ML20078C951
| ML20078C951 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/21/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078C940 | List: |
| References | |
| NUDOCS 8309280366 | |
| Download: ML20078C951 (8) | |
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[ *4 UNITED STATES w
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 e.... i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 61 TO FACILIT't OPERATING LICENSE NPF-3 TOLEDO EDISON COMPANY AND CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346
- 1. 0 Introduction By letter dated July 5,1983 (Ref.1), Toledo Edison Company (the licensee) made application to modify the Davis-Besse Nuclear Power Station, Unit No. 1, Technical Specifications to permit operation for a fourth cycle.
The analysis performed and the resulting modifications to the Technical Specifications are described in the Unit 1, Cycle 4 Reload Report (Ref. 2).
The safety analysis for the previous third cycle of operation at Davis-Besse 1 is being used by the licensee for the proposed fourth cycle of operation.
Where conditions are identical or limiting in the third cycle analysis, our previous evaluation (Ref. 3) of tnat cycle continues to apply.
1.1 Description of the Cycle 4 Core The Davis-Besse Cycle 4 core will consist of 177 fuel assemblies, each of which is a 15x15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube.
Cycle 4 will operate in bleed-and-feed mode with core reactivity control supplied mainly by soluble baron in the reactor coolant and supplemented by 53 full length control rod assemblies (CRAs).
In addition, eight axial power shaping rods (APSRs) e.re provided for additional control of the axial power distrio' ution.
No burnable poison rods will be utilized in the Cycle 4 core.
The length of Cycle 4 is expected to be 240 effective full power days (EFPD) of operation, marginally lower than the 268 EFPD accumulated during Cycle 3.
The licensed core full power level remains at 2772 MWt.
2.0 Evaluation of the Cycle 4 Core l
2.1 Fuel System Design s
The 48 Babcock and Wilcox (B&W) Mark-B4 fuel assemblies loaded as Batch 6 at end of Cycle 3 (EOC 3) are mechanically interchangeable with Batches 1D, 28, 4, 5A and 5B fuel assemblies previously loaded at Davis-Besse Unit 1.
The cladding l
stress, strain and collapse analyses are bounded by conditions previously l
analyzed for Davis-Besse 1 or were analyzed specifically for Cycle 4 using l
methods and limits previously reviewed and approved by the NRC.
End-of-life l
fuel rod internal pressures have also been analyzed using previously-approved methods and limits.
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8309280366 830921 PDR ADOCK 05000346 P
The thermal behavior of the fuel in the Cycle 4 core has been analyzed with I
two B&W fuel thermal performance codes, TAFY-3 (Ref. 4) and TACO-2 (Ref. 5).
Although both of these codas have been approved for use in safety analysis, we i
believe (Ref. 6) that only the newer TACO series of codes are capable of cor-rectly calculating fission gas release (and therefore rod pressure) at high i
burnups.
Babcock & Wilcox has responded (Ref. 7) to this concern with an analytical comparison between the TAFY-3 code and an earlier version of TACO l
called TACO-1 (Ref. 8).
In this response, they have stated that the fuel rod internal pressure predicted by TACO-1 is lover than that predicted by TAFY-3 for fuel rod exposures of up to 42 mwd /kgU.
The licensee has stated that the maximum expected exposure of any fuel rod during Cycle 4 is less than this amount.
We find this acceptable.
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For the Loss of Coolant Accident (LOCA) analysis (Section 7.2 of the Reload i
Report), the volume-averaged fuel temperature and fuel rod internal pressure j
were calculated for Cycle 4 as a function of linear heat rating.
The licensee has stated that these conditions are bounded by those used in the generic LOCA t
analysis for Davis-Besse Unit 1.
i As mentioned previously, B&W currently has several fuel performance codes which j
are approved and could be used to calculate LOCA initial conditions.
The older TAFY-3 code was used for the generic LOCA analysis cited in the Cycle 4 Reload Report.
Information obtained by the NRC staff (Ref. 9) indicates that the TAFY-3 code predictions do not produce higtr~ calculated peak cladding tempera-tures in the generic LOCA analysis than the newer TACO-1 or TACO-2 codes as l
suggested by the licensee.
The issue involves excessive fuel densification and i
lowered fuel rod internal gas pressures at beginning of life.
Babcock and Wilcox has proposed a method of resolving this issue which has been adopted by Toledo Edison Company (Ref. 10).
