ML19326B887
| ML19326B887 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/17/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19326B881 | List: |
| References | |
| NUDOCS 8004180615 | |
| Download: ML19326B887 (6) | |
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NUCLEAR REGULATORY CCMMtsslON
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENCMENT NO. 31 TO FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE - UNIT NO. 1 DOCKET NO. 50-313 1.0 Introduction By letter dated December 28,1978 (Reference 1), as supplemented by letters dated January 17, and 30,1978, and March 3,1978 (References 2, 3 and 4, respectively), the Arkansas Power and Light Company (AP3L or the licensee) requested an amendment to Facility Operating License No. OPR-51. The amendment would modify the Technical Specifications for Arkansas Nuclear One, Unit No.1 (ANO-1) for Cycle 3 cperation.
2.0 Evaluation The ANO-1 reactor core censists of 177 fueled assemblies, each containing a 15x15 array of fuel rods. Each 15x15 array contains 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube.
For Cycle 3 operations all Batch 2 assemblies will be discharged frpm the core. Five once-burned Batch 1 fuel assemblies will be reloaded into the center of the core. Sixty (60) Satch 3 4:remblies and 56 Batch 4 assemblies will be shuffled into new locatiers. Fifty-six (56)
Batch 5 fresh assemblies will occupy the core periphery and eight interior locations. Tables 4-1 and 4-2 of ?.eference 1 contain su=maries of fuel design parameters, dimensions and thermal analysis parameters f.r the fuel batches which will be burned in Cycle 3.
5 3 activity control will be supplied by 61 full length Az-in-Cd control rids and soluble boron shim. In addition, eight (8) partial length axial pcwer shaping reds (APSRs) are provided for control of the axial power distributions. Control red interchanges or burnable poison rods are unnecessary for Cycle 3 oceration.
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9 2.1 Fuel' Mechanical Desion The Batch 5 fresh fuel uses the Mark B-4 fuel assembly design which was initially used in Batch 4 during Cycle 2.
The reload fuel assemblies' incorporate minor changes in the spacer grid corner cells which reduce spacer grid interaction during handling. Additionally, dynamic impact testino has shown that the spacer grids have a higher seismic capabi.lity and consequently an increased safety margin over the values reported in. Reference 5.
The dynamic impact testing techniques are described in Reference 6.
Creep collapse time was calculated to be in excess of 30,000 effective full power hours (EFPH) whicn is longer than the projected three cycle exposure of 25,584 EFPH. The calculation of creep collapse time was performed using the power history of the limiting fuel assembly. As was done in Cycle 2, the CROV computer code was used to predict the collapse time (Reference 7). The licensee stated (Reference 3) that the CROV code demonstrated its ability to conser-vatively predict cladding collaose.
Additional conservatisms used an the CROV calculations were that no credit was taken for fission gas release; the cladding thickness used in CROV was the lower tolerance limit (LTL) of the as-built measurements; and the lowest as-fabricated pellet densities were assumed to be located in the worst case power region of the core.
The fuel clad strain analysis 'was perfomed using a number of conser-vative assumptions: maximum allowable fuel pellet diameter and density, lowest pemitted tolerance for the cladding inner diameter, conserva-tively high local pellet burnup, and conservatively high heat generation rate. This insures that the 1.0", limit on cladding plastic circumferen-tfal strain is not violated.
i The Batch 5 fuel assembly design is based upon established concepts and utilizes standard comoonent materials. Therefore, on the bases of the analyses presented and previous successful operations with equivalent fuel, we conclude that the fuel mechanical design for Cycle 3 operations-is acceptable and does not decrease the safety margin.
2.2 Fuel Thermal'Desien
- The Batch 5 fuel produces no significant differences in fuel thermal performance relative to the other fuel remaining in the core. As was-done in the Cycle 2 reload calculations, the linear heat rate (LHR) capability of ANO-1 was calculated using the TAFY cc: router code (Reference 9). The nominal LHR for Cycle 3 varies from a value of -5.77 for the Batch 1 fuel to 5.20 for tne Batch 5 fuel. The LHR capability varies frem 19.40 for Batch 3 to 20.15 for Satches a and 5.
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4 l?ut densification power spike model for Cycle 3 used the conservative combination of initial density and enrichment to calculate the spike factor. The power spike modcl is the same as that presented in Reference 10 with modifications to Fg and Fk. These changes reflect additional data from operating reactors which support a different approach and yield less severe penalties due to power spikes. Based on the analyses presented in Reference 1 and comparison with the allowable Linear Heat Generation Rate (LHGR) for fuel centerline melt considerations (Reference 11), the fuel thermal design for the ANO-1 Cycle 3 core is acceptable and does not decrease the safety margin.
2.3 Fuel Material Design Cycle 3 fuel for ANO-1 will not have any significant material changes from previous cycles. Batch 4 started the use of a Zircaloy-4 (Zy-4) spacer material rather than Zirconium dioxide (Zr02) material. The use of Zy-4 spacer material is continued in Batch 5 assemblies.
It was concluded in Reference 12 that the change from Zr02 to Zy-4 does not affect the primary coolant system chemistry. Therefore, the fuel material design for ANO.1 Cycle 3 operations is acceptable.
2.4 Nuclear Analysis Physics parameters were calculated for the AMO-1 Cycle 3 core. There are minor differences between Cycle 3 and the Cycle 2 reference cycle physics parameters since Cycle 3 is not yet an equilibrium cycle.
However, the differences in these parameters are minor.
The licensee requested a change in the ANO-1 Technical Specification regarding the correction of the hot zero power (HZP) measured moderator temperature coefficient (MTC) to compare with the 955 power Technical Specification limit (Reference 2). The proposed change would allow the use of cycle dependent parameters measured in the physics startup testing to project or extrapolate the 955 power value. The current Technical Specification requires a Technical Specification change each cycle because the cycle dependent corrections to the MTC at HZP are explicitly stated in the Technical Specification. We find that this approach will eliminate an unnecessary administrative step and -
is therefore acceptable.
The licensee also proposed a change in the plant Technical Specifica-tions increasing the allowable quadrant tilt from 3.4% to a.925. The additional peaking allowed is a result of the statistical combination of the nuclear uncertainty factor, the hot channel factor, and the rod bow peaking penalty. He find that this Technical Specification is acceptable and does not decrease the safety margin.
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. The only significant proposed operational procedure change is the
. proposed Technical Specification change of the axial power shaping rod (APSR) position limits. The APSR position limits would provide added control of power peaking to insure that peak power limits for Los: of Coolant Accident (LOCA) conditions would not be violated.
We find that, based on the AP&L's nuclear' analysis techniques and their comitmsnt to perform acceptable physics startuo testing, the ANO-1 nuclear analysis is acceptable. We also find the proposed Technical Specifications of APSR position limits and the usual regulating control rod and imbalance limits, which assure that the loss of coolant accident (LOCA) LHGR limits are not exceeded, are acceptable.
2.5 Thermal-Hydraulic Analyses The thermal-hydraulic analyses for ANO-1 Cycle 3 were perfomed using previously approved methods and models per the ANO-1 Final Safety Analysis Report (FSAR). The only change in the thermal-hydraulic analysis for Cycle 3 is the removal of the densification power spike from Departure from Nucleate Boiling Ratio (DNBR) calculation, resulting in an increase in the minimum calculated steady-state ONBR from 1.84 for Cycle 2 to 1.90 for Cycle 3.
The maximum fuel rod bow, calculated using the interim NRC fuel rod
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bow model, is 11.25 and occurs at the end of Cycle 3.
The licensee provides the requisite margin by the flux / flow trip setpoint of 1.060 and the variable low-pressure trip. We find that the thermal-hydraulic analysis for ANO-1 Cycle 3 operations is acceptable.
2.6 Accident and Transient Analysis The generic Babcock and Wilcox (B&W)' Loss of Coolant Accident (LOCA) analysis is contained in BAW-10103 (Reference 13). The analysis in BAW-10103 is generic since the limiting valt.s of key parameters for all plants in the category I (177 FA-lowered loop) Nuclear Steam Supply System (NSSS) are used. The combir'ttion of average fuel temperature
~ and pin pressure data, for the lift lime of the fuel, as used in the BAW-10103 LOCA limits analysis is conservative comoared to those used in the Cycle 3 reload analysis. In Reference 14, S&W submitted a change to the BAW-10103 LOCA analysis because of an incorrect pressure drop assumed for the inlet nozzle region. The correction incorporates a revised reactor coolant system pressure distribution.
The result is that the peak clad temperature in the revised calcula-tion is 20600F for the unruptured node and 18260F for the ruptured node. This is a reductim of 860F and 2400F, respectively, relative to the BAW-10103 results. Therefore, tne analysis presented in BAW-10103 is valid for the reload cycle.
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. Relative to plant transients, the Cycle 3 evaluation is bounded by the FSAR, the fuel densification report (Reference 15) and previous cycle analyses.
We conclude that the' LOCA analyses performed for ANO-1 meet 10 CFR 50.46 criteria and insure that the plant can be operated without undue risk to the public safety.
2.7. physics Startup Tests The proposed physics startup program is discussed in Reference 4.
The licensee has comitted to conduct physics startup tests to insure that the significant aspects of the ANO-1 Cycle 3 core would be within the acceptable criteria. These include control rod functional tests, scram times, control rod worth tests, temperature reactivity coefficient tests, and power distribution tests. The licensee has also committed to provide a report on these tests within 45 days after completion of the test program. The program has been reviewed and found acceptable.
3.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluint types or total amounts nor an increase in power level and will. tot result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 851.5(d)(4), that an j
environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the -
amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3)
J such activities will be conducted in compliance with the Comission's l
regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: March 17,1978
n Re'ferences 1.
Letter, Rueter (AP&L) to Davis (NRC), Docket No. 50-313, December 28, 1977, forwarding the Arkansas Nuclear One, Unit No.1, Cycle 3 Reload Report, BAW-1471.
2.
, Letter, Cavanaugh (AP&L) to Davis (NRC), Docket No. 50-313, January 17, 1978.
3.
Letter, Cavanaugh (AP&L) to Davis (NRC), Docket No. 50-313,.
January 30, 1978.
4.
Letter, Williams (AP&L) to Reid (NRC), Docket No. 50-313, March 3, 1978.
5.' BAW-10035, " Fuel Assembly Stress and Deflection Analysis for Loss-of-Coolant Accident and Seismic Excitation," June, 1970.
6.
BAW-10133, ' Mark-C Fuel Assembly Topical Report on LOCA - Seismic Analyses," October 1977.
7.
BAW-1433, " Arkansas Nuclear One, Un'it No.1 - Cycle 2 Reload Report," November,1976.
8.
BAW-10084P-A, " Program to Cetermine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse," January, 1975.
9.
BAW-1004, "TAFY - Fuel Pin Temperature and Gas Pressure Analysis,"
May,'1972.
- 10. BAW-10055, Rev.1, " Fuel Densification Peport, June,1973.
- 11. Standard Review Plan, Section 4.4, po. 4.4-2 and 4.4-3.
- 12. Letter, Davis (NRC) to Phillips (AP&L), Safety Evaluation for Amendment No. 21 to Facility Operating License No. DPR-51, March 31, 1977.
- 14. Letter, Taylor (B&W) to Baer (NRC), July 8,1977.
- 15. BAW-1391, " Arkansas Nuclear One, Unit Mo.1, Fuel Censification Reper*," June, 1973.