ML20084D007
| ML20084D007 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/20/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20084C991 | List: |
| References | |
| TAC-53319, TAC-53320, NUDOCS 8405010120 | |
| Download: ML20084D007 (12) | |
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SAFETY EVALUATION REPORT RELATED TO AMEN 0 MENT N0.w TO FACILITY OPERATING LICENSE NPF-9 AND TO AMENDMENT NO.13T0 FACILITY OPERATING LICENSE NPF-17 DUKE POWER COMPANY INTRODUCTION By letter dated December 12,1983 (Ref.1) Duke Power Company (the licensee) made application to amend the Technical Specifications of McGuire Units 1 and 2 to reflect transition from standard design Westinghouse fuel assemblies (STD) to Westinghouse optimized fuel assemblies (CFA). This transition will be carried out over 4 cycles for each unit. The December 12, 1983 submittal supports the first phase of transition for McGuire Unit 1 from STD to 0FA during Cycle 2.
The February 20, 1984 change, which corrected an inconsistency between the time constant values for McGuire Units 1 and 2, was part of the original application as noticed (49FR7893). The March 23, 1984 changes, addressing withdrawal of the prior request for deletion of Technical Specification reporting requirements, and correction of Technical Specifi-cation bases were not in the application as noticed.
However, these changes do not substantially affect the application as noticed. Additional infor-mation was also provided by letter dated March 9,1984.
t In addition to the reload changes indicated above, there are two non-reload
- related changes included. These involve deletion of the Boron-Injection System for Unit 1 (Technical Specification 3/4.5.4) and revised control rod insertion limits for both Units 1 and 2 (Technical Specification 3/4.1.3.6).
These changes are included in this safety evaluation.
EVALUATION 1.0 Fuel Mechanical Design McGuire Units 1 and 2 have been operating with Westinghouse 17x17 low para-sitic fuel assemblies.
This fuel is designated by Westinghouse as STD.
The licensee has stated (Ref.1) that both McGuire units will be refueled
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- --...-- =. - - -. -. =-- 2-with Westinghouse 17x17 reconstitutable optimized fuel assenblies (0FA). The OFA fuel has similar design features to the STD fuel. The major differences are the use of six intennediate (mixing vane) Zircaloy grids for the OFA fuel versus six intermediate (mixing vane) Inconel grids for j STD fuel, a reduction in fuel rod, guide thimble, and instrumentation tube diameters, and the replacanent of a standard bottom nozzle with a i reconstitutable botton nozzle. l t j The OFA has been designed to be mechanically compatible with the STD design, I reactor internal interfaces, fuel handling and refueling equipment, and f spent fuel storage racks. The top and bottom Inconel (non-mixing vane) l grids of the 0FA are nearly identical to the Inconel grids of the STD design. The only difference is that the spring and dimple heights have been modified to accommodate the reduced diameter of the OFA fuel rod. ] The six intennediate (mixing vane) grids are made of Zircaloy rather than j Inconel which is currently used in the STD design. The Zircaloy grids i have thicker straps than the Inconel grids; also, the Zircaloy grid height j is 2.25 inches compared to the Inconel grid height of 1,32 inches. These dimensional changes were made'to campensate for differences in material i strength. The Westingnouse 17x17 0FA grid design, which was described in i WCAP-9500-A, has been reviewed and approved by the NRC staff (Ref. 2). In perfonning our review of the 17x17 0FA fuel for McGuire Unit 1 Cycle 2 we focused on those issues identified in the staff's SER on Westinghouse OFA fuel j as requiring a plant specific review. Our evaluation of those issues follows, i j 1.1 Cladding Collapse I j The licensee uses an approved method described in WCAP-8381 (Ref. 3) to analyze cladding collapse and has stated (Ref.1) that the cladding collapse is not predicted during Cycle 2 operation. i i i i i i } t -,er,, - - - -,,,,,,,, - --r, -,,,,,,r ,,,,,.,,n,,e---e. g.w,, ,,w-,,,-,, ,g -en-h-Q g,--, -.wn, .emy.-- ->-r---, -~,g. w----
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i l l l 1.2 Post-Irradiation Surveillance The licensee will implement a fuel inspection program on the irradiated and discharged optimized fuel fra the initial reload region. The program will involve visual examinations on a sample to detect crud buildup, rod bowing, grid strap conditions and missing components. The licensee stated (Ref. 4) that additional fuel inspections would be perfomed depending on the results of operational monitoring, including coolant activity, and the visual fuel inspections. The staff finds this surveillance program for the OFA fuel to be acceptable. 1.3 Demonstration Assembly The reload will contain one 0FA demonstration fuel assembly with the addition of three intemediate flow mixer grids and 88 removable fuel rods. The remov-able rods contain fuel enriched at 3.10 w/o U-235, and the remaining rods are 3.20 w/o U-235. The staff finds the use of this one demonstration assembly acceptable since the licensee has stated that it will be located in a non-limiting location. This is especially important because of the inconpatibility in the number of grids between the demonstration assembly and the OFA and STD fuel designs. 2. Nuclear Design The licensee stated that the transition from the STD to the OFA design will not result in changes from the current nuclear design basis. Although the core physics characteristics are slightly different for the OFA fuel compared to STD, we have reviewed these differences and detemined that the differences are within the normal range of variations seen from cycle to cycle. Thus, we find the change to 0FA acceptable. A number of changes to the McGuire Technical Specifications have been proposed by the licensee as part of the transition to 0FA fuel. These changes include the positive moderator temperature coefficient (MTC) specification; the 0.3 multiplier in the F limit function; a reduction in the required shutdown AH margin (SDM) to 1.3%ap; F surveillance employing a relaxed axial offset q control (RAOC). These Technical Specification changes are addressed later in the Technical Specification section of this SER.(Section 6).
