ML18018B909

Forwards Results of Dec 1984 LOCA Reanalysis to Justify Raising Fq Limit (Total Peaking).Increased Limit Will Allow More Operating Flexibility & Increase Fuel Economy.Draft FSAR Pages to Be Included in Future Amend Also Encl

ML18018B909
Person / Time
Site:	Harris
Issue date:	05/07/1985
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NLS-85-111, NUDOCS 8505130339
Download: ML18018B909 (78)

Text

REGULA Y INFORMATION DISTRIBUTIO SYSTEM (RIDS)

~

'CCESSION NBR!8505130339 DOC ~ DATE! 85/05/07 NOTARIZED:- NO DOCKET pACXu.50>>400 AUTH BYNAME." Shearon Harris AUTHOR AFFILIATION Nuclear'ower Planti Unit>> li Carolina 05000400 ZIMMERMANiS,R, Carolina Powe~ 8 Light Co.

REC IP ~ NAMEI RECIPIENT AFF IL.IATION DENTONrH: R ~ Office of Nuclear Reactor Regulationr Director.

SUBJECT:

For war ds- r esul ts of Dec 1984LOCp r e~ analysis to Just) fy raising FQ ] imit (total peaking) Incr eased limit will allow

~

mor e> oper ating fl exabi ) i ty 8, incr ease fueil economy, Draf t:

FSAR pages to be included in future amend =also encl.

DISTRIBUTION CODE! BOglD COP lES RECK I VED I LTR 'NCL '

I ZE'!

TITLE<- Licensing Submittal PSAR/FSAR Amdts 8, Rel-ated Cor respondence NOTES:

REC I PI EN T'D COPIES' RECI P COPIES CODE/NAME~ TTR ENCL' IENT'D'ODE'/NAME'RR LTTR ENCL NRR/DL/ADL LB3>> BC 1 0 NRR LB3 LA 1 0 BUCKLEYrB 01 1 1 INTERNAL! ACRS 6 6 ADM/LFMB 0 EI.D/HDS1 1- 0 IE FILE 1 1E/OEPER/EPB 36- 1 1 IE/DQAYT/QAB21 1 1 NRR ROE'gM ~ L~ NRR/OE/AEAB 1 0 NRR/OE/CEB 1 1 NRR/DE/EHEB 1 NRR/DK/EQB 13>> 2 2 NRR/DE/GB 28 2 2' NRR/DE/MEB 18 1 1 NRR/OE/MTEB NRR/DE/SAB 1 1 2517'RR/DK/SGEB 1

NRR/DHFS/HFEB40 1 1 NRR/DHFS/LQB 1 NRR/OHFS/PSRB 1 1 32'RR/DL/SSPB 1 0 NRR/DSI/AEB .1- 1 NRR/DSI/ASB 1 1 10 26'RR/DSI/CPB 1 1 NRR/DSI/CSB 09 1 1 NRR/OS I/ICSB 1 NRR/DS1/METB 12 1 1 19 16'RR/DSI/PSB 1 NRR/DSI/RAB 22" 1 1>>

NRR/DS I/RSB 23- 1 1 1 RGN2 "3 1 0 EXTERNAL; BNL.(AHDTS ONLY) 1 1 DMB/DSS (AMDTS) 1 1 LPDR 03>> NRC PDR 02'NL-1 1 NSIC 05 1 GRUELrR 1 1 TOTAI. NUMBER OF COPIES REQUIRED:= LTTR 51 ENCL>> 43

)') ~

'i

,Il I II ,

I (8

f '(i I'l g", " f ( f > C t lil l li I '9

pl ii II -Ir 4 C Ii I ht )iIT>> 'f (.Itl

%fan

~ I ~

e'I It ll 'll C)Q "(W,l( ) f ) Wfi,'t f) I ~ f jQ )r ~ II J ECW z)Vff ittrti W X f W rl i

~ f Art II <) ( I ',i l'it "/ I t( 'lllllt, f pWW f It '( rt 7f I+ I C) II'h

~ > "g '

