100,40) 20 0 I 0 20 40 60 80 100 APSR POSITION,%WD 8-19

REFERENCES 1 Oconee Nuclear Station, Units 1, 2, and 3 Final Safety Analysis Report, Docket Nos. 50-269, 50-270, and 50-287.

2 A. F. J. Eckert, H. W. Wilson, and K. E. Yoon, Program to Determine In reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084P-A, Rev. 3 Babcock & Wilcox, October 1980.

3 BPRA Retainer Design Report, BAW-1496, Babcock.& Wilcox, May 1978.

4 TACO - Fuel Performance Analysis, BAW-10087A, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, August 1977.

5 Oconee Unit 3, Cycle 6 - Reload Report, BAW-1634, Babcock & Wilcox, August 1980.

6 B&W Version of PDQ07 Code, BAW-10117A, Babcock & Wilcox, January 1977.

7 Core Calculational Techniques and Procedures, BAW-10118, Babcock & Wilcox, October 1977.

8 Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Babcock & Wilcox, May 1977.

9 Oconee 3 Fuel Densification Report, BAW-1399, Babcock & Wilcox, November 1973.

10 L. S. Rubenstein (NRC) to J. H. Taylor (B&W) Letter, "Evaluation of Interim Procedure for Calculating DNBR Reductions Due to Rod Bow," October 18, 1979.

11 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 3, Babcock

& Wilcox, July, 1977.

12 J. H. Taylor (B&W) to S. A. Varga (NRC), Letter, July 18, 1978.

13 Power Peaking Nuclear Reliability Factors, BAW-10119, Babcock & Wilcox, January 1977.

A-1

14 Normal Operating Controls, BAW-10122, Babcock & Wilcox, August 1978.

15 Verification of the Three-Dimensional FLAME Code, BAW-10125A, Babcock &

Wilcox, August 1976.

16 Oconee Nuclear Station Reload Design Methodology Technical Report NFS-1001, Rev. 4, Duke Power Company, Charlotte, North Carolina, April 1979.

17 CASMO - A Fuel Assembly Burnup Program, AE-RF-76-4158, Studsvik Energiteknik AB, June 1978.

1 W. 0. Parker, Jr. (Duke) to H. R. Denton (NRC), Letter, October 16, 1981.

19 J. H. Taylor (B&W) to L. S. Rubenstein (NRC), Letter, September 5, 1980.

20 Oconee Unit 2, Cycle 6 - Reload Report, BAW-1691, Babcock & Wilcox, August 1981.

21 J. H. Taylor (B&W) to S. A. Varga (NRC), Letter, "BPRA Retainer Reinsertion," January 14, 1980.

A-2 8

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