The method relies on reduced peak linear heat rate (PLHR) limits at low core elevations for the first 24 effective full power l
days (EFPD) of operation based on comparison of TAFY-3 and TACO-2 calculated l
LOCA initial conditions.
The method is similar to an older TAFY-3/ TACO-1 i
comparison used in the Davis-Besse 1 Cycle 3 safety analysis.
However, the resulting PLHR reduction is different for each code.
In addition to the issue of initial fuel temperatures and rod internal pres-sures used in the LOCA analysis, a second issue involving cladding swelling and rupture models has affected the proposed Cycle 4 operating limits for Davis-Besse 1.
In late 1979, tha NRC staff reviewed Emergency Core Cooling System (ECCS) fuel cladding models in light of new data.
Adequacy of the models then in use was questioned and new models, developed as Appendix K acceptance cri-teria, were presented in NUREG-0630 (Ref. 11).
Each fuel vendor was then asked to show how, in light of the new models, the plants analyzed with their analyt-ical methods continued to meet the applicable LOCA limits.
The B&W response (Ref. 12) concluded that the impact of the NRC models was small and did not result in analytical results in excess of the LOCA limits.
A more recent B&W calculation (Ref. 13), however, found that the cladding swel-ling and rupture models presented by the staff have a non-trivial effect on LOCA peak cladding temperatures in B&W 177 fuel assembly plants.
Because this
. calculation was applicable to all B&W plants, the licensee was requested (Ref. 14) to provide supplemental calculations for Davis-Besse Unit 1 similar to those provided in Reference 13.
The licensee's responses (Refs. 15 to 18) culminated in the supplemental calculation (Ref. 10) cited previously.
This j
calculation, which considers both fuel densification (TAFY-3/ TACO-2) and clad-ding swelling and rupture effects, results in low core elevation PLHR limits which are more restrictive than those which consider only fuel densification.
The licensee has proposed (Ref. 2) modification to the Davis-Besse 1 Technical Specifications which account for these reduced PLHR limits.
In general, the supplemental calculation utilizes previously approved methods except for the substitution of the NRC cladding models.
However, there are segments of the analysis (e.g. THETA 1-B - Ref.19) that are currently under-going NRC review.
Babcock & Wilcox has also presented results from a calcula-tion using a new FLECSET' heat transfer correlation (Refs. 20 and 21).
This correlation appears to offset the NUREG-0630 penalties.
The licensee has not yet claimed these FLECSET benefits, however, because the benchmarking and other final evaluations of FLECSET have not been completed and provided to the NRC for review.
Considering the above, we conclude that the licensee's proposed Technical Speci-fication changes are both appropriate and necessary.
Since these operating limits are more restrictive than those previously used at Davis-Besse Unit 1, since they are only needed for a brief time period, and since potential for unused compensating benefits may exist, we, thertfore, conclude that the oper-ating restrictions imposed on an interim basis are acceptable for incorporating 4
the NUREG-0630 penalties until our final evaluation of FLECSET is completed.
- 2. 2 Nuclear Design To support Cycle 4 operation of Davis-Besse Unit 1, the licensee has provided analyses (Ref. 2) using analytical techniques and design bases established in B&W reports that have oeen approved by the NRC staff.
The validity of the methods also has been reinforced through predictions of a.; umber of cycles for this and other reactors.
The licensee has provided a comparison of the core physics parameters (Ref. 2) for Cycles 3 and 4 as calculated with these techniques.
We reviewed the characteristics comptred to previous cycles, and find them acceptable for use in the Cycle 4 accident and transient analysis, as described in Section 2.4 of this evaluation.
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There are no significant core design changes between the reference Cycle 3 and Cycle 4 designs.
The Cycle 4 core was shuffled in a manner to minimize the carryover effect on quadrant tilt.
The Cycle 4 design cycle length is 240 days, i
whereas the Cycle 3 design cycle length was 268 days.
No significant operation-al or procedural changes exist for Cycle 4 with regard to axial or radial power shape, xenon, or tilt control.
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Due to the differences in design cycle lengths, the critica'l boron concentra-tions for Cycle 4 differ from those of Cycle 3.
Because of different isotopic distributions, Cycle 4 control: rod worths, ejected' rod worths, and stuck rod l
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. worths differ from those of Cycle 3.
