ML19274E512

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Forwards Proposed Tech Spec Change to Permit Increase to 1.93 in Allowable LOCA Peaking Factor Limit.Amend Explains Basis for LOCA Reanalysis.Check for Amend Fee Encl
ML19274E512
Person / Time
Site: Zion  File:ZionSolutions icon.png
Issue date: 03/22/1979
From: Reed C
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 7903290157
Download: ML19274E512 (66)
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{{#Wiki_filter:Commonwealth Edison One First National Plaza Chicago, Illinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 March 22, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Zion Station Units 1 and 2 Proposed Change to Facility Operating License Nos. DPR-39 and DPR-48 hPC Docket Nos. 50-295 and 50-304 Reference (a) : February 16, 1979 letter from A. Schwencer to Cordell Reed

Dear Mr. Denton:

Pursuant to 10 CFR 50.59, Commonwealth Edison Company hereby requests a change to Operating License Nos. DPR -39 and DPR-48, Appendix A, Technical Specifications. The purpose of this amendment is to revise the Zion Technical Specifications to increase the allowable LOCA peaking factor limit from the current value of 1.86 to a value of 1.93 based on an ECCS reanalysis performed in accordance with the Westinghouse ECCS Evaluation Model approved by the Staff in August 1978. The proposed changes to the Zion Station Technical Specifications are enclosed in. contains the LOCA reanalysis for Zion Units 1 and 2. The LOCA reanalysis was performed for a volume of 818.65 cubic feet per accumulator tank. Using this volume the peak clad temperature calculated for the limiting break (DECLG, CD = 0.8) is 2156.44 F with an F peaking factor limit g of 1.93. This reanalysis differs from that most recently approved by the NRC Staff in Reference (a) in that the: (1) Average fuel pellet temperature was lowered 20 F based on Zion generic fuel temperature data, and O b ny 7903290 /67

Commonwealth Edison NRC Docket Nos. 50-295/304 Mr. Harold R. Denton: March 22, 1979 (2) Average fuel pellet temperature was lowered an additional 650F to remove a "modeling conservatism" that was previously added to the average fuel temperatures calculated by the PAD computer code. The average fuel pellet temperature that exists at the initiation of a hypothetical LOCA has a direct effect on the peak clad temperature achieved in the course of the LOCA. This is due to the greater stored energy in the pellet which directly contributes to the increase in clad temperature. Additional information concerning the above temperature changes is provided below: Use of Zion Generic Fuel Temperature Data The previous LOCA analyses for the Zion units have used Westinghouse generic fuel parameters for detennining the initial fuel pellet temperatures. These generic parameters are conservatively chosen to bound the nonnal fuel cycle to fuel cycle variations associated with the fuel assembly fabrication process. Although the generic values are not required by 10 CFR 50.46 Appendix K, they have been used to preclude the possibility of frequent costly reanalyses whenever a given fuel cycle's "as-built" parameters are not bounded by those utilized in the currently approved plant specific ECCS evaluations. However, in order to increase the allowable LOCA peaking farcor limit (Fg), which currently restricts the operating fAexibility of Zion Unit 1 and which will limit the operating oower level of Zion Unit 2 during the next fuel cycle, Westinghouse >erformed the LOCA reanalysis contained in Attachment 2 using

ion generic fuel parameters.

These Zion generic fuel parameters are bounding parameters based on previous fuel reload data and anticipated future reload data. Utilization of these Zion generic fuel parameters reduces the initial average fuel pellet temperature for tha analysis by approxiamtely 200F. As a result, at each Zion reload review Commonwealth Edison will verify that specific reload fuel parameters are bounded by the Zion generic fuel parameters used in the analysis of Attachment 2.

