ML19269C781

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Forwards Proposed Tech Spec Changes to Conform to ECCS Reanalysis & Revised LOCA Analysis for Mod of Accumulator Vols to Appropriate Levels.Amend Is Combined Class 3 & 1 Amend.Appropriate Fee Remittance Encl
ML19269C781
Person / Time
Site: Zion  File:ZionSolutions icon.png
Issue date: 02/02/1979
From: Reed C
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 7902120191
Download: ML19269C781 (90)
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Text

{{#Wiki_filter:' Corr monwealth Edison Y. One first National Plaza. Chicago, Illinois / ~ Address Reply to: Post Office Box 767 / Chicago Illinois 60690 February 2, 1979 THIS DOCUMENT CONTAINS POOR QUAllTY PAGES Thl3 000m.E C0h.C43 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Zion Station Units 1 and 2 Proposed Change to Facility Operating License Nos. DPR-39 and DPR-48 NRC Docket Nos. 50-295 and 50-304 References (a) : April 2', 1978 letter from A. Schwencer to Cordell Reed containing Order for Modification of License (b): January 29, 1979 letter from C. Reed to Harold R. Denton titled " Zion Station Units 1 and 2 Revised ECCS Evaluation"

Dear Mr. Denton:

Pursuant to Reference (a) and 10 CFR 50.59, Commonwealth Edison Company hereby request a change to Operating License Nos. DPR-39 and DPR-48,. Appendix A, Technical Specifications. The purpose of this amendm0nt is to revise the Zion Technical Specifications to conform to the ECCS rc?.nalysis required by Reference (a) and performed in accordance with the Westinghouse ECCS Evaluation Model approved by the Staff in August 1978 and transmitted to the NRC Staff per Reference (a). The proposed change to the Zion Station Technical Specifications is enclosed at Attachment 1. contains the revised LOCA analysis for Zion Units 1 and 2. The reviaad LOCA analysis 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, Cp = 0.8) is 2174.180F with an Fg peaking factor limit of 1.86. As indicated in Reference (b), although the current accumulator volumes differ from the reanalysis value, these differences are more than compensated for by the additional margins stated in Reference (b). After final approval of the reevaluation contained herein, Commonwealth Edison intends to modify the accumulator volumes to the appropriate levels at the first outage of sufficient duration to accommodate the change. 0 7'16 4 / W W g4

1 p Conunonwealth Edison NRC Docket Nos. 50-295/304 Mr. Harold R. Denton: February 2, 1979 The technical specification changes of Attachment 1 also include Limiting Conditions for Operation, Surveillance Requirements and Bases for base load opcration without the use of APDMS type surveillance. For base load operation, the additional restriction of limiting the dI band to +3% AI was incorporated for the reactor operator to narrow the rarae of power distributions and thus ensure that thc Fg(Z) core peaking factor will always be less than the limiting Fg(Z) value. ";he technical specification change also incorporates the parameter / (defined as 10.8% of the APDMS turn-on power level, P ), which T is used to modify the " outer wings" of the41 band to ensure that margin to the LOCA limit is always maintained. The 10.8% value is a conservacive application of the 11% value currently in use. The above considerations, proposed technical specification changes (Attachment 1) and ECCS Revised LOCA Analysis have been revicacd and approved by Commonwealth Edison On-Site and Off-Site Review with the conclusion that there are no unreviewed safety questions. Pursuant to 10 CFR 170, Commonwealth Fdison has determined that this proposed amendment is a combined Class III and Class I Amendment. As such, Commonwealth Edison has enclosed a fee re-mittance in the amount of $4,400.00 for this proposed mmendment. Commonwealth Edison has concluded that the proposed amendment change does not involve a significant hazards consideration since the calculations made in support of this anendment are consistent with well defined and established analysis methods that have received previous NRC Staff approval. Please address any additiona) Tuestions that you might have to this office. Three (~) signed originals and thirty-seven (37) copies of this letter are provided for your use. Very truly yours, C.k9er Cordell Reed Assistant Vice-President attachments (2) SUBSCRIBED and SWO to before m9 this bl , day of _ hf l id / / A lt i 1979. hn$ni311 IY M CAVD Nyhary Public kJ

e ATTACH'iENT 3 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 47 67 174 45a 47A 68a 46 63a 69a The following pages have been added: 46a 69b 46b 69c

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIRalENT 3.2.2 Power Distribution Limits 4.2.2 Power Distribution A. Hot Channel Factor Limits

  • A.

