ML20132E290
| ML20132E290 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/16/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20132E265 | List: |
| References | |
| NUDOCS 8508010762 | |
| Download: ML20132E290 (11) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 77 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA POWER CORPORATION, ET AL.
CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302
1.0 INTRODUCTION
In a letter dated April 25,1985 (Ref.1), Florida Power Corporation (the licensee) made application to modify the Technical Specifications for Crystal River Unit 3 to permit operation for a sixth cycle. The safety analyses performed and the resulting modifications to the plant Technical Specifications are described in the Cycle 6 reload report (Ref. 2).
The safety analysis for the previous fifth cycle of operation at Crystal River Unit 3 is being used by the licensee as a reference for the proposed sixth
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cycle of operation. Where conditions are identified as limiting in the fifth cycle analysis, our previous evaluation (Ref. 6) of that cycle continues to apply.
1.1 Description of the Cycle 6 Core The Crystal River Unit 3 core consists of 177 fuel assemblies, each of which is a 15 x 15 array containing 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. Cycle 6 will operate in a feed-and-bleed mode with core reactivity control supplied mainly by soluble boron in the reactor coolant and supplemented by 60 full length control rod assemblies (CRAs) and 44 burnable poison rod assemblies (BPRAs).
In addition, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The licensed core full power level is 2544 MWt.
1.2 Significant Areas of Review for this Reload For the most part, Cycle 6 of Crystal River Unit 3 will be identical in 0508010762 850716 ADOCK 050 g 2 DR
6 operation to Cycle 5, and most Technical Specification changes such as reactor core safety limit, trip setpoints, rod insertion limits and imbalance limits are the result of the changes associated with the insertion of new fuel, cycle lifetime and the time of withdrawal of APSR(s) which are often made in Babcock
& Wilcox (B&W) reactors. However, there were two significant changes associated with this reload. First, the APSRs were changed from using Ag-In-Cd to Inconel for neutron absorption. This change affects both the i
fuel design and nuclear performance of the core which are evaluated in Sections 2 and 3 of this SE. Second, the use of crossflow models which can predict flow redistribution effects in an open lattice reactor core are used to determine Departure from Nucleate Boiling Ratio (DNBR) margins. This change affects the thermal-hydraulic design of the core which is evaluated in Section 4 of this SE.
2.0 EVALUATION OF THE FUEL SYSTEM DESIGN 2.1 Fuel Assembly Mechanical Design The 60 B&W Mark-8415 x 15 fuel assemblies loaded as Batch 7 are mechanically interchangeable with Batches 6A, 68, 7A, and 7B' fuel assemblies loaded previously at Crystal River Unit 3.
The Mark-B4 fuel assembly has been previously approved (Ref. 3) by the NRC staff and is utilized in other B&W nuclear steam supply systems. A comparison of the fuel design parameters for the various fuel assemblies in Cycles 5 and 6 can be found in Table 4-1 of Reference 2.
2.2 Fuel Rod Design Although all batches in the Crystal River Unit 3 Cycle 6 core will utilize the same Mark-84 fuel design and are mechanically interchangeable, the Batch 8 assemblies will incorporate a slightly higher average enrichment. The 60 assemblies will contain 3.49 w/o U-235. The cladding stress, strain and collapse analyses for the Cycle 6 fuel rod design are bounded by conditions previously analyzed for Crystal River Unit 3 using methods and limits previously reviewed and approved by the NRC. We find that no further review
. of these areas is necessary.
2.3 Fuel Thermal Design All fuel in the Cycle 6 core is thermally similar. The fresh Batch 8 fuel inserted for Cycle 6 operation introduces no significant differences in fuel thermal perfomance relative to the fuel remaining in the core. The themal analyses for all fuel were pe' fomed with the TACO 2 code. Nominal undensified input parameters used in this methodology are provided in Table 4-1 (Ref. 2).
Densification effects are accounted for in the TACO 2 code densification model.