Commonwealth Edison NRC Docket Nos. 50-295/304 Mr. Harold R. Denton: May 22, 1979 Removal of PAD Computer Code Conservatism The Westinghouse computer code, PAD, is used to determine fuel temperatures, pressures and other parameters required for both fuel rod design evaluations and for safety analyses. Incorporated into this code when used for Appendix K LOCA analyses are conservative features that contribute 2000F 0 to 250 F of margin compared to the best estimate fuel temperatures used in Appendix K LOCA analyses. Although not required by 10 CFR 50.46 Appendix K, Westinghouse added an additional 650F for "model conservatism" during the development phase of PAD. A subsequent review of the overall conservatisms in the calculation of fuel temperatures as calculated for in the LOCA analysis has indicated that the 650F model conservatism is not necessary. In fact, elimination of this 650F conservatism still leaves a 2000F to 2500F fuel temperature conservatism in the analysis which results in a conservative analysis of the initial stored energy in the fuel and the peak clad temperatures with the current LOCA/ECCS evaluation models. In addition, based on the Westinghouse " Eighteen Case" analyses, a total peaking factor (Fg) of 2.10 for Zion Unit 1 Cycle 4 and 2.17 for Zion Unit 2 Cycle 4 could occur for the full range of power distributions, including load follow maneuvers, allowable under Constant Axial Offset Control (CAOC). Therefore, in order to accommodate the increase in the F peaking factor g limit from 1.86 to 1.93 axial power distribution monitoring type surveillance will be utilized for power levels above 91.9% and 88.9% of rated power for Zion Units 1 and 2, respectively. In conjunction with the revised peaking factor limit of 1.93, the normalized Fg(Z) envelope, Ky(Z), has also been modified and is included in Attachment 1. The above considerations, proposed technical specification changes (Attachment 1) and ECCS LOCA Reanalysis (Attachment 2) have been reviewed and approved by Commonwealth Edison On-Site and Off-Site Review with the conclusion that there are no unreviewed safety questions.

Commonwealth Edison NRC Docket Nos. 50-295/304 Mr. Harold R. Denton: March 22, 1979 Currently, Zion Unit 2 is in a refueling outage and is expected to return to service for Cycle 4 operation on or about April 22, 1979. With the current Fg peaking factor limit of 1.86 Zion Unit 2 Cycle 4 operation will be restricted during the forthcoming summer peak period to approximately 85% of rated power for the full range of power distributions, including load follow maneuvers, allowable under Constant Axial Offset Control (CAOC). Although a Base Load mode of operation will case this restriction to about 92-93% of rated power, this type of operation significantly restricts the operational flexibility of the unit. However, an increase in the F peaking g factor limit to a value of 1.93 would reduce this costly power restriction by about 4%. Therefore, with these considerations in mind, Commonwealth Edison requests that the NRC Staff expeditiously review and approve the amendment change contained herein prior to the return to operation of Zion Unit 2 for Cycle 4 operation. Pursuant to 10 CFR 170, Commonwealth Edison has determined that this proposed mnendment is a combined Class III and Class I Amendment. As such, commonwealth Edison has enclosed a fee remittance in the amount of $4,400.00 for this proposed amendment. Commonwealth Edison has concluded t'lat the proposed amendment change does not involve a significant hazards consideration since the calculations made in support of this amendment are consistent with well defined and established analysis methods that have received previous NRC Staff approval. Please address any additional questions that you might have to this office. Three (3) signed originals and thirty-seven (37) copies of this letter are provided for your use. Very truly yours, Cordell Reed Assistant Vice-President attachments (2) SUBSCRIBED and SW0 before me thig)9 m_ p to , day of. Ndlleb , 1979. N A 4Lt#/ lH N I GodilG D N6jary Public U

ATTACFl4ENT 1 ZION STATION UNITS 1 AND 2 NRC DOCKET NOS. 50-295 AND 50-304 Proposed Technical. Specification Changes The following pages have been revised: 45 and 63a

SURVRILLANCE REQUIRbl4ENT LIMITING CONDITION FOR OPERATION 3.2.2 Power Distribution Limits 4.2.2 Power Distribution A. Ilot Channel Factor Limits

  • A.

Ilot Channel Factor Limits 1.1 At all times, except during physics 1.1 Following initial core loading tests at $$.75% rated power **, th and at a minimum of regular hot channel factors defined in the effective full power monthly bases must meet the following limits: intervals thereafter, power distribution maps, using the Units 1 and 2 movable detector system, shall be made to confirm that the hot ~ 1.93/P x K1(Z), for P*p.5 channel factor limits of this P (Z) g(Z)f F = L __ L 3.a6 x K1(z), for e g.5 specification are satisfied and E fl.55 1+0.2(1-Ph xRBP, Following initial loading and each subsequent reloading, a where: power distribution map using the Movable Detector System, FO(Z) =Fg(Z) limit; shall be made to ccnfirm that I' ~ power distribution limits are l 1.93 =P constant (LOCA limiting value)) met, in the full power con-q figuration before a unit is fraction of rated power at which operated above 75% of rating. P = andF]H the core operated during Pg measurement) Iq(Z) factor from Figure 3.2-9 selected = at the core elevation, Z, of the measured Fg;

  • The hot channel factors above are defined for a period not to exceed the predicted minimum time to collapse exposure levels for each fuel region as referenced in the bases.
    • During physics tests which may exceed these hot channel factor limits, the reactor may be in this condition for a period of time not to exceed eight hours continuously..

Figure 3 2-9 Hot Channel Factor Normalized Operating Envelope for Units 1 and 2 F, Constant (LOCA Limiting Value) = 1.93 ~. i 't i i i \\ ~O%' m>. JN'\\ /P m m s s ) \\

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a M e ATTACHMENT 2 ZION STATION UNITS 1 AND 2 NRC DOCKET NOS. 50-295 AND Sd-304 ECCS LOCA REANALYSIS

LOCA RZAUALYSIS The Loss of Coolant Accident (LOCA) has been re-analyred for Zion Units 1 and 2. The following information amends the Safety Analysis Report section on Major Reactor Coolant System Pipe Ruptures. The description of the various aspects of the LCCA analysis is 23 given in NCAP-8339 The individual computer codes which comprise the Westinghouse Emergency Core Cooling System (2CCS) evaluation model are described in detail in separate reports [3-6} along with code modifications specified in references 7, 10, and 11. The analysis presented here was performed with the February 1978 version of the evaluation model which includes modifications delineated in references 12, 13, 14 and 15. The analysis in this section was performed with lower steedy state fuel temperatures. Zion generic pellet temperatures were used for 0 Zion rather than Uestinghouse generic pellet temperatures (-20 F). Also, a 650F PAD computer code uncertainty was removed. Nith the 0 0 removal of the 65 F PAD uncertainty a 200 to 250 F fuel temperature conservatism still remains in the analysis. Therefore, removal of this uncertainty still results in a conservative analysia. This uncertainty is not required by Appendix K of 10CFR50 Part 46. Results The analysis of the loss of coolant accident is performed at 102 percent of the licensed core pover rating. The peak linear po'ar and total core power used in the analysis are given in Table 2.

Table 1 presents the occurrence time for various events throughout the accident transient. Table 2 presents selected input values and results from the hot fuel rod thermal transient calculation. For these results, the hot spot is defined as the 1scation of maximum peak clad temperatures. That location is specified in Table 2 for each break analyzed. The location is indicated in feet, which presents elevation above the bottom of the active fuel stack. Table 3 presents a summary of the various containment systems parameters and structura] parameters which were used as input to the COCO computer code b0 used in this analysis. Tables 4 and 5 present reflood mass and energy releases to the containment, and the broken loop accumulator mess and energy release to the containment, respectively. The results of several sensitivity studies are reported These results are conditions which are not limiting in nature and hence are reported on a generic basis. Figu cs 1 through 17 present the transients for the principal parameters for the break sires analyzed. The following items are noted: Figures 1 - 3: Quality, mass velocity and clad heat transfer coefficient for the hotspot and burst locations. Figures 4 - 6: Core pressure, break flow, and core pressure drop. The break flow is the sum of the flowrates from both ends of the guillotine breek. The core pressure drop is taken as the pressure just before ths core inlet to the pressure just beyond the core outlet.