I ! l l 3. Themal-Hydraulic Design i The themal hydraulic design analysis of this mixed core was performed with l the improved themal design procedure (ITDP) (Ref. 5), the THINC-IV code and i the WRB-1 critical heat flux correlation. The THINC-IV code as described in WCAP-7956 and its design application described in WCAP-8050 has been approved for safety analysis. The WRB-1 correlation has been approved for 17x17 f' 0FA with a DNBR limit of 1.17. The ITDP has been approved (Ref. 6) with a condition requiring the use of plant-specific uncertainty distributions of 4 i the pertinent parameters for the design DNBR analysis. In response to a staff question, the licensee in a letter dated March 9,1984 (Ref. 4) provided measurement uncertainty values of the pressurizer pressure, reactor i l coolant temperature, reactor power and RCS flow rate. A detailed measure-i ment component uncertainty breakdown of these parameters and the statistical method of combining these camponent uncertainties are also provided. In addition, the licensee also provided a calculation based on the uncertainty f distributions of the pertinent parameters to derive the design DNBR limits of 1.335 and 1.316 for the typical cell and the thimble cells, respectively. This analysis was perfomed with the same sensitivity factors used in ~ WCAP-9500 which the licensee has detemined to be applicable for the McGuire j stations. Therefore, the design DNBR limits of 1.34 and 1.32 for the typical and thimble cells, respectively, as provided in the report are acceptable. The safety analysis is performed with plant-specific safety DNBR limits of 1.49 and 1.47, respectively, for the typical and thimble cells. Therefore, there is about 10 percent DNBR margin available for design flexibility, i The use of the ITDP for the analysis of a transitional mixed core has previously been approved (Ref. 7) with a condition requiring a penalty on l DNBR to account for the uncertainty associated with interbundle crossflow in j the mixed core. For the 17x17 0FA fuel, a 5 percent penalty was previously proposed as a generic application for WCAP-9500 by Westinghouse (Ref. 8). l However, a comprehensive review by the licensee of all transition patterns ] versus an all 0FA pattern over appropriate ranges indicates that the maximum j required penalty is 1.9% DNBR and the 5 percent DNBR penalty is unnecessarily l conservative. Therefore, the licensee proposed to apply 2 percent DNBR penalty for the transitional mixed core and we find it acceptable. I u......
. The fuel rod bow penalty must be accounted for in the safety analysis. Using the approved method described in WCAP-8691, Revision 1 (Ref. 9), the maximu.. rod bow penalties are detennined to be less than 6 percent DNBR for the 17x17 0FA fuel and less than 5 percent for the 17x17 standard fuel. The rod bow penalty and the transitional mixed core penalty can be compenstated by the available thennal margin of 10 percent discussed earlier. 4. Transients and Accident Analyses All of the non-LOCA transients and accidents were reanalyzed to include the following major design changes: a. replacing 60 STD fuel assemblies with 60 Region 4 optimized fuel assemblies, b. use c,f positive moderator temperature coefficient, c. reduction of shutdown margin to 1.3% Ap, e d. use of the Improved Thermal Design Procedure with both the WRB-1 and W-3 DNB correlations, e. change in the Nuclear Hot Channel Factor (Fg) All the transients and accidents were done using approved methods ano acceptable initial conditions. The results presented were acceptable since they did not violate the specified acceptable fuel design limits as required by GDC 10. The analyses of those transients reanalyzed with the Westinghouse Improved Thennal Design Procedure were done in accordance with our SER on this method (Ref. 6).
o ) 5. LOCA The LOCA analysis was done using Westinghouse methods which have been approved by the staff. The large break LOCA was shown to be more limiting than the small break LOCA - the results of the analysis were within the limits of 10CFR50.46. The maximum cladding tenperature for the large break LOCA is 2175*F which is less than the acceptance criteria of 2200*F. The maximum local metal-water reaction is well below the embrittlement limit of 17% as required by 10CFR50.46 and the total core metal-water reaction is less than 0.3% as canpared with the one percent criterion of 10CFR50.46. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided. Thus, we conclude that the LOCA analysis is acceptabic and that a loss of coolant accident will not result in a health hazard to the public. 6. Technical Specification Changes We have reviewed the Technical Specification changes proposed in Reference 1 and modified in References 10 and 11 for McGuire Units 1 and 2 and find them acceptable. The changes and reasons for their acceptability are discussed bel ow. 1. Section 2.1.1, Reactor Core Safety Limits. The changes in this section have been made for the core coolant average temperature limit as a function of thermal power and primary system pressure. This change is necessary due to the use of the improved thermal design procedure where the uncertainties associated with the coolant temper-ature, pressure and power are treated statistically, and the safety analysis is performed with nominal values of these parameters. This results in the permissible operating space of the safety limit being reduced to reflect the fact that the nominal values of these parameters are used as the safety limit. This change is acceptable. l
( s 2. Section 2.2.1, Reactor Trip System Setpoints for OT AT and ORtT for Unit 1. The changes in this section have been made in accord-ance with standard procedures (Ref.12) used for all cycles of all Westinghouse designed reactors which have been approved by the staff, and are, therefore acceptable. 3. Section 3/4.1.1.1, Shutdown Margin-Tavg > 200*F. The changes in this section reduce the shutdown margin for Unit 1 to 1.3% from 1.6%. The steamline break accident analysis was perfonned with the reduced shutdown margin with acceptable results as discussed in Section 4 of this evaluation. These changes are, therefore, acceptable. 4. Section 3/4.1.1.3, Moderator Temperature Coefficient. The changes in this section allow the moderator coefficient to be slightly positive, up to 5 pcm/*F and up to 70% of rated thennal power for Unit 1. The accident analysis was conservatively perfonned with this coefficient at full power, except the rod ejection analysis which used a variable coefficient consistent with the changed value. The results were acceptable as discussed in Section 5 of this evaluation, and, therefore, these changes are acceptable. 5. Section 3/4.1.3.4, Rod Drop Time. The changes in this section increase the Unit 1 allowable control rod drop time to 3.3 seconds from 2.2 seconds. The Unit 2 control rod drop time is already specified at 3.3 seconds. The increased control rod drop time was taken into account in the scram curves used for the accident analysis with acceptable results as discussed in Section 4 of this evaluation. These changes are, therefore, acceptable. 6. Section 3/4.1.3.6, Control Rod Insertion Limits. The changes in this section revise the control rod insertion limits for both Units 1 and 2. As a result of hardware limitations and an error j' made in calculation of the insertion limits, the existing Technical
e i. Specifications are slightly more limiting than necessary for D bank insertion. The revised limits wer? used throughout the McGuire safety analysis and are, therefore, acceptable. i l 7. Section 3/4.2.1, Axial Flux Difference Limits. The changes in this section implenent a Westinghouse power distribution control method-ology called Relaxed Axial Offset Control (RAOC). This is an approved methodology (Ref.13) and is, therefore, acceptable. l' F ( Z). The 8. jection 3/4.2.2, Heat Flux Hot Channel Factor changes in this section implenent a peaking factor (F ) surveil-q lance technical specification which has been approved (Ref.14) and is, therefore, acceptable. 4 l l 9. Section 3/4.2.3 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Fac tor. There are two sets of changes in this section. The first involves the elimination from Figure' 3.2-3 of. the R2 function which* - accounts for fuel rod bow penalty. Since the calculation using the approved method described in WCAP-8691, Revision 1 has resulted in i rod bow penalty of less than 6%, the~ rod bow penalty is compen-i sated for by the available thermal margin. Therefore the elimination of R -function is acceptable. The second involves the 2 multiplier in the nuclear enthalpy rise hot channel factor (FAH) i to increase as equation for Unit 1. This multiplier allows FAH power decreases. This occurs naturally in a Westinghouse reactor as feedback froa the negative reactivity coefficient decreases, and as control rods are inserted in the reactor. The existing j, multiplier of 0.2 allows a linear increase in F from 0 at full ~ 3g power to 20t at zero power. The proposed multiplier of 0.3 allows a linear, increase which is 30% at zero power. This has been taken into account in the themal hydraulic analysis. In addition the same multiplier change has been approved for a number of Westinghouse reactor: ~ tn recent years, and is, therafore, acceptable. i = J ~
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- 10.
