ML20062L348
| ML20062L348 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/17/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20062L347 | List: |
| References | |
| NUDOCS 9312300006 | |
| Download: ML20062L348 (7) | |
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UNITED STATES li NUCLEAR REGULATORY COMMISSION g
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wasnincton, o.c. 2oswoooi SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO.112TO FACILITY OPERATING LICENSE NPF-3_5 AL4D AMENDMENT N0.106 TO FACILITY OPERATING LICENSE NPf-52 DUKE POWER COMPANY. ET AL.
I CATAWBA NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414
1.0 INTRODUCTION
By letter dated September 7,1993, Duke Power Company, et al. (the licer ae),
submitted a request for changes to the Catawba Nuclear Station, Units 1 ad 2, Technical Specifications (TS). The changes are required by the reloading of Unit I with Mark-BW fuel for operation in Cycle 8.
Duplicate TS pages have been created to preserve certain TS values unchanged for Unit 2.
As discussed with the licensee, the Table of Contents'has also been revised to reflect the ;
new page numbering sequence.
The requested changes would revise the TS to (a) reduce the slope of the axial power imbalance per,alty in the overtemperature-delta temperature reactor protection system trip setpoint equation, and (b) increase the boron concentration limits in the cold leg accumulators (CLA), the refueling water storage tank (RWST), and the reactor coolant system and refueling canal during MODE 6 conditions.
The changes in boron concentrations are required to offset the additional reactivity needed for an increase in cycle length from 350 effective full power days (EFPD) to 390 EFPD and to offset the increased l
positive reactivity inserted following the cooldown of a core with a higher percentage of Mark-BW fuel.
2.0 EVALVATION The Catawba Unit 1 plant recently completed operating in Cycle 7 with a core that was more than 2/3 loaded with B&W Fuel Company (BWFC) Mark-BW 17x17 fuel.
The staff's evaluation of TS changes _ made for that cycle of operation was reported in Reference 1.
The Catawba Unit 1 Cycle 8 (CIC8) core will consist of 193 fuel assemblies.
Seventy-six of these are fresh assemblies manufactured by the BWFC; 117 are Mark-BW assemblies that have completed one or two cycles; and 9 are Westinghouse OFA assemblies that have completed 3 cycles.
The design cycle length of CIC8 is increased to 390 EFPD from the previous cycle's 350 EFPD. The use of the Mark-BW fuel design in Catawba has been previously approved as summarized in Reference 1.
)
There are no significant changes in the fuel assembly design, the fuel rod design, the thermal or material design of the Cycle 8 fuel from that previously described in Reference 1 for Cycle 7.
The licensee states that 9312300006 931217 PDR ADOCK 05000413 P
(a)the replacement Batch 10 of park-BW fuel assemblies will have an enrichment 2
of 3.65 percent by weight of U (b) calculated cladding creep collapse time is predicted to be greater than fuel residence time, (c) cladding stress is predicted to be within ASME limits, and (d) cladding strain is predicted to be below the one percent strain limit. The thermal design analysis indicates that the maximum predicted fuel rod burnup values in Cycle 8 will be less than the cladding collapse burnup values.
The limiting fuel rod internal pressure has been found to be less than reactor coolant system pressure.
The licensee states that these parameters have been analyzed with approved methodology (References 2, 3, 4 and 5). The results of these analyses are consistent with previously established limits and were performed by DPC with the methodology described in approved topical reports, and therefore, are acceptable.
2.1 Fuel System Desian Epclear Desian The core physics parameters for Cycle 8 were generated similarly to those for Cycle 7, using the PDQ-7 and E'31-N0DE-P computer codes and methodology that l
has previously been approved by the NRC staff (report DPC-NF-2010-A, Reference 6 and DPC-NE-3001-PA, Reference 12).
The physics analysis described in DPC-NF-2010-A is intended to determine the values of safety related parameters including those describing the core power distribution, reactivity worths and coefficients, and the reactor kinetics characteristics.
The DPC-NE-3001-PA report describes the methodology used by the licensee to ensure that the accident analysis for a defined reference core conservatively bounds the reload core.
The important key safety parameters for each FSAR Chapter 15 event are identified, and the methods for calculating these parameters are described.
In reload aplications, the licensee shows that the reference i
analysis described in the report is bounding by demonstrating that the event-specific key safety parameters of the reload core are within the conservative envelope of the reference analysis.
Also, as for Cycle 7, the' reactor protection limits and core operational limits were analyzed with approved methodology (Reference 7).
Control Reauirement The value of the required shutdown margin occurs at the end-of-cycle and at hot zero power conditions. Sufficient net available control rod worth, including a maximum worth stuck rod and appropriate calculation uncertainties, exist to meet shutdown margin requirements.
These results were developed usipg approved methods and incorporated appropriate assumptions and are, therefore, acceptable.
Thermal-Hydraulic Desian 1
The thermal performance of Cycle 8 fuel was analyzed using NRC-approved methodology (Reference 8).
The analysis methodology is consistent with that for Cycle 7 analyses in that they are based on the BWCMV departure from nucleate boiling (DNB) correlation (Reference 9) with a generic statistical DNB ratio limit of 1.40.
A margin is added to account for uncertainties,
including the transition from Westinghouse 0FA fuel to Mark BW fuel, to arrive at a design DNBR value of 1.55.
Accident Analysis The licensee has evaluated the following listed anticipated operational occurrences and postulated accidents:
increase in feedwater flow, excessive load increases, steam system piping failures, turbine trip, feedwater system pipe break, partial loss of forced reactor coolant flow, complete loss of forced reactor coolant flow, reactor coolant pump shaft seizure, uncontrolled rod bank withdrawal at power, dropped rod / rod bank, statically misaligned rod, single rod withdrawal, startup of an inactive reactor coolant put.p, boron dilution, rod ejection, steam generator tube failure, and loss-of-coolant accidents (LOCA). With the exception of the steamline break event and the post-LOCA subtriticality analyses, the licensee found the C108 thermal-hydraulic and physics parameters to be bounded by the existing Final Safety Analysis Report (FSAR) Chapter 15 analyses. The licensee's analysis of the steam line break event with the more positive CIC8 boron worth (B0C boron worth changes from -7.81 to -7.28 pcm/ppmb and E0C boron worth changes from -
8.91 to -8.31 pcm/ppmb from CIC7 to C108) showed the existing limiting case to be unchanged and to require no changes to the TS. The post-LOCA subcriticality analysis required an increase in RWST concentration from 2000 to 2175 ppm and an increase in CLA minimum concentration from 1900 to 2000 ppm.
The maximum RWST and CLA limits, are accordingly, increased to preserve an operating band.
These changes are reflected in TS 3.1.2.5, 3.1.2.6, 3.5.1, 3.5.4, and 3.9.1.
In addition, the increases in RWST and CLA maximum boron concentration limits necessitated a reanalysis of the post-LOCA boron precipitation evaluation and of the post-LOCA containment sump pH. The post-LOCA boron precipitation analysis requires a reduction in the time that the operator must initiate recirculation through the hot leg from 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. The post-LOCA sump pH analysis indicated that the existing range in the TS BASES is acceptable.
The licensee has also found that the slope and breakpoint of the over-temperature delta-temperature reactor trip function, as specified in TS Table 2.2-1, may be changed to remove some conservatism in the CIC7 value for Unit 1.
These analyses were performed using methodology as described in the topical reports listed as references in the licensee's application. These topical reports have been reviewed previously by the NRC and have been found acceptable as stated in the safety evaluations for those reports.
The staff's review of the CIC8 reload parameters found them to be bounded by the accident analysis assumptions stated by the licensee, and are therefore acceptable.
2.2 Technical Soecification Chances (1) Table 2.2-1 Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, Note 1(iii),
is revised for Unit I such that for each percent Al that the power imbalance
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-q is more positive than +3%, the oT trip setpoint shall be reduced by q*525f versus the previous value of 2.316%.
1.
A duplicate of this TS page is created as applicable to Unit 2 to retain the current value of 2.316% for Unit 2.
This is acceptable.
(2) TS 3.1.2.5, 3.1.2.6, and 3.5.1 Catawba Unit 1 Cycle 8 (C108) is characterized by an increase in the length of I
the fuel cycle from 350 EFPD to 390 EFPD as the transition is made to a higher percentage of Mark BW fuel. The licensee's analyses show that higher minimum boron concentrations in the RWST and the CLAs are needed to offset the additional reactivity needed to accomplish the longer cycle length and to offset the increased positive reactivity insertion, that would accompany the l
post-LOCA cooldown of a core with a higher percentage of Mark BW fuel.
The increased positive reactivity insertion derives from a generally more negative moderator temperature coefficient associated with the Mark BW fuel's smaller water to uranium ratio.
Associated with an increase in the minimum boron concentration is an increase in the maximum concentration since an allowable operating space range of concentrations must be preserved.
The increase in the minimum concentration is reflected in TS 3.1.2.5,3.1.2.6l and 3.5.4 for the RWST (2000 ppm to 2175 ppm) and in TS 3.5.1 for the CLA (1900 ppm to 2000 ppm).
The increase in the maximum concentration is reflected in TS 3.5.1 (2100 ppm to 2275 ppm) for the CLA and TS 3.5.4 for the RWST (2100 ppm to 2275 ppm).
The CLA volume weighted average boron concentration range in TS 3.5.1 is increased from 1800 ppm - 1900 ppm to 1900 ppm - 2000 ppm for the same reasons as cited above.
Also, for TS 3.9.1, the required reactor coolant system (RCS) and refueling canal minimum boron concentration during MODE 6 conditions has been revised i
from 2000 ppm to 2175 ppm.