The licensee took into account ejected rod worths and their adherence to shutdown margin requirements in the develop-ment of rod position limits for Cycle 4.
The licensee presented an analysis of shutdown margin adequacy as a function of predicted control and stuck rod worths.
This analysis allowed for a 10 percent uncertainty on net rod worth and for flux redistribution.
It shows considerable margin in excess of requirements.
We, therefore, conclude that the licensee has demonstrated adequate provision of shutdown margin for Cycle 4.
In addition, control rod worth measurements are made during startup tests.
These confirm the adequacy of predicted control rod worths.
2.3 Thermal-Hydraulic Design The thermal-hydraulic performance for Cycle 4, in which the fresh Batch 6 fuel is hydraulically and geometrically similar to the other fuel in the Cycle 4 core, is identical to that of Cycle 3.
The thermal-hydraulic design evaluation supporting Cycle 4 operation is based on the methods and models previously used in Cycle 3 as described in References 22 and 23.
The design conditions are given in Table 1 and are identical fcr Cycles 3 and 4.
A rod bow topical report (Ref. 24) was submitted and approved (Ref. 25) since the last fuel cycle.
This report addressed the mechanisms and resulting local conditions of rod bow.
The conclusion was that rod bow penalty is insignificant and is~ offset by the reduction in power production capability of the fuel assemblies with irradiation.
Therefore, there is no resulting rod bow penalty for Cycle 4.
The flux / flow trip setpoint for Cycle 4 has been established as 1.069 '(Ref. 26) and was 1.070 for Cycle 3.
This setpoint and other plant operating limiti are based on criteria that meet the design minimum Departure from Nucleate Boiling Ratio (DNBR) limit of 1.30 calculated using the BAW-2 correlation.
The minimum DNBR at 112 percent of full power is 1.79 for Cycle 4 which is the same as for Cycle 3.
The MRC staff finds that the thermal-hydraulic design is acceptable since the Cycle 4 and Cycle 3 (previously approved) design conditions are identical and acceptable design methods have been used in the analysis.
2.4 Accident and Transient Analys~is Acceptability of core thermal, thermal-hydraulic, and kinetics parameters, in-cluding the reactivity feedback coefficients and control rod worths, was dis-cussed in Sections 2.2 and 2.3.
The licensee concluded, by examination of the Cycle 4 values of these parameters with respect to acceptable previous cycle values, that transients and accidents for Cycle 4 are bounded by previously accepted analyses.
A supplemental ECCS calculation (Ref. 10) for Davis-Besse 1 has resulted in reduced PLHR limits at lower core elevations (see Section 2.1).
The new LOCA
. limits were used in the generation of more stringent control rod insertion and imbalance limit curves for the first 24 EFPD of Cycle 4.
These revised curves have been included in the proposed Technical Specifications for Cycle 4.
2.5 Technical Specification Modifications 1
The pertinent Technical Specifications have been revised for Cycle 4 operation to account for changes in power peaking and control rod wcrths as discussed in Sections 2.2 and 2.4.
We have reviewed these changes as proposed in Reference 2 and find them all acceptable.
- 2. 6 Summary We conclude from the examination of Cycle 4 ccre 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 Davis-Besse Nuclear Power Station's ability to operate safely during Cycle 4.
h 3.0 Environmental Consideration We have determined that the amendment does 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 the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR S51.5(d)(4),
that an environmental impact statement, or negative declaration and environmen-tal 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) 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 Commission's 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.
The.following NRC personnel contributed to this Safety Evaluation:
H. Balukjian, M. Dunenfeld, J. Voglewede.
Dated:
September 21, 1983 I
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. Table 1 Davis-Besse Cycles 3 and 4 Thermal-Hydraulic Design Conditions Design power level, MWt 2772 System pressure, psi 3 2200 Reactor coolant flow, gpm 387,200(b)
Reactor coolant flow, % design 110 Vessel inlet / outlet coolant temp., 100% power, F 557.7/606.3 Ref design radial-local power peaking factor 1.71 Ref design axial flux shape 1.5 cosine with tails Hot channel factors Heat flux (F"9) q)
Enthalpy rise (F 1.011 1.014 Flow area 0.98 2
1.89x10s(a)
Avg heat flux, 100% power, Btu /h-ft Max heat flux, 100% power, Btu /h-ft2 4.85x105(a)
CHF correlation
.BAW-2 Minimum DNBR (at 112% power)( )
1.79 (a) With thermally expanded fuel rod OD of 0.43075 inch.