Hot Channel Factor Limits 1.1 At all times, except during physics tests at $ 75% rated power **, the 1.1 Following initial core loading hot channel factors defined in the and at a minimum of regular bases must meet the following limits: effective full power monthly intervals thereafter, power Units 1 and 2 distribution maps, using the movable detector system, shall N = f(r be made to confirm that the hot g(Z)f l.86/P x Ky(Z), for P>.5 F Fg channel factor limits of this L 3.72 x K1(Z), for P f.5 specification are satisfied and F fl.55 1+0. 2 (1-P) XRBP, H Following initial loading and ~ where: each subsequent reloading, a power distribution map using Fg(Z) =Fg(Z) limit; the Movable Detector System, L shall be made to confirm that power distribution limits are 1.86 = F constant (LOCA limiting value)) q met, in the full power con-P = fraction of rated power at which figuration before a unit is operated above 75% of rating. the core operated during FQ andFjH measurement) K (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 exporure levels for each fuel region as referenced in the bases.
    • During physics tests which may exceed these het channel factor limits, the reactor may be in this condition for a period of time not to exceed eight hours continuously.

LIMITING CONDITION FOR OP,ERATION SURVEILLANCE REQUIREMENT 3.2.2.A.l.1 (RodBowPenalty)=(1-{hH Penalty), RBP andFfH Penalty is obtained from Figure 3.2-6 as a function of fuel region average burnup in MWD /MTU. The measurement of total peaking g Meas, shall be increased by factor, F three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error. The measurement of enthalpy rise hot channel factor, ((H, shall be increased by four percent to account for measure-ment errors. 1.2 If the measured hot channel factors exceed the limits in Item 3.2.2.A.1.1 of this specification, the reactor power and the high neutron flux trip setpoints shall be reduced in direct proportion to the excess over the peaking factor which is limiting for that unit. - 45a -

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQdIREMENT 3.2.2.A.l.2 (Con't) 4.2.2.A.1.2 If subsequent in-core mapping cannot, 2.1.a within a 48 hour period, demonstrate that the hot channel factor limits This APDMS type surveillance is are met, the reactor shall be brought conducted with at least two incore to the hot shutdown condition with return thimbles at the following frequency: to power authorized only for the purpose of physics testing. (1) At least once per 10 hours, if above P

  • T (2) Immediately snd every two hours If the turn-on power fraction, P,

for ten hours following the (defined in the bases) islesst3an events listed below: 1.00, then APDMS type surveillance must beconductedtomeet[F3 (z)]g, (de fined the APDMS (a) Raising the power, if above type surveillance set-point, P in the bases) while the power is above PT, except when surveillance is being done for BASE LOAD operation (Section (b) Moving the control bank of 3. 2. 2. A. 2. 2.). rods more than an accumulated total of 5 steps in any one 2.1.b direction,.if above PT* If the conditions of Section 2.1.b 3.2.2.A.2.1.a cannot be met, then immediately take action (and reduce Conduct APDMS type surveillance with power) asnecessarytoensure[F3(z)], at least two incore thimbles is met or reduco power to below immediately after taking action to P

ce-establish F (z)6 Fj (z),.

T. l i LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.2.A. 4.2.2.A. 2.2.a 2.2.a g I BASE LOAD operation can be used at powers Not Applicable _between PT and the power limited by (z] or1.00 (whichever is lower). BASE a F ~ BAD bperation can replace operation under L APDMS type surveillance only if Section i

3. 2.2. A.2.2. b is satisfied.