II

' ti It ri s 7 Cr I '( lit 'ir g 'll Wi'I~ it Itr l,Q ) ,ji,fW W'l <

"> f )

lit 'X II JJ rl r I, I (il 4 f~

Wg

'I

,i(ttlhh Ct ( th l

> WI'W f

)gt I' r Ii P

l l' 1

it f

i I,'(Hg', l ll C

ii t(, g(

) Ji l e i

'I lt Ir( p 1I II la

S!M 0 Carolina Power & Light Company SERIAL: NLS-85-111 MN 0 7 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. 1 - DOCKET NO.50-000 LOCA RE-ANALYSIS

REFERENCE:

April 23, 1985 letter from A. B. Cutter (CPRL) to H. R. Denton (NRC).

Dear'r. Denton:

Carolina Power R Light Company submits information (Attachment 1) to justify raising the F~ limit (total peaking) from 2.10 to 2.32 for the Shearon Harris'uclear Power Plant. A larger margin to the peak clad temperature limit of 2200'F is also demonstrated. The increased F~ limit will allow more operating flexibilityarid increase fuel economy to be achieved.

Draft FSAR pages are included as Attachment 2. These FSAR pages will be included in a future amendment. Technical Specification changes were included in the revised "pen and ink" Technical Specifications transmitted via the referenced letter.

Please review the'ttached information and revise the Safety Evaluation Report, as necessary. If you have any questions, please contact Mr. Gregg A. Sinders at (919) 836-8168.

Yours very truly, S. tmmerman Manager Nuclear Licensing Section GAS/ccc (1307GAS)

Attachments Cct Mr. B. C. Buckley (NRC) Mr. Wells Eddleman Mr. G. F. Maxwell (NRC-SHNPP) Mr. 3ohn D. Runkle Mr. H. Richings (NRC-CPB) Dr. Richard D. Wilson Dr. 3. Nelson Grace (NRC-RII) Mr. G. O. Bright (ASLB)

Mr. Travis Payne (KUDZU) Dr. 3. H. Carpenter (ASLB)

Mr. Daniel F. Read (CHANGE/ELP) Mr. 3. L. Kelley (ASLB)

Wake County Public Library 8505130339 850507 PDR ADOCK 05000400 A PDR 411 Fayettevilte Street o P. O. Box 1551 o Raleigh, N. C. 27602

ATTACHMENT 1 RESULTS OF THE DECEMBER 1980 LOCA RE-ANALYSIS FOR THE SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 BY WESTINGHOUSE ELECTRIC CORPORATION FOR CAROLINA POWER R LIGHT COMPANY Introduction Westinghouse Electric Corporation has developed improved analytical techniques which allow an increase in the F~ limit (total peaking), while still meeting the acceptance criteria for LOCA. This rePort provides information to substantiate an increase in the F~ limit for the Shearon Harris Nuclear Power Plant (SHNPP).

Descri tion of the Anal sis The large break LOCA ECCS analysis for SHNPP was performed at a power level of 2775 MWt, with a F& (Hot Channel Factor) of 1.55. The analysis utilized initial fuel conditions generated by the Revised Pad Thermal Safety Model (WCAP-8720, iAddendum 2). Pertinent analysis assumptions include a six percent steam generator tub~

plugging level, 17xl7 standard fuel, and an accumulator water level of 1,050 ft.

(including piping between the accumulator and check valves). The analysis was performed with a modified version of the 1981 Westinghouse ECCS evaluation model as described in WCAP-9220-P-A, Revision 1 (proprietary). This version includes the BASH computer code documented in WCAP-10266 (proprietary). At the present time, the BASH code has yet to be approved by the Nuclear Regulatory Commission. Also, Statistical Evaluation of LOCA Heat Source Uncertainty, WCAP-10395 (proprietary), was applied.

The break sizes analyzed in the study were the CD 0.0, 0.6, and 0.8 DECLG. The worst break is with C = 0.8, which resulted in a peak clad temperature of 1820'F at an F~

(total peaking Pactor) oi 2.32. The analysis result demonstrates conformance wit%

10 CFR 5086 requirements for a large break ECCS LOCA analysis for SHNPP.

Conclusion The above information justifies raising the F~ limit (total peaking) from 2.10 to 2.32.

