ML19312C500
| ML19312C500 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/21/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C496 | List: |
| References | |
| NUDOCS 7912160062 | |
| Download: ML19312C500 (9) | |
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NUCLEAR REGULATORY COMMISSION
,. t WASHINGTON. D. C. 20555
, /;"Q os, o j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 52 TO LICENSE NO. DPR-38 AMENDMENT NO. 52 TO LICENSE NO. DPR-47 AMENDMENT NO. 49 TO LICENSE NO. OPR-55 DUKE POWER C0f'PANY OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3 DOCKET NOS. 50-269, 50-270 AND 50-287 Introduction By letter dated September 6,1977(l) Duke Power Company (the licensee) requested caanges to the Technical Specifications appended to the Oconee Unit 3 operating License for Cycle 3 operation.
Evaluation The Oconee Unit 3 reactor core consists of 177 fuel assemblies. All of the Batch 2 fuel assemblies will be discharged at the end of Cycle 2.
Five oncegrned Batch 1 fuel assemblies, with an initial enrichment of 2.01 wt%
U, will be reloaded into the central portion of the core.
Batches 3, 4 and 4A with initial enrichments of 3.00, 2.53 and 2.64 wt%
235, respectively, will be shuffigg5 0
an initial enrichment of 3.02 wt%
u, will occupy primarily the core periphery _ and eight interior locations.
Fuel Assemoly Mechanical Desion The types of fuel assemblies and pertinent fuel design parameters and dimensions for Oconee 3, Cycle 3 are listed in Table 4-1 of the attachment to reference 1.
Batches 3, 4 and 4A fuel are essentially the same as Batch 1 fuel. The Mark 84 tresh fuel assemblies (Batch 5) incorporate minor design modifications to end fittings and spacer grid corner cells.
The latter change reduced spacer grid interaction during handling.
In addition, improved dynamic impact testing methods show that the spacgr grids have a higher seismic capability and thus increased safety margin.Il 7 912160 gf 2 L
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- The Batch 5,15 x 15 (Mark B-4), fuel assembly design and the Batch 1, 15 x 15 (Mark B-3), fuel assembly design have been previously reviewed and accepted by us for use in Oconee Unit 3.
Also, these types of fuel assemblies are currently operating in Oconee Unit 3.
The reload fuel assemblies, therefore, do not represent any unreviewed change in mechanical design from the reference cycle.
Each fuel assembly design has been taken into account in the various mechanical analyses. The Batch 3 fuel is generally limiting, because of its relatively low initial fuel pellet density, and previous incore exposure. The results of these analyses have shown that the mechanical design differences between fuels for Cycle 2 and Cycle 3 are negligible and are acceptaole.
Creep collapse analyses were performed for tnree-cycle fuel assembly power histories. Batches 3 and 4 were analyzed using as-built data. The Batch 3 fuel is more limiting for cladding collapse due to its previous incore exposure time. The creep collapse analyses were performed based on the conditions (JJsgt forth in reference 2 which have been previously found acceptable.
The collapse time for the most limiting assemnly was conservatively determined to be more than 30,000 EFPH (effective full-power hours), which is longer than the maximum design exposure for the total of three cycles.
The Oconee 3 stress parameters were enveloped by a conservative fuel rod stress analysis. The following conservatisms with respect to Oconee 3 fuel were used in the analysis:
lower post-densification internal pressure, lower initial pellet density, higher system prassure, and higher thermal gradient across the cladding.
For design evaluation, the primary stress must be less than two-thirds of the minimum specified unirradiated yield strength, and all stresses must be less than the minimum specified unirradiated yield strength.
In all cases, the margin is in excess of 307,.
The fuel design criteria specify a limit to the cladding plastic circum-ferential strain of 1.0"..
The pellet design is established for plastic cladding strain of less than l'. at maxinum design local pellet burnup and heat generation rate values, which are consideraoly higher than the values for Oconee 3 fuel. This will result in an even greater margin than the analysis demonstrated. The strain analysis is also based on the maximum manufacturing specifications values for the fuel pellet diameter and density and the icwest permitted manufacturing specifications tolerance for the cladding internal diameter.
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' The lirear heat generation rate (LHGR) capabilities are bgd on center-line fuel melt and were established using the TAFY-3 code with fuel densification to 96.5% of theoretical density. Two of the Batch 5 fuel assemblies were loaded with fuel pellets which have a different nominal density (91% TD) and diameter than the Batch 5 fuel assemblies. Based on these characteristics a LHGR of 19.74 KW/ft has been established for these two fuel assemblies. These assemblies will be placed in non-limiting locations during their entire core life. For Cycle 3, the licensee has specified that if these fuel assemolies and placed in locations M-14 and E-2, they will not experience LHGR's greater than 19.15 KW/ft.