2.2.b 2.2.b l Prior to going to BASE LOAD operation and After waiting t'.te 24 hours, analyze a, prior to discontinuing APDMS type full core flux map near the power level surveillance (section 3.2.2.A.2.1), maintain that is limited by APDMS type surveillance. the following conditions for 24 hours: Prior to goirg above the power level limited by APDks type surveillance, (1) Power must be maintained between determine from this flux map the power P and the power limited by APDMS limited by _g(z) for BASE LOAD operation. F T type surveillanc(.. (2) oI wi, thin a 13% 4 I targe band. 2.2.c 2.2.c If the conditions of Section 3.2.1.A.2.2.b For BASE LOAD operation, the following are satisfied, then BASE LOAD operation may surveillance requirements shall apply: replace operation with APDMS type surveillance provided the following is (1) At less than or equal to the power maintained: level that was determined to be limited by[Fg(z] for BASE LOAD operation, obtaih a full core flux map at least one week, two weeks and' monthly after achieving BASE LOAD operation. - 46a -

~ LIMITING CONDITION FOR OPER ATION SURVEILLANCE REQUIREMENT 3 2.2.A 4.2.2.A 2.2.c (Continued) 2.2.c. (1) Power between and the power (2) A flux difference alarm shall limited by @q -)]n or 1.00 indicate non-conformance with (whichever is lower). the +3% AI target band for BASH" LOAD operation. If the (2) AI within the AI target band alarm is temporarily out of as per section 3.2.2.A.4 and service, conformance with the 3.2.2. A.5, except use +3% AI applicable limit and the flux target band instead of the +6, difference shall be logged -7% AI target band. hourly for the first 24 hours and half-hourly thereafter. 2.2.d If any of the requirements of section 2.2.d Not Applicable 3.2.2.A.2.2.c. are not maintained then power must immediately be reduced to below the power limited by AFDMS type surveillance (section 3.2.2.A.2.1.) and APDMS type surveillance must be ini-tiated if the power is above P

  • T 3.2.2.A.3 4.2.2.A.3 The target flux difference at a given The reference equilibrium indicated power level, Po, is determined by noting axial Clux difference as a function of the indicated axial flux difference at the power level (called the target flux power level with equilibrium xenon condi-difference) shall be determined at least tions established in the core, part once per equivalent full power quarter.

length rods fully withdrawn, and with the The target difference must be updated full length rod bank more than 190 steps every effective full power month. This withdrawn. Po for the purpose of deter-may be done using the measured value for mining the target value, should be as that month or by linear extrapolation high a power level as practicable. The using the two mo73 recent measured va-target flux difference at any other level, lues. The initial target flux difference P, is equal to the target value of PO on a reload may be determined from design multiplied by the ratio, P/P. predictions. O -46b-

LIMITING CONuITION FOR OPERATION SURVEILLANCE REQUIREMENT 4*

  • ^

3 2.2. A.4 4. Except during physics tests, during 4. Not applicable excore calibration procedures and except as modified by Itemo 5 through 7, the indicated axial flux difference shall be maintained within +6, (7% of the target flux l difference this defines the SI target band on axial flux differ-ence). 5. At a power level greater than 90 5. A f1 tx difference alarm shall indicate non-l percent of P, l conformance with the AI target band around the T target value for operation at power levels 51 If the indicated axial flux l above 90% of P. If the alarm is temporarily T difference deviates from its out of service, conformance with the applica-target band, either the devia-ble limit and the flux difference shall be tion shall be eliminated or the 1cgged hourly for the first 24 hours and reactor power shall be imme-half-hourly thereafter. diately reduced to a level no l greater than 90 percent of PT and Item 6 applies. l 6. At a power level no greater than 90 6. A flux difference alarm shall indicate non-percent of P, conformance with the limit on time (one hour T penalty time in 24) that the 6I target 6.1 The indicated axial flux band m be exceeded for operation at or be-difference may deviate from its low 9 of P. If this alarm is temporarily T l liI target band for a maximum out of service, conformance with the applica-penalty time of 1 hr (cumula-ble limit and the flux difference shall be tive) in any 24 hour period logged hourly for the first 24 hours and half-provided the flux difference hourly thereafter. does not exceed an envelope C LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3 ,2.A.6 bounded by + T percent and - T percent 4.2.2.A at a power of 90% of P and increa-T sing by +1 percent and -1 percent for each 2 percent of rated power below 90% of P. Whereif = 10.8% of PT T rounded down to the nearest percent. 6.2 If item 6.1 is violated then the reactor power shall be reduced to no greater than 50% power and the high neutron flux setpoint reduced to no greater than 55% of rated values. 7. At a power level no greater than 50 per-7. Not applicable. cent of rated power. 7.1 The indicated axial flux difference may deviate from its target band. 7.2 A power increase to a level greater than 50 percent of rated power is contingent upon the indicated axial flux difference not being outside its target band for more tisan one hour (cumulative) out of the pre-ceding 24 hour period. -47A-