(1347GAS/ccc )

S)KPP FSAR maneuvers are studied to determine the general behavior of the local power density as a function of core elevation.

These cases represent many possible reactor states in the life of one fuel cycle and they have been chosen as sufficiently definitive of the cycle by comparison with much more exhaustive studies performed on some 20 or 30 different, but typical, plant and fuel cycle combinations. The cases are described in detail in Reference 4.3.2-6, and they are considered to be necessary and sufficient to generate a local power density limit which, when increased by 5 percent for conservatism, vill not be exceeded with a 95 percent confidence level. Many of the points do not approach the limiting

( envelope, however they are part of the time histories which lead to the hundreds of shapes whi,ch do define the envelope. They also serve as a check that the reactor studied is typical of those studied more exhaustively.

Thus, it is not possible to single out any transient or steady state condition which defines the most limiting case. It is not even possible to separate out a small. number which form an adequate analysis. The process of generating a myriad of shapes is essential to the philosophy that leads to the required level of confidence. A maneuver which provides a limiting case for one reactor fuel cycle (defined as approaching the line of Figure 4.3.2-21), is not necessarily a limiting case for another reactor or fuel cycle with different control bank worths, eniichments, burnup, or coefficient. Each shape depends on the detailed history of operation up to that time and on the manner in which the operator conditioned xenon in the days immediately prior to the time at which the power distribution is calculated.

The calculated points are synthesized from axial calculations combined with radial factors appropriate for rodded and'nrodded planes in the first cycle.

In these calculations, the effects on the unrodded radial peak of xenon redistribution that occurs following the withdrawal of a control bank (or banks) from a rodded region is obtained from two-dimensional X-Y calculations.

A 1.03 factor to be applied on the unrodded radial peak was obtained from calculations in which xenon distribution was preconditioned by the presence of control rods and then allowed to redistribute for several hours. A detailed discussion of this effect may be found in Reference 4.3.2-6. %ma- u&ee~

~~ heve-I.a '-c R.V The envelope drawn over the calculated (max F Power) points in Figure 4.3.2-2l presents an upper bound envelope on )ocal po~er density versus elevation in the core. It should be emphasized that this envelope is a conservative representation of the bounding values of local power density.

Expected values are considerably smaller and, in fact, less conservative bounding values may be justified with additional analysis or surveillance requirements. For example, Figure 4.3.2-21 bounds both BOL and EOL conditions, but without consideration of radial power distribution flattening with burnup, i.e., both BOL and FOL points presume the same radial peaking factor. Inclusion of the burnup flattening effect would reduce the local po~er densities corresponding to EOL conditions which may be limiting at the higher core elevations.

4.3.2-10

SH:~FP FSAR Finally, as previously discussed, this upper bound envelope is based on procedures of load follow which require operatior. within an allowed deviation from a target equilibrium value of axial flux difference. The procedures are detailed in the Technical Specifications and are folloved by relying only upon excore surveillance suppLemented by the normal monthly full core map requirement and by computer"based alarms on deviation and time of deviatior.

r rom thr allowed flux difference band.

Allowing for fuel densification effects the average linear power at 2775 Mvt is 5.44 kW/ft. From Figure 4.3.2-21, the conservative upper bound value of normalized local pover density is 2.32 corresponding to a peak linear power of 12.9 kW/ft. at 102 percent power.

Accident analyses for SHNPP are preserted in Chapter 15 of the FSAR. The FQ, of ~

results of ghye analyses determined a limiting value of total peaking factor, un8er normal operation, including load following maneuvers'his v~lue is derived from the conditions necessary to satisfy the limiting condicions specified in the LOCA analyses of FSAR Section 15.6.5. As noted above, an upper bound envelope of FO x Power equal to 2.32 x K(Z), as shovn in FSAR Figure 4.3.2-21, results from operation in accordance with Constant Axial OfGset Control procedures using eQecore surveillance only.