Figures 7 - 9: Clad temperature, fluid temperature and core flow. The clad and fluid temperatures are for the hot spot and burst locations. F igure s 10 - 11: Downcomer and core water level during reflood, and flooding rate. F igures 12 - 13: Emergency core cooling system flowrates, for both accumulator and pumped safety injection. Figures lo - 15: Containment pressure and core power transients. F igure s 16 - 17: Break energy release during blowdown and the containment wall condensing heat transfer coefficient for the worst break. Conclusions - Thermal Analysis For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet b0 That is: the Acceptance Criteria as presented in 10CFR50.46 1. The calculated peak clad temperature does not exceed 2200 F based on a total core peaking factor of 1.93. 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircalloy in the reactor. 3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17% are not exceeded during or after cuenching. 4. The core temperature is reduced and decay heat is removed for an extended period of time, as recuired by the long-lived radioactivity remaining in the core.

i References for Section 15.4.1 1. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled fluclear Po',ter Reactors",10CFR50.46 and Appendix K of 10CFR50.46. Federal Register, Volume 39, i umber 3, January 4,1974. 2. Bordelon, F.l1., itassie, H.'.l., and Zordan, T. A., " Westinghouse ECCS Evaluation ftdel-Suanary", UCAP-8339, July 1974. 3. Bordelon, F.I1., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant", UCAP-8302 (Proprietary Version), WCAP-8306 (!!an-Proprietary Version), June 1974. 4. Bordelon, F.!1., et al., "LOCTA-IV Program: Loss-of-coolant Transient Analysis", WCAP-8301 (Proprietary Version), WCAP-8305 (i!on-Proprietary Version), June 1974. 5. Kelly, R.D., et al., "Ca culational Itdel for Core Reflooding after a loss-of-Coolant Accident (UREFLOOD Code) ". WCAP-8170 (Proprietary Version), WCAP-81?l (l on-Proprietary Version), June 1974. 6. Bordelon, F.li., and 11urphy E.T., " Containment Pressure Analysis Code (C0CO)",UCAP-8327 (Proprietary Version), WCAP-8326 (l:on-Proprietary Version), June 1974. 7. Bordelon, F.it., et al., "The Westinghouse ECCS Evaluation I odel-Suppl e-mentary Information",WCAP-8471 (Proprietary Version),WCAP-8472 (lion-Proprietary Version), January 1975. 8. Salvatori,R., "Uestinghouse ECCS - Plant Sensitivity Studies", L' CAP-8340 (Proprietary Version), UCAP-8356 (lion-Proprietary Version), July 1974. 9. Del eted u.

3 i i 10. " Westinghouse ECCS Evaluation Ebdel, October,1975 Versions",WCAP-8622 4 (Proprietary Version), WCAP-862S (Non Proprietary Version), November,1975. 11. Letter from C. Eicheldinger of Westinghouse Electric Corporation to D.B. Vassalo of the Nuclear Regulatory Concission, letter number NS-CE-924, January 23, 1976. 12. Kelly, R.D., Thompson, C.M., et al., "Westinghause Emergency Core Cooling System Evaluation !bdel for Analyzing Large LOCA's During Operation with One Loop Out of Service for Plants without Loop Isolation Valves",WCAP-9166, February,1978. 13. Eicheldinger,C., " Westinghouse ECCS Evaluation Model, February 1978 Version", WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version), February, 1978. 14. Letter from T.M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TfM-1981, Nov.1, 1978. 15. Letter from T.M. Anderson of Westinghouse Electric Corporation to Tedesco of the Nuclear Regulatory Commission, letter number NS-TMA-2014, Dec.,11,1978.

TABLE 1 - ~ ~ w LARGE BREAK - TIME SEQUENCE OF EVENTS EVENT OCCURRENCE TIME (SECONDS) DECLG, CD= 1.0 DECLG,CD= 0.8 DECLG, CD = 0.6 Accident Initiation 0.0 0.0 0.0 Reactor Trip Si9nal 0.665 0.668 0.673 Safety Injection Signal Og 0.48 0.53 Start Accumulator Injection 14.0 14.4 16.6 End of ECC Bypass 26.701 27.443 27.835 End of Blowdown 29.532 29.499 32.493 40.449 41.306 41.24 Bottom of Core Recovery Accumulators Empty 49.637 50.01 52.49 Start Pumped ECC Injection 25.44 25.48 25.53 e