Figure 3.2-9 Hot Channel Factor Normalized Operating Envelope for Unita 1 and 2 F Constant (LOCA Limiting Value) = 1.86 q ~ (6.0,1.000) 1.0 (11.4,0.932) .8 (12.0,0.763) cr .6 E> B 2 e .4 I E l i x .2 0 o 2 4 6 8 10 12 Core Height (Feet) - 63a -

Fg(Z), Height Dependent Heat Flux Hot Channel ItshouldbenotedthatFfH, is based on Factor, is defined as the maximum local heat integral and is used as such in the DNB flux on the surface of a fuel rod at core calculations. Local heat fluxes are obtained elevation Z divided by the average fuel rod by using hot channel and adjacent channel heat flux, allowing for manufacturing explicit power shapes which take into account ~ tolerances on fuel pellets and rods. variations in horizontal (x-y) power shapes throughout the core. Thus, the horizontal Fg, Nuclear Heat Flux Hot Channel Factor, is power shape at the point of maximum heat flux defined as the maximum local fuel rod linear is not necessarily directly related to F H. power density divided by the average fuel rod linear power density, assuming nominal fuel An upper bound envelope, Fg(Z) limit, has been pellet and rod dimensions, determined from extensive analyses considering E 11 operating maneuvers consistent with the Fg, Engineering Heat Flux Hot Channel Factor, technical specifications on power distribution is defined as the allowance on heat flux re-control as given in Section 3.2.2. The results quired for manufacturing tolerances. The of the loss of coolant accident analyses based engineering factor allows for local variations on this upper bound envelope indicates the peak in enrichment, pellet density and diameter, clad temperature will not exceed the 2200 F 0 surface area of the fuel rod and eccentricity limit. The Fg (Z) limit includes the higher upper of the gap between pellet and clad. Combined head temperature considerations (12). statistically, the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux. FfH,NuclearEnthalpyRiseHotChannelFactor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. The procedures for axial power distribution The alarms provided are derived from the plant control referred to above are designed to minimize process computer which determined the one the effects of xenon redistribution on the axial minute averages of the operable excore detector - power distribution during loadfollow maneuvers. outputs to monitor dgI in the reactor core and Basically control of flux difference is required alerts, the operator when AI alarm conditions to limit the difference between the current value exist. Two types of alarm messages are output. of Flux Difference ( 6I) and a reference value Above a preset power level, an alarm which corresponds to the full power equilibrium message is output immediately upon detennining value of Axial offset ( Axial Offset = 6I/ fractional a delta flux exceeding a preset band about a power). The reference value of flux difference target delta flux value. Below this preset varies with power level and burnup but expressed power level, an alarm message is output if the as axial offset it varies only with burnup. 6I exceeded its allowable limits for a preset cumulative amount of time in the past 24 hours. lThetechnicalspecificationsonpowerdistribution For periods during which the alarm on flux control assure that the Pq limit is not exceeded and difference is inoperable, manual surveillance xenon distributions are not developed which at a will be utilized to provide adequate warning of later time, would cause greater local power peaking significant variations in expected flux even though the flux difference is then within the differences. However every attempt should be limits specified by the procedure. made to restore the alarm to an operable condition as soon as possible. An deviations The target (or reference) value of flu 7 dirterence from the target band during manual logging shall is determined as follows. At any time '"St be treated as deviations during the entire equilibrium xenon conditions have been escablished, preceeding logging interval and appropriate the indicated flux difference is noted with part actions shall be taken. This action is necessary length rods withdrawn from the core and with the to satisfy NRC requirements; however more full length rod control rod bank more than 190 frequent readings may be logged to minimize the steps withdrawn (i.e. normal full power operating penalty associated with a deviation from the position appropriate for the time in life, target band to justify continued operation usually withdrawn farther as burnup proceeds). at the current power. This value, divided by the fraction of full power at which the core was operating is the full power The times that deviations from the band occur value of the target flux difference. Values for are normally accumulated by the computer. all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector of the 6 I target band are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring target flux difference overy month. For this reason, the specification provides two methods for updating the target flux difference. -68a-

significantly different from those resulting For Condition II events the core is protected from operation within the target band. The from overpower and a minimum DNBR of 1 30 by instantaneous consequences of being outside the an automatic protection system. Compliance ~ band, provided rod insertior. limits are observed, with operating procedures is assumed as a is not worse than a 10 percent increment in precondition for Condition II transients, peaking factor for flux difference in the range however, operator error and equipment mal-Hicated) percent (+ Y percent to - Y percent in- + ( Y +3 ) functions are separately assumed to lead to increasing by +1 percent for each 2 the cause of the transients considered. percent decrease in rated power. Therefore, while the deviation exists the power level is limited lto90%ofPT or lower depending on the indicated flux difference. If, for any reason, flux difference is not con-ltrolledwithinthe AI target band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 per-cent is required to protect against potentially more severe consequences of some accidents. As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to the equilibrium full power condition as possible. This is accomplished, without part length rods, by using the boron system to position the full length control rods to produce the re-quired indicated flux difference. It is accom-plished, when part length control rods are used, by using the boron system to position the full length control rods in a desired range and simultaneously using the part length bank to con-trol flux difference. The difference between these two methods is in the rate at which a return to full power can be accomplished when there is an increase in plant power demand. -69a-

Above the TURN-ON power fraction, PT, additional c. R 1 n R) j= 1 axial power distribution monitoring system type n 1:1 (APDMS type) surveillance is required except during Ej,least n=6 incor.e flux maps repre-Base Load Operation. APDMS type surveillance in-for thimble j, is determined from volves measurement of normali?ed axial power dis-at tributions, Fj(Z), for thimble j using movable sentative of the full range of the aI incore instrumentation. This surveillance involve 6 target Band Operations. Most recent ensuring or taking actiort to gnsure that Fj(7.) is flux maps will be used to update Kj ~ j (Z}} so that either by making n> 6 or by replacing F less tha its.setpoint, P (Z) $ g,F (Z[ L. s, Q LQ By lim 1 ting the core average the old flux maps with the recent maps. axial power distribution the total power peaking factor FQ(Z) can be limited since all other com-meas p ponents remain relatively fixed. The remaining Qi part of the total power peaking factor can be Rij=hijl2 1AX derived based on incore measurements i.e. an effective radial peaking factor, R, can be deter-and F j(Z) is the normalized axial mined as the ratio of the total peaking factor i results from a full core flux map and the axial distr 1Dution at elevation Z from peaking factor in a selected thimble. thimble j in map i which had a measured peaking factor without uncertainties Of

eas, Equations, definitions, and Explanations:

F"qi g 1 1 '7 1 (Fq(Zh L 2 1"1 I J-Rij), (Ej)(1 + F )(1.03)(1.07) $ = ""~1 8 ~ j RJ If APD/S eg,uipment is used for this surveillance, then LPj(Zy a will be further reduced by 15% to is standard deviation of R3 and is account for equipment round off errors. J derived from n flux maps or 0.02, which-ever is greater. The main criteria for b. PT = Turn-On power fraction selecting thimbles to be used in APDMS type surveillance will be based on Fq constant (LOCA limiting value) selecting those witn the smallest 0" ThevalueofFjwillalsobeusedds. a maximum Fq constant from 4-loop design guide to evaluate when older flux map predictions of swing-and-base load operation data is no longer adequate. 1.03=Ph e. f. 1.07 is the sum of two factors, 5% measurement uncertainty associated with the use of the incore flux measurement system and 2% additional conservatism for APDMS type sur- -69b-veillance technique.