All the other fuel assemblies in the Cycle 3 core are thermally similar.
The fresh Batch 5 fuel inserted for Cycle 3 operation introduces no sig-nificant difference in fuel thennal performance relative to the other fuel remaining in the core, and their LHGR limit has been established as 20.15 KW/ft.
The power spike model used for Cycle 3 analyses is the same as that used for Cycle 2.
The power spike factor and gap size were based on unieradi-ated Batch 4 and Batch 5 fuel (94.0% TO) with an assumed enrichment cf 3.0 wt% 2350 These values are conservatively high for Batch 1 and Batch 3 fuel.
The thermal analysis of the fuel rods assumed an in-reactor densification to 96.5% of theoretical den ty. The analytical methods used are the same as those for Cycle 2.(D These analyses were based on the lower tolerance limit of the fuel density specification and assumed isotropic diametral shrinkage and anisotropic axial shrinkage resulting from fuel densi fication.
The Batch 5 fuel assemblies are not new in concept, nor do they use different materials. Therefore, the chemical compatibility of all possible fuel-cladding-coolant assembly interactions for the Batch 5 fuel assemblies are identical to those of the present fuel.
This fuel as proposed for reload in Oconee 3 has had considerable operating experience. Because of this experience, the similarity of the Batches 1 and 5 fuel and because the fuel assemblies for Cycle 3 4
operation will not exceed ay design life limits, we conclude that operation is acceptable.gn and the fuel thennal design for Cycle 3 the fuel mechanical desi 4
. The core configuration for Cycle 3 differs slightly from that of Cycle 2 in that the Batch 2 fuel removed at the end of Cycle 2 is the Mark B-3 fuel assembly design, and the fresn Batch 5 fuel inserted for Cycle 3 is the Mark B-4 assembly design. Mark B-4 assemblies differ from the Mark B-3 primarily in the design of the end fitting, which results in a slight reduction in flow resistance for the B-4 design. No credit was taken in the analyses for the increased flow to the Mark B-4 assemblies, located in the hottest core locations, as a result of slight changes in the core flow distribution or for the increase in the system ficw resulting from the reduction in total ' ore pressure drop.
However, the slight reduction c
in flow rate of the Mark B-3 assemblies (because of the icwer flow resistance of the B-4 assemblies) was considered.
The BAW-2 CHF correlation (8) was used for thermal-hydraulic analysis of Cycle 3.
This correlation has been reviewed and approved for use with the Mark B fuel assembly design.(9)
The effect of fuel densification on minimum DNBR is primarily a result of the reduction in active fuel length, which increases the average heat flux. The Cycle 3 DNBR analysis was based on a cold densified active length of 140.2 inches, a value selected to apply gene-ically to a number of B&W plants.
This is a conservative method of appiying the densification effect since all the fuel assemolies in Cycle 3 have longer densified lengths and because no credit is taken for axial thermal expansion of the fuel column.
The potential effect of fuel rod bow on DNBR can be considered by incor-porating suitable margins into DNB-deminated core safety limits and reactor protection system setpoints. The maximum rod bow magnitude would be calculated from the equation ob = 11.5 + 0.069 M, where ab is the rod bcw magnitude in mils and BU is the burnup is MWD /mtu. The resultant DNBR penalty based on the maximum predicted assembly burnup at the end of Cycle 3 is approximately 6.0%.
However, since this rod bow model has not yet been found acceptable, the maximum rod bow magnitude was calculated using the NRC approved interim model, AC/Co= 0.065 + 0.001449 GI is the initial gap.
where AC is the rod bow magnitude (in mils) and Co The resultant DNBR penalty, based on the maximum predicted assembly burnup at end of Cycle 3, is 11.2%.
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Nuclear Analyses Table 5-1 of the attachment to reference 1 compares the core physics parameters of Cycles 2 and 3.
The values for both cycles were generateo using PD007. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. The shorter Cycle 2 produced a smaller cycle differential burnup than is expected for Cycle 3.
The accumulated average core burn-up will be higher in Cycle 3 than in Cycle 2 because of the presence of the or.ce-burned fuel of Batches 1, 3, 4 and 4A.
The critical boron concentrations for Cycle 3 are higher than for Cycle 2 because of a higher fuel enrichment in Cycle 3.