This maintains consistency in the RCS and RWST concentrations during MODE 6 conditions and is acceptable.
lhe changes in the minimum boron concentrations require consideration of (a) adequate shutdown margin during all modes of norman operation, and (b) post-LOCA subtriticality.
Boron concentration changes may also affect the post-LOCA recirculated coolant pH analysis. As noted earlier, DPC's analysis found that previous results remained bounding and no TS changes were required.
The. licensee's submittal of January 13, 1993 (Reference 17), which has been incorporated by reference by the licensee, references an approved topical i
report, DPC-NF-2010 A, Section 4.0, as providing the methodology for normal operation shutdown margin determination.
The minimum required shutdown margin values,1.3% A0 for Taya 2 200 *F and 1.0% 40 for T
< 200 *F) are included as Limiting Conditions for Operation in TS 3.1.1.1,y,d 3.1.1.2.
The an calculated value for CIC8 is reported as 1.367%op in Table 5-2 of the application.
Since it was determined in accordance with the approved DPC Topical Report DPC-NF-2010 and is otherwise in accordance with TS 3.1.1, it is acceptable.
I I
9 The licensee's submittal of January 13, 1993, references the post-LOCA subcriticality analysis methodology as provided in FSAR Section 15.6.5.2.
i Also, the required all-rods-out (AR0) critical boron concentration, which must be bounded by the calculated boron concentration during the sump recirculation mode, is determined using approved methodology in DPC-NF-2010 A, Section 9..
Since the principal parameter in this determination, the limiting ARO critical boron concentration, has been determined using approved methodology we conclude that the post-LOCA subcriticality analysis is acceptable.
l The change in the maximum boron cor. centration requires assurance that boron precipitation is precluded following a LOCA. As stated in the licensee's application of September 7,1993, post-LOCA boron precipitation would be prevented with a reduction in the hot leg recirculation initiation time from 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. This is acceptable.
3.0 STATE CONSULTATION
r In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (58 FR 57847 dated October 27,1993).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common i
defense and security or to the health and safety of the public.
Principal Contributors:
R. E. Martin, PD II-3 S. L. Wu, SRXB Date:
December 17, 1993
[
t References 1.
Letter, R. E. Martin, NRC, to M. S. Tuckman, DPC, forwarding amendments numbered 101 and 95 for Catawba Units 1 and 2, respectively, dated-September 14, 1992.
2.
BAW-10172-PA, Mark-BW Mechanical Design Report, Babcock & Wilcox (B&W)
Topical Report, December 19, 1989.
3.
DPC-NE-2001-PA, Rev.1, Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel, Duke Power Ompany (DPC) Topical Report, October 1990.
4.
BAW-19984-A, Rev. 2, Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, B&W Topical Report, October 1978.
5.
BAW-10141-PA, Rev.1, TAC 02 - Fuel Performance Analysis, B&W Topical Report, June 1983.
6.
DPC-NF-2010-A, McGuire Nuclear Station / Catawba Nuclear Station Nuclear Physics Methodology for Reload Design, DPC Topical Report, June 1985.
7.
DPC-NE-20ll-PA, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, DPC Topical Report, March 1990.
8.
DPC-NE-2004-PA, McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Mathodology using VIPRE-01, DPC Topical Report, December 1991.
9.
BAW-10159-PA, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid fuel Assemblies, B&W Topical Report, July 1990.
10.
BAW-10173-PA, Rev. 2, Mark-BW Reload Safety Analysis for Catawba and McGuire, B&W Topical Report, February 20, 1991.
- 11. DPC-NE-3000P, Rev.1, Thermal-Hydraulic Transient Analysis Methodology, DPC Topical Report, May 1989.
12.
DPC-NE 'l001-PA, Rev.1, Multidimensional Reactor Transients and Safety Analysi Physics Parameters Methodology, DPC Topical Rep 9rt,' November 1991.
)
13.
BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and McGuire
. Units, B&W Topical Report, May 1991.
14.
BAW-10168-A, B&W Loss-of-Coolant Accident Evaluation Model For Recirculating Steam Generator Plants, B&W Topical Report, January 1991.
- 15. DPC-NE-1003-A, Rev.1, McGuire Nuclear Station / Catawba Nuclear Station Rod Swap Methodology Report for Startup Physics Testing, DPC Topical Report, December 1986.
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- 16. DPC-NE-3002-A, McGuire Nuclear Station / Catawba Nuclear Station FSAR Chapter 15 System Transient Analysis Methodology, DPC Topical Report, November 1991.
17.
Letters, M. S. Tuckman, DPC, to NRC, Catawba Nuclear Station, Docket Nos.
50-413 and 50-414, McGuire Nuclear Station, Docket Nos. 50-369 and 50-370, Technical Specification Amendment, Relocation of Cycle-Specific Parameter Limits, dated January 13, and April-26, 1993.
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