Bases: Base load operation involves restricting operation to a narrow range of power distributions to en-sure that F9 (z)f[F This will be verified by(z)] ki core flux fu map surveillances at the intervals identified in Section 3.2.2.A.2.2. The power level at which [F IZI) L will be reached will be O determined in a conservative manner from full core flux maps taken at lower powers. Also, oper-ation near the power level at which [F (z)] first occurs will be just-191ed by full core flux maps. 69c

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.8 5 Accumulator System 4.8 5 Accumulator System (Table 4.8-4) A. The four accumulator systems shall satisfy the following conditions A. Surveillance and testing of the whenever the reactor coolant system accumulator system shall be pressure exceeds 1000 psig except performed as follows: as specified in 3.8.5.A.5 1. Each accumulator shall be 1. The pressure 2nd level of the pressurized to at least 600 accumulator tanks shall be psig and ghall contain minimum checked once a shift. At each of 770 ft3 and a maximum of 5% increase in level, the 818 ft3 of water. boron concentration will be checked. 2. Each accumulator shall contain 2. The accumulator boron concen-water borated to at least tration shall be checked 2000 ppm. monthly. 3. Each accumulator's isolation 3. The accumulator check valve valve shall be open, operability (SI-8948A, B, C, and D and SI-8956A, B, C,and D) will be verified at each refueling outage by opening the accumulator outlet isola-tion valve (MOV-SI8808A, B, C, and D) and verifyinE a de-crease in accumulator level and by a leakage test to de-termine that no gross valve leakage is experienced. 4 Isolation Valves Not Applicable. -174-

6 ATTACIBtENT 2 ZION STATION UNITS 1 AND 2 NRC DOCKET NOS. 50-295 AND 50-304 ) ECCS REVISED LOCA ANALYSIS

LOCA REAflALYSIS The Loss of Coolant Accident (LOCA) has been re-analyzed 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 LOCA analysis is given in WCAP-8339[2] The individual computer codes which comprise the Westinghouse Emergency Core Cool-ing System (ECCS) evaluation model are described in detail in separate reports [3-6] along with code modifications specified in references 7,10, and 11. The anal-ysis presented here was performed with the February 1978 version of the evalua-tion model which includes modifications delineated in references 12, 13, 14, and 15. Resul ts The analysis of the loss of coolant accident is performed at 102 percent of the licensed core power rating. The peak linear power and total core power used in the analysis are given in Table 2. Table 1 presents the occurence time for various events throughout the accident transient. Table 2 presents selected input values and results from the hot fuel rod therm-al transient calculation. For these results, the hot spot is defined as the location of maximum peak clad temperatures. That location is specified in Table 2 for each break analyzed. The location is indicated in feet, which pre-sents elevation above the bottom of the active fuel stcck. Table 3 presents a surmiary of the various containment systems parameters and structural parameters which were used as input to the C0C0 computer code [6] used in this analysis.

't Tables 4 and 5 present reflood mass and energy releases to the containment, and the broken loop accumulator mass and energy release to the containment, respect-ively. The results of several sensitivity studies are reported [8] These results are for conditions which are not limiting in nature and hence are reported on a generic basis. Figures 1 through 17 present the transients for the principal parameters for the break sizes 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 break. The core pressure drop is taken as the pressure just before the 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. Figures 10 - 11: Downcomer and core water level during reflood, and flooding ra te. Figure' 12 - 13: Emergency core cooling system flowrates, for both accumulator and pumped safety injection. Figures 14 - 15: Containment pressure and core power transients. ~ Figures 16 - 17: Break energy release during blowdown and the contain-ment 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 the Acceptance Criteria as presented in 10CFR50.46 El3. That i;: 0 1. The calculated peak clad temperature does not exceed 2200 F based on a total core peaking factor of 1.86 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 quenching. 4. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

References for Section 15.4.1 1. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors",10CFR30.46 and Appendix X of 10CFR50.46. Federal Register, Volume 39, Number 3, January 4,1974. 2. Bordelon, F.M., Massie, H.W., and Zordan, T.A., " Westinghouse ECCS Evaluation Fbdel-Sumary", WCAP-8339, July 1974. 3. Bordelon, F.M., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant", WCAP-8302 (Proprietary Version), WCAP-8306 (Non-Proprietary Version), June 1974. 4. Bordelon, F.M., et al., "LOCTA-IV Program: Loss-of-coolant Transient Analysis", WCAP-8301 (Proprietary Version), WCAP-8305 (Non-Proprietary Version), June 1974. 5. Kelly, R.D., et al., " Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code) ". WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Prcprietary Version), June 1974. 6. Bordelon, F.M., and Murphy E.T., " Containment Pressure Analysis Code (C0CO)",WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), June 1974. 7. Bordelon, F.M., et al., "The Westinghouse ECCS Evaluation Model: Supple-mentary Information",WCAP-8471 (Proprietary Version),WCAP-8472 (Non-Proprietary Version), January 1975. 8. Salvatori,R., " Westinghouse ECCS - Plant Sensitivity Studies", WCAP-8340 (Proprietary Version), WCAP-8356 (Non-Proprietary Version), July 1974. 9. Deleted

10. " Westinghouse ECCS Evaluation Model, October,1975 Versions",WCAP-8622 (Proprietary Version), WCAP-8623 (Non-Proprieta'y Version), November,1975. 11. Letter from C. Eicheldinger of Westinghouse Electric Corporation to D.B. Vassalo of the Nuclear Regulatory Commission, letter number NS-CE-924, January 23, 1976. 12. Kelly, R.D., Thompson, C.M., et al., " Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation with One Loop Out of Service for Plants without Loop Isolatio: Valves",WCA)- 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 Comission, letter number NS-TMA-1981, Nov.1, 1978. 15. Letter from T.M. Anderson of Westinghouse Electric Corporation to Tedesco of the Nuclear Regulatory Comission, letter number NS-TMA-2014, Dec.,11,1978.

TABLE 1 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 Signal 0.665 0.668 0.673 Safety Injection Signal 0.44 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 Bottom of Core Recovery 40.449 41.306 41.24 Accumulators Empty 49.637 50.01 52.49 Start Pumped ECC Injection 25.44 25.48 25.53

TABLE 2 LARGE BREAK - ANALYSIS INPUT AND RESULTS Quantities in the calculations: Licensed core power rating 102% of 3250 MWt Total core peaking factor 1.86 Peak linear power 102% of 12.612 kw/ft Accumulator water volume 818.65 cubic feet per tank Accumulator pressure 615 PSIA Humber of Safety Injection Pumps Operating 3 Steam Generator Tube Plugging Level 1% percent (uniform) Fuel Parameters - Cycle 1 Region 1 Resul ts DECLG,$D " - I

  • S-DECLG, CD=

0.8 DECLG,CD= 0.6 Peak clad temperature (OF) 2012.55 2174.18 1960.57 Location (feet) 6.25 5.75 7.75 Maximum local clad / water reaction (%) /.85 6.98 J.122 Location (feet) 6.25 '5.75 7.75 Total core clad / water reaction (%) <0.3 < 0. 3 __ <0.3 Hot rod burst time (seconds) 31.4 30.0 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 Number of Pumps Operating 3 Runout Flow Rate 3600 gpm/each Actuation Time 18 see SAFEGUARDS FAN COOLERS Number of Fan Ccolers 5 Fastest Post-Accident Initation of Fan Coolers 38 sec blRUCTURAL HEAT SINKS 2 Thickness (in) Area (ft ) .25 steel,12 concrete;.004 pint 54447 .25 steel,12 concrete;.004 pa int 1502' 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 print 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 44 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 Energy Release @ Time Mass Flow (BTU /sec x 10-(sec) (LBM/sec) 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 236.703 450.61 18.3687 297.903 470.66 16.7054

TABLE 5 BROKEN LOOP ACCUMULATOR MASS AND ENERGY RELEASES TO TiiE CONTAINMENT DECLG BREAK CD = 0.8 IIME MASS FLOW ENERGY RELE.".SE (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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