Section 3/4.5.1.1 Accumulators-Cold Leg Injection. The changes in this section reflect the assumptions of accumulator borated water volume used in the LOCA analysis for Unit I with acceptable results as discussed in Section 5 of this evaluation. These changes are, therefore, acceptable. 11. Section 3/4.5.4, Boron Injection System Boron Injection Tank. This change eliminates the boron injection tank (BIT) fran the Technical Specifications for Unit 1. A previously approved Technical Specifi-cation change reduced the BIT to 2000 ppn. The analysis supporting that change showed that the steamline break accident is safety mitigated without consideration of the BIT functioning. This led to licensing of Unit 2 without a BIT. This change is, therefore, acceptable. 12. Section 6.9.1.12, Radial Peaking Factor Report. This section is appropriately changed to meet the requirements for F surveillance q for Unit 1 discussed in item 8 above and the procedures approved in Reference 14. CONCLUSION The Commission made a proposed determination that the amendments involve no significant hazards consideration which was published in the Federal Register (49 FR 7893) on March 2,1984, and consulted with the state of North Carolina. No public comments were received, and the state of North Carolina did not have any comments. In conclusion the staff finds the proposed changes to the plant technical specifications to be acceptable and 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 these amendments will not be inimical to the common defense and security to the health and safety of the public.
o 10 - ENVIR0fMENTAL CONSIDERATION We have detennined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any signf ficant environmental impact. Having made this determination, we have further concluded that the amendments involve an' action which is insignificant from tha standpoint of environmental impact and, pursuant to 10 CFA 51.5(d)(4), that an environmental impact statensnt or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments. t e 4 k \\ i I 4 4 E L i
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I REFERENCES 1. Letter to H. R. Denton (NRC) from H. S. Tucker (Duke Power) "McGuire Nuclear Station, Docket Nos. 50-369 and 50-370, McGuire 1/ Cycle 2 l I OFA Reload", December 12, 1983. 2. L. S. Rubenstein (NRC) memorandum, " Safety Evaluation Report on WCAP-9500", to R. L. Tedesco (NRC), May 15, 1981. j 3. R. A. George (et, al), " Revised Clad Flattening Model", WCAP-8381, July 1974. l l 4. Letter from H. B. Tucker (Duke Power Company) to H. R. Denton, "McGuire Nuclear Station, Docket Nos. 50-369 and 50-370, McGuire 1/ Cycle 2 0FA I Reload" March 9,1984. 5. H. Chelemer, L. H. Boman, D. R. Sharp, " Improved Thermal Design Procedure", WCAP-8567P, July 1975. 6. D. F. Ross, Jr. (NRC) memorandum, " Improved Thermal Design Procedure, WCAP-8567/8568"; " Application of the THINC-IV program to PWR design, WCAP-8054"; "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids, WCAP-8762"; to D. B. Vassallo (NRC) April 10, 1978. 7. Memorandum from L. S. Rubenstein to T. M. Novak, " Supplemental Infomation on WCAP-9500 and WCAP-9401/9402 Mixed Core Compatibility", December 14, 1982. 8. Letter from E. P. Rahe (Westinghouse) to J. R. Miller (NRC), NS-EPR-2643, " Supplement to WCAP-9500 and WCAP-9401/9402 NRC Safety Evaluation 4 Report Mixed Core Compatibility Items Supplemental Infomation", August 17, 1982. i 1 _j
e ! 9. WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation", July 1979, Westinghouse Corp.
- 10. Letter to H. R. Denton (NRC) from H. B. Tucker (Duke Power) "McGuire Nuclear Station, Docket Nos. 50-369 and 50-370, McGuire 1/ Cycle 2 0FA Reload", February 20, 1984.
- 11. Letter to H. R. Denton (NRC) from H. B. Tucker (Duke Power) "McGuire Nuclear Station, Docket Nos. 50-369 and 50-370, McGuire 1/ Cycle 2 0FA Reload", March 23, 1984.
12. S. L. Ellenberger, et al., " Design Basis for the Thermal Overpower AT and Thennal Overtenperature aT/ Trip Functions", WCAP-8745, March 1977.
- 13. Memorandum to F. J. Miraglia from L. S. Rubenstein, " Review of Westinghouse Topical Report NS-EPR-2649, Part A", November 17, 1982.
- 14. Memorandum to F. J. Miraglia from L. S. Rubenstein, " Review of Westinghouse Topical Report NS-EPR-2649, Part B", November 17, 1982.
Principal Contributors: A. Gill, Core Performance Branch, DSI Y. Hsii, Core Performance Branch, OSI D. Fieno, Core Performance Branch, OSI R. Lobel, Core Performance Branch, DSI M. Dunenfeld, Core Performance Branch, DSI V. Leung, Reactor Systems Branch, DSI R. Birkel, Licensing Branch No. 4, DL}}