Linear heat rate (LHR) to fuel melt capability for all fuel was detemined with the TACO 2 fuel pin performance code. The analysis perfomed for Cycle 6 demonstrates that 20.5 kW/ft is a conservative limit to preclude centerline fuel melt (CFM) for all fuel batches.
The maximum fuel rod burnup at the end of cycle (EOC) 6 is predicted to be less than 40,500 mwd /mtu. Fuel rod internal pressure has been evaluated with TACO 2 for the highest burnup fuel rod and is predicted to be less than the nominal reactor coolant system pressure of 2200 psia.
All fuel thermal design analyses have been perfomed using TACO 2 which has been previously reviewed and approved by the NRC staff (Ref. 4) and is therefore acceptable.
2.4 Gray APSR Design 1
Tne gray APSRs that are to be used in Cycle 6 were designed to improve creep life. Cladding thickness and red ovality control, which are the primary factors controlling the creep life of a stainless steel material, have been improved to extend the creep life of the gray APSR. The minimum design cladding thickness of the Mark-8 APSR is 18 mils, while that of th? gray APSR is 24 mils. Additionally, the gap width between the end plug and the Inconel
- absorber material was reduced. Finally, the ovality in the gap area will also be controlled to tighter tolerances. The gray APSR design was analyzed by B&W to demonstrate that it meets specified design requirements. The APSR was analyzed for cladding stress due to pressure, temperature, and ovality.
It was found that the gray APSR has sufficient cladding and weld stress margins. The gray APSR was also analyzed for cladding strain due to thermal and irradiation swelling. The results of B&W analysis showed that no cladding strain is induced due to thermal expansion or irradiation swelling of the Inconel absorber.
We have reviewed the mechanical design of the gray APSRs as presented by B&W in Reference 2 and find it acceptable.
2.5 Operating Experience B&W has accumulated operating experience with the Mark-B 15 x 15 fuel assembly at all of the eight operating B&W 177-fuel assembly plants. A sumary of this operating experience as of October 31, 1984, is given on page 4-4 of Reference 2.
3.0 EVALUATION OF THE NUCLEAR DESIGN Table 5-1 (Ref. 2) compares the core physics parameters for Cycles 5 and 6 designs. The cycle burnup (80C to EOC) will be smaller for Cycle 6 than for Cycle 5 because of the shorter Cycle 6 length. Differences in cycle length, feed batch size and enrichment, BPRA loading, shuffle pattern, and rod group designations for Cycle 6 account for the differences in the physics parameters from those of Cycle 5.
The critical baron concentrations for Cycles 5 and 6 are given in Table 5-1.
The control rod worths differ between cycles due to the gray APSRs and changes in radial flux and burnup distributions. Calculated ejected rod worths and their adherence to criteria are considered by B&W at all times in life and at all power levels in the development of the rod position limits presented in Section 8 of Reference 2.
The maximum stuck rod worths are less at BOC 6 and
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, 4 greater at EOC 6 than those for Cycle 5.
The adequacy of the shutdown margin with Cycle 6 stuck rod worths is demonstrated in Table 5-2 of Reference 2.
The following conservatisms are applied to B&W's shutdown calculations:
A 1.
Poison material depletion allowance.
2.
Ten percent uncertainty on net rod worth.
3.
Flux redistribution penalty.
4 Flux redistribution was accounted for since the shutdown analysis was calculated using a two-dimensional model. The shutdown calculation at the end l
of Cycle 6 was analyzed at 400 EFPD and E0C. The latest time (210 EFPD) in core life at which the APSRs are inserted will be 400 EFPD.
To support Cycle 6 operation of Crystal River Unit 3, the licensee has provided analyses (Ref. 2) using analytical techniques and design bases established in B&W reports that have been approved by the NRC staff. The licensee has provided a comparison of the core physics parameters for Cycles 5 and 6 as calculated with these techniques. We find the predicted characteristics acceptable because they use approved techniques, the validity of which has been reinforced through a number of cycles of predictions for this and other reactors. As a result of our review of the characteristics j
compared to previous cycles, we agree with their use in the Cycle 6 accident and transient analysis, as discussed in Section 6 of this Safety Evaluation.
4.0 EVALUATION OF THERMAL-HYDRAULIC DESIGN j
The objective of the thermal-hydraulic review is to confirm that the design of the reload core has been accomplished using acceptable methods, and that acceptable safety margin is available from conditions which would lead to fuel damage during normal operation and anticipated transients.
4 Except for the steady state analysis codes discussed below, the therwel-hydraulic models and methodology used for Cycle 6 are the same as used for Cycle 5.
The effect of rod bow on DNBR was accounted for using the 4
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analysis presented in Reference 5 which has been reviewed and approved by the NRC staff.
The important thermal-hydraulic parameters are very similar with the exception of the design axial peak (1.50-e 1,65) for Cycles 5 and 6 as summarized in i
Table 1.
However, the principal steady state thermal-hydraulic analysis codes were charged from CHATA. TEMP to LYNXT, LYNX 1, and LYNX 2.
LYNX 1 and LYNX 2 tave been previrusly reviewed and cpproved for this application by the NRC 1
staff. The LYNXT computer code is currently under NRC staff review. However, i
this review has progressed sufficiently to approve its use in the steady state mode for Cycle 6 of Crystal River Unit 3.
LYNXT was previously approved by the staff for the steady state analysis of Arkansas Unit 1 Cycle 7 The use of LYNXT in the transient mode is currently being questioned by the staff and i
j LYNXT may not be approved for such application. This is not a problem for Crystal River Cycle 6 since the previously reviewed and approved code RADAR i
was used to perform the transient thermal-hydraulic analysis of the core, i
The minimum DNBR at the design overpower condition is equal to 2.07. Although the design axial peak has been increased from 1.50 > 1.65, the benefits of crossflow analyses have resulted in additional DNBR margins relative to Cycle 5.
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TABLE 1
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THERMAL-HYDRAULIC DESIGN ANALYSIS CYCLE 5 CYCLE 6 RATED POWER, MWt 2544 2544 DESIGN POWER, MWt 2568 2568 s
REACTOR COOLANT FLOW.,GPM 374880 374880 EFFECTIVE FLOW FOR HEAT TRANSFER, %
91.9 90.9/
REFERENCE DESIGN FAH 1.71 1.71
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REFERENCE DESIGN AXIAL POWER SHAPE 1.50 1.65 g-
' COSINE COSINE
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CHF CORRELATION BEW-2 B&W-2 DESIGN DNBR LIMIT 1.30 1.30 t\\
PRINCIPAL T-H ANALYSIS CODES
- CHATA, LYNXT, TEMP LYNX 1, LYNX 2 MINIMUM DNBR 9 112% OVER POWER 2.05 2.07 MINIMUM DNBR 9 CORE PROTECTION 31.4 1.6 5
I SAFETY LIMIT!
' 1.7 1.9 TRANSIENT ANALYSIS CODE RADAR RADAR l
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Based on the similarities between the major thermal-hydraulic parameters of Cycles 5 and 6 and use of approved and/or acceptable (LYNXT) analysis methods, we find the thermal hydraulic performance of Cycle 5 to be acceptable.
5.0 TECHNICAL SPECIFICATIONS As indicated in our review, the operating characteristics for Cycle 6 were calculated with well-established, approved and/or acceptable methods.
In addition, we agreed in Section 3 with the licensee's evaluation of control rod worths and their role in the establishment of control rod position limits.
The Technical Specification changes proposed in References 1 and 2 are a reflection of these analyses and are therefore acceptable.
The licensee's submittal proposes changes to support operating Cycle 6 for Crystal River Unit 3.
These changes include:
1,.
Reactor core safety limits and trip setpoints for reactor thermal power i
and axial power imbalance.
2.
Minimum boric acid and borated water volumes.
3.
Regulating and axial power shaping rod group insertion limits.
4,.
Axial power imbalance limits.
A 6.0 EVALUATION OF ACCIDENT AND TRANSIENT ANALYSIS e
\\ The licensee has examined each FSAR accident analysis with respect to changes in Cycip 6 parameters to determine their effect on the plant thermal performa'rre during hypothetical transients. The key parameters having the i
greatest effect on determining the outcome of a transient or accident are the i
core thermal parameters, thermal-hydraulic parameters, and physics and kinetics parameters. Core thermal properties used in the FSAR accident analysis were design operating values based on calculational values plus uncertainties.
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-g-Table 1 compares the thermal-hydraulic parameters for Cycles 5 and T.
These parameters are the same for both cycles. A comparison of the key kinetics parameters from the FSAR and Cycle 6 is provided in Table 7-1 of Reference 2.
These comparisons indicate no significant changes or changes in the conservative direction, except for the initial conditions for the four-pump coast-down and locked-rotor accidents. However, the B&W analysis in Reference 2 finds that the locked-rotor accident evaluated at 102% of 2568 MWt for Cycle 3 operation remains valid for Cycle 6.
Also, the Cycle 4 four-pump coastdown analysis, performed with an initial power level of 102% of 2544 MWt and a pump monitor delay time of 1.5 seconds, bounds Cycle 6 and remains valid. In addition, the single reactor coolant pump coastdown analysis accounted for equipment errors and delay times associated with the Rosemont flow transmitters, which replace the BY transmitters used in previous cycles.
These analyses have been previously reviewed and approved by the NRC staff and are therefore acceptable.
Generic LOCA analyses for the B&W 177-fuel assembly (FA) lowered-loop NSSS have been performed using the final acceptance criteria emergency core cooling systems (ECCS) evaluation model. These analyses used the limiting values of key parameters for all plants in the 177-FA lowered loop category, and therefore, are bounding for Crystal River Cycle 6 operation. Further details on plant-specific aspects of these analyses are discussed in Section 7.2 of Reference 2.
A comparison of the radiological doses calculated for Cycle 6 to those previously reported in the FSAR shows that all Cycle 5 dose values are either bounded by FSAR values or are a small fraction (10%) of the 10 CFR 100 limits.
Thus the radiological impact of accidents during Cycle 6 are not significantly different from those described in Chapter 14 of the FSAR.
7.0 EVALUATION FINDINGS We conclude from the examination of Cycle 6 core thermal and kinetic properties, with respect to acceptable previous cycle values and with respect to the FSAR values, that this core reload will not adversely affect the ability of Crystal River Unit No. 3 to operate safely during Cycle 6.
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8.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
We have detennined that the amendment involves 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 Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical j
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 this amendment.
9.0 CONCLUSION
We have 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, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Dated: July 16,1985 Principal contributors:
G. Schwenk
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REFERENCES 1.
Letter from G. R. Westafer (Florida Power) to H. R. Denton (NRC),
" Technical Specification Change Request No.135," dated April 25, 1985.
2.
" Crystal River Unit 3 Reload Report," 88W Company Report BAW-1860, dated April 1985.
3.
J. F. Stolz (NRC) letter to J. A. Hancock (Florida Power) transmitting Amendment 48 to Facility Operating License No. DPR-72, dated December 4, 1981.
4.
C. O. Thomas (NRC) letter to J. H. Taylor (Babcock & Wilcox) transmitting
" Safety Evaluation of BAW-10141," dated April 13, 1983.
5.
J. C. Maxler, et al., " Fuel Rod Bowing in B&W Fuel Designs, Revision 1,"
Babcock & Wilcox, Lynchburg, Virginia, BAW-10147-P-A.
6.
J. F. Stolz (NRC) letter to W. S. Wilgus (FPC) transmitting Amendment No.
64 with Safety Evaluation of Crystal River Unit 3 Cycle 5 Reload, dated July 12, 1983.
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