(b) Telecon, G. Bradley, Toledo Edison, to A. DeAgazio, NRC, September 1, 1983.
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- REFERENCES
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1.
R. P. Crouse (Toledo Ed) letter to J. F. Stolz (NRC) dated July 5, 1983.
I 2.
" Davis-Besse Nuclear Power Station Unit 1, Cycle 4 - Reload Report,"
Babcock & Wilcox Company Report BAW-1783, May 1983.
Attachment to Reference 1 above.
3.
J. F. Stolz (NRC) letter to R. P. Crouse (Toledo Ed) on " Amendment No. 45 to Facility Operating License' No. NPF-3; Cycle 3 Operation," dated July 28, 1982.
4.
C. D. Morgan and H. S. Kao, "TAFY - Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.
5.
Y. H. Hsii et al., " TACO 2:
Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10141P, January 1979.
6.
D. F. Ross, Jr. (NRC) letter to J. H. Taylor (B&W) dated January 18, 1978.
7.
J. H. Taylor (S&W) letter to L. S. Rubenstein (NRC) dated September 5, 1980.
8.
R. H. Stoudt et al., " TACO:
Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-30087P-A, Rev. 2, August 1977.
9.
R. O. Meyer (NRC) memorandma for L. S. Rubenstein (NRC) on "TAFY/ TACO Fuel Performance Models in B&W Safety Anal > sis" dated June 10, 1980.
10.
R. P. Crouse (Toledo Ed) letter to J. F. Stolz (NRC) dated May 6, 1983 and transmitting " Bounding Analytical Assessment of NUREG-0630 on LOCA and kW/ft Limits," B&W Document No. 77-1142162-00.
11.
D. A. Powers and R. O. Meyer, " Cladding Swelling Models for LOCA Analysis,"
U. S. Nuclear Regulatory Commission Report NUREG-0630, April 1980.
12.
J. H. Taylor (B&W) letter to L. S. Rubenstein (NRC) dated October 28, 1980.
13.
J. W. Cook (Consumers Power) letter to H. R. Denton (NRC) dated April 2, 1982 and transmitting B&W Report No. 12-1132424, Revision 0, " Bounding Analysis Impact Study of NUREG-0630."
14.
T. M. Novak (NRC) letter to R. P. Crouse (Toledo Ed) dated July 13, 1982.
15.
R. P. Crouse (Toledo Ed) letter to T. M. Novak (NRC) dated August 13, 1982.
16.
R. P. Crouse (Toledo Ed) letter to T. M. Novak (NRC) dated October 19, 1982.
e o 17.
R. P. Crouse'(Toledo Ed) letter to J. F. Stolz (NRC) dated February 17, 1983.
18.
R. P. Crouse (Toledo Ed) letter to J. F. Stolz (NRC) dated April 5, 1983.
19.
" Babcock & Wilcox Revisions to THETA 1-B, a Computer Code for Nuclear Reac-tor Core Thermal Analysis (IN-1445) - Revision 3," Babcock & Wilcox Company Report BAW-10094, Rev 3, February 1981.
20.
N. Lee, S. Wong,'H. C. Yeh, and L. E. Hochreiter, "PWR FLECT SEASET Unblocked Bundle, Forced and Gravity Reflood Test Data Evaluation and Analysis Report, NUREG/CR-2256 (EPRI NI-2013 or WCAP-9891), November 1981.
21.
G. P. Lilly, et al., PWR FLECHT Skewed Profile Low Flooding Rate Test Series Evaluatiot Report," WCAP-9183, November 1977.
22.
" Davis-Besse Unit 1 Fuel Densification Report," Babcock & Wilcox Company Report BAW-1401, April 1975.
- 23. to " Application to Amend Operating License for Removal of Burnable Poison Rod and Orifice Rod Assemblies," Babcock & Wilcox Company Report BAW-1489, Rev. 1, May 1978.
24.
" Fuel Rod Bowing in Babcock & Wilcox Fuel Designs," Babcock & Wilcox Com-pany Report BAW-10147P (Proprietary), April 1981.
25.
C. O. Thomas (NRC) letter to J. H. Taylor (B&W) dated February 15, 1983.
26.
G. Bradley (Toledo Ed) telecon with A. DeAgazio (NRC) dated September 1, 1983.
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