TABLE 2 ~ ~ LARGE BREAK - ANALYSIS INPUT AND RESULTS Quantities in the Calculations: Licensed Core Power Rating 102% of 3250 MWt 1.93 MWt Total Core Peaking Factor Peak Linear Power 102% of 13.086 kw/ft Accumulator Water Volume 818.65 cubic feet per tank 615 PSIA Accumulator Pressure 3 Humber of Safety Injection Pumps Operating 1 percent (uniform) Steam Generator Tube Plugging Level Fuel Parameters - Cycle 1 Reg.on 1 Results 0.8 DECLG, CD" 10 DECLG, C DECLG, C = = D D Peak Clad Temperature (*F) 2000.89 2156.44 1974.46 Location (feet) 7.5 5.75 7.5 Maximum Local Clad / Water Reaction (%) 4.73 6.71 3.29 6.25 5.75 7.75 Location (feet) Total Core Clad / Water Reaction (%) <0.3 <0.3 <0.3 Hot Rod Burst Time (seconds) 3.14 30.00 34.4 Location (feet) 6.25 5.75 5.75

TABLE 3 CONTAINMENT DATA 6 3 NET FREE VOLUME 2.736 x 10 ft INITIAL CONDITIONS Pressure 14.7 psia Temperature 900F RWST Temperature 62 F Service Water Temperature 330F 0 Outside Temperature -10 F SPRAY SYSTEM r Number of Pump ~s Operating 3 Runaut Flow Rate 3600 gpm/each Actuation Time 18 see SAFEGUARDS FAN COOLERS Number of Fan Coolers 5 Fastest Post-Accident Initation of Fan Coolers 38 see STRUCTURAL HEAT SINKS 2 Thickness (in) Area (ft ) .25 steel,12 concrete;.004 pdnt 54447 .25 steel,12 concrete;.004 pa int 15026 18 concrete 15500 .25 steel,12 concrete 2000 12 concrete 36000 9 concrete 7000 .25 steel,12 concrete 16000 .25 steel 54860 .375 steel;.004 pa int 89300 0.6249 steel 1060 5.25 steel, 12 concrete 1147 .64 steel,12 concrete 1400 10.51 steel, 12 concrete 186 24.25 steel,12 concrete 54 .75 steel,12 concrete 4?0 7.287 steel,12 concrete 603.94 12.0308 steel,12 concrete 180.93 0.25 steel, 12 concrete 14862. 0.25 steel,12 concrete 3712. 0.375 steel 32000.0

TABLE 4 Reflood Mass and Energy Release to the Containment DECLG BREAK CD = 0.8 Time Mass Flow Energy Release, (sec) (LBM/sec) (BTU /see x 10 7 41.306 0.0 0.0 42.106 6.727 0.868 47.434 42.02 5.4226 55.003 126.75 15.497 ~ 67.003 148.23 16.5844 81.003 362.4 22.6632 96.103 400.72 23.0909 112.203 409.57 22.5839 148.003 423.73 21.28874 188.803 436.8 19.87726 f 236.703 450.61 18.3687 297.903 470.66 16.7054 O

~ TABLE 5 BROKEli LOOP ACCUMULATOR MASS Afl0 EilERGY RELEASES TO THE C0flTAlfiMEflT DECLG BREAK C0 = 0.8 TIME MASS FLOW EllERGY RELEASE (sec) (1bm/sec) (BTU /sec) 0.0 2838.6 -169239.5 1.0 2666.8 158995.8 2.0 2523.9 150479.2 3.0 2402.1 143216.6 4.0 2297.0 136948.7 5.0 2204.6 131443.4 6.0 2122.6 126552.1 7.0 2048.8 122153.4 8.0 1981.6 118148.8 9.0 1920.1 114476.6 10.0 1863.2 111086.9 11.0 1810.6 107951.2 12.0 1761.8 105042.1 13.0 1716.3 102327.0 14.0 1674.1 99813.1 15.0 1634.9 97475.4 16.0 1598.1 95281.7 17.0 1563.6 93224.6 18.0 1531.6 91318.0 19.0 1502.3 89571.9 20.0 1474.9 87935.7 21.0 1449.0 86392.4 22.0 1424.7 84945.6 23.0 1401.7 83575.0 24.0 '1379.9 82274.6 25.0 1359.2 81036.6 26.0 1532.9 85672.4

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