The control rod worths are sufficient to maintain the required shutdown margin for Cycle 3.
The maximum stuck rod worths for Cycle 3 are less than those in Cycle 2.
The adequacy of the shutdown margin with Cycle 3 rod worths has been demon-strated analytically. The shutdown calculations conservatively used a poison material depletion.llowance and 10% uncertainty on net rod worth.
The same calculational methods and design information were used to obtain the nuclear design parameters for Cycles 2 and 3.
In addition, for Cycle 3 there are no significant operational procedure changes from the reference cycle procedures with regard to axial or radial power shape control, xenon control or tilt control.
In view of the above and the fact that startup tests (to be conducted prior to power operation) will verify that the significant aspects of the core performance are within the assumptions of the safety analysis, we find the licensee's nuclear analyses for Cycle 3 to be acceptable.
Thermal-Hydraulic Analyses The major acceptance criteria which are used for the thermal-hyaraulic design are specified in Standard Review Plan (SRP) 4.4.
These criteria establish acceptable limits on departure from nucleate boiling (DNB). The thermal-hydraulic analyses for Oconee Unit 3 Cycle 3 reload were made using previously approved models and methods. Certain aspects of the thermal-hydraulic design are new for the Cycle 3 core and are discussed bel ow.
The thermal-hydraulic design evaluation in support of Cycle 3 operation used the methods and models described in references 5, 6 and 7.
Cycle 3 analyses and resulting setpoints have been based on 106.5% of the design reactor coolant (RC) system flow rate. Cycle 2 analyses used 107.6% of design flow based on a measured ficw value of 110.0%. The reduced flow rate has been selected for Cycle 3 analyses to provide consistency witn Oconee Units 1 and 2.
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/ The pressure-temperature limit curve shown in Fiqure 2.1-1C of the Technical Specifications provices the basis for the variable low-pressure trip setpoint. The curve shows for all modes of reactor operation locus of points for which the calculated minimum DNBR is equal to 1.30 (BAW-2) plus the margin required to offset an 11.2%
DNBR reduction due to rod bow.
The specific credits used in this analysis to account for rod bow are as follows:
% DNBR credit 10.2 Credit for rod bow penalty already
=
included in analysis 1.0 Credit for flowa.rea reducti3n
=
factor in analysis 4
none claimed Credit for plant excess flow (3.5%
=
available)
Total 11.2 The flux /ficw trip setpoint was determined by analyzing an assumed two-pump coastdown starting from an initial indicated power level of 102% plus flux medsurement and heat balance errors (equal to 108%
full power in core). The specific credits used in this analysis to account for rod bow are as follows:
" DNBR credit 5.8 Credit for rod bow penalty already
=
included in analysis 1.0 Credit for flow area reduction
=
factor in analysis Credit for 2% (3.5% available) 4.4
=
excess RC flow Total 11.2 l
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, In summary, a reactor coolant flow rate based on actual measured flow with uncertainties and conservatisms was used in the Oconee Unit 3 Cycle 3 thennal hydraulic analyses.
The licensee has also assured us that there will be sufficient margin in the reactor coolant flow rate (at least 108.5% of design) to compensate for the difference between the approved and the not yet approved rod bow.
model s.
Based on our review, we find that the licensee has included appropriate conservatisms in the analyses and that the proposed Technical Specifications provide assurance that the criteria of SRP 4.4 will be met. Therefore, we conclude that the thermal-hydraulic analyses are acceptable.
Accident and Transient Analyses The accident and transient analyses as provided by the licensee demonstrate that the Oconee FSAR analyses conservatively bound the predicted conditions of the Oconee Unit 3 Cycle 3 core and are, therefore acceptable. The licensee has stated that each FSAR accident analysis has been examined, with respect to changes in Cycle 3 parameters, to detemine the effects of the reload and to ensure that performance is not degraded during hypothetical transients. The core thennal parameters used in the FSAR accident analyses were design operating values based on calculated values plus uncertainties. FSAR values of core thennal parameters were compared with those calculated in the Cycle 3 analyses. For each accident of the FSAR, a discussion and comparison of the key parameters from the FSAR and Cycle 3 was provided by the licensee to show that the initial conditions of the transient are bounded by the FSAR analysis. The effects of fuel densification on the FSAR accident results have been evaluate Oconee Unit 3 fuel densification recort.p5;gnd are reported in the Since Cycle 3 reload fuel assemblies contain fuel rods with tneoretical density higher than tnose considered there, the conclusions derived in that report are valid for Oconee Unit 3 Cycle 3.
Calculational techniques and mathods for Cycle 3 analyses remain consistent with those used for the FSAR.
No new dose calculations were performed for the reload report. The dose considerations in the FSAR are based on maximum peaking and burnup for all core cycles; therefore, the dose con-siderations are independent of the reload batch.
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, A review of the ECCS U-baffle pressure drop error has been perfomed and documented in reference 10. The review considered a reanalysis of the reactor coolant system pressure loss characteristics and the effects and ECCS performance.
The review found the current ECCS perfomance analysis acceptable for all three Oconee units.
Reference 10 also found that a new surveillance testing program of the reactor internals vent valves is acceptable for all three Oconee units. The review considered the impact of these changes on ECCS perfomance and the adequacy of the surveillance techniques.
Startup Tests A startup program will be conducted to verify that the core performance is within the assumptions of the safety analyses and provide the necessary data for continued plant operation. The startup test program is similar to that previously approved for Cycle 2 operation. Within 90 days following completion of physics testing the licensee will pro-vide a summary of the test program results. This startup test program and reporting schedule are acceptable.
Environmental Consideration We have determined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that these amendments involve an action which is insignificant frcm the standpoint of environmental impact and pursuant to 10 CFR [51.5(d)(4) that an environmental impact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) tnere is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date: November 21, 1977 l
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l REFERENCES l
1.
Letter from W. O. Parker, Jr., (Duke Power Cor.pany) to Edson G. Case, (flRC) dated September 6, 1977.
2.
Program to Ceternine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAH-10034, Rev.1, Babcock & Wilcox, November 1976.
3.
Letter from A. Schwencer (NRC) to J. F. Mallary (B&W) dated Jan. 29, 1975 4.
C. D. Morgan and H. S. Xao, TAFY - Fuel Pin Temoerature and Gas Pressure Analysis, 2AW-100aa, Babcock & Wilcox, May 1972.
5.
Oconee 3 Fuel Densification Recort, BAW-1399, Babcock & Wilcox November 1973.
6.
B. J. Buescher and J. W. Pergram, Babcock & Wilcox Model for Predicting In-Reartor Densification, BAW-10083P, Rev.1, Babcock
& Wilcox, November 1976.
7.
Oconee Nuclear Station, Units 1, 2, and 3, Final Safety Analysis Report, Docket Nos. 50-269, 50-270, and 50-287.
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8.
Correlation of Critical Heat Flux in a Bundle Cooled by Pressuri:ed Water, BAW-10000A, Babcock & Wilcox, June 1976.
9.
Letter from J. Stolz (NRC) to K. E. Surke (B&W), dated April 15, 1975.
10.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment,Nos. 45, a5 and 42 to Facility License Nos.
OPR-38, OPR-47 and CPR-55, Duke Power Ccmoany, Oconee Nuclear Station Uni: Nos. 1, 2 and 3, July 29, 1977.
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UNITED STATES NUCLEAR REGULATORY CCMMISSION DOCKET NOS. 50-269, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENCMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Comission) has issued Amendment Nos.
52,
52 and 49 to Facility Operating Licenses Nos. OPR-38, DPP-47 and OPR-55, respectively, issued to Duke Pcwer Company which revised Technical Specifications for operation of the Oconee Nuclear Station Unit Nos.1, 2 and 3, located in Oconee County, South Carolina. The amendments are effective as of their date of issuance.
The amendments revise the Technical Specifications to establish operating limits for Unit 3 Cycle 3 operation.
The application for these amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Ccamission's rules and regulations. The Ccmission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments.
Prior public notice of these amendments was not required since the amendments do not
'volve a significant hazards consideration.
The Cemission has detemined that the issuance of these amendments will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an envirormental impact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with issuance of these amendments.
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,... For further details with respect to this action, see (1) the application for amendments dated September 6,1977,(2) Amendment Nos.
52 52 and 49 to Licenses Nos. DPR-38, DPR-47 and CPR-55, respectively, and (3) the Comission's related Safety Evaluation. All of these items are available for public inspection at the Comission's Public Document Room,1717 H Street, N.W., Washington, D.C.
20555 and ac the Oconee County Library, 201 South Spring, Walhalla, South Carolina 29691. A copy of items (2) and (3) may be obtained upon request addressed to the U.'S. Nuclear Regulatory Comission, Washington, D.C.
20555, Attention:
Director, Division of Operating Reactors.
Dated at Bethesda, Maryland, this 21st day of Neverter 1977.
FOR THE NyCLEAR REGULATORY COMMISSION
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'A.Ic'hwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors L
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