~

Q x

'32 n

ys me port to be-issue~

rThe APDMS This system and normal is u

a rveillance ool to verify comp iance vith lizes a four section excore det ctor (which ed) to p. ovi on-line real-time onitoring o l mits F

of as been ca (Z).

F is Z)~

ibrated 0

system vides audible and visual alarms en a predet mined APDM alarm

<set po t and power di ribution are excee ed. Core s nning is i tiated by lexc ding a power r ge setting, by exc ding a control rod step dead band, or Immr ual ly.

To determine reactor protection system setpoints vith respect to power distributions, three categories of events are considered: namely rod control equipment malfunctions, operator errors of commission, and operator errors of omission. In evaluating the three categories of events, the core is assumed to hv operating withir. the four const.airts described above.

Th~ first category comprises uncontro'led rod withd.aval (with rods moving ir.

the normal bank sequence) for full length banks. Also included are motiors of the full length banks below their irsertior. Limits, vhich could be caused, for examp)e, by uncontrolled dilution or reactor coolant cooldown. Power disr.ributions vere calculated throughout the occurrences, assuming short term corrective action; that is, no trar.sient xenon effects vere considered to result from the malfur.ctior.. The event vas assumed to occur from typical normal operating situations vhich include norm'al xenon trarsierts It was furrher assumed ir. determining the power distributions that total core pover level would be limited by reactor trip to belov l)8 percent. Since the study is to determine protection limits vith respect to power and axial offset, no credit vas taken for trip setpoint reduction due to flux difference. Results are given in Figure 4.3.2-22 ir. units of kW/ft. The peak pover density vhich can. occur in such ever.ts, assuming reactor trip at or below 118 percent, 'is Amendment No. 14

SHNPP FSAF.

less than that required for center-line melt, including uncertainties and densificatior. effects.

The second category, also appearing in Figure 4.3.2-22, assumes that the operator mispositions the full length rod bank in violation of the insertion limits and creates short term conditions not included in normal operating conditions.

The third category assumes that the operator fails to take action to correct a flux difference violation. The results shown on Figure 4 '.2-23 are FQ multiplied by 102 percent power including an allowance for calorimetric error.

The figure shows that provided the assumed error in operation does not continue for a period which is long compared to the xenon time constant, the peak linear power does not exceed 18 kM/ft. including the above factors.

lt is nt ipated that future analyyes will permit a pise in thy F o lynitpo~ P 2.32 or eater. This rgse in the/'limit will eliminlte the nee'd to ~er~e h APD S Analyses of possible operating power shapes show that the appropriate hot channel factors F~ and F AH for peak local power density and for DNB analysis at full power are the vazues given in Table 4.3.2-2 and addressed in the Technical Specific'ations.

F< can be increased with decrgasing power as shown in the Technical Specifications. Increasing F AH with decreasing power is permitted by the DNB protection setpoints and allows radial power shape changes with rod insertion to the insegtion limits as dgscribed in Section 4.4.4.3. The allowance for increased Fpermitted is F QH 1.55 [1 + 0.2 (I-P)). This becomes a design basi5 ~~H criterion which Xs used for establishing acceptable control rod patterns and control bank sequencing Likewise fuel loading patterns for each cycle are selected with consideration of this design criterion. The worst values of F bH for possible rod configurations occurring in normal operation are used in verifying that this criterion is met. Typical radial factors and radial power distributions are shown in Figures 4.3.2-6 through 4.3.2-]1 ~ The worst values generally occur when the rods are assumed to be at their insertion limits. Haintenance of constart axial offset control establishes rod positions which are above the allowed rod insertion limits thus providing inc. eased margin to the F, criterion. As discussed in Sectior. 3.2 of Re:erence '.3.2-7, it has beer. deter ined that provided the above conditions are observed, the Techrical Specificatior. limits are met. These limits are taken as input to the thermal-hydraulic design basis as described in Section 4.4.4.3.].

1'hen a situation is possible in normal operation which could result in local power densities in excess of those assumed as the pre-condition for a subsequent hypothetical accident, but which would not itself cause fuel failure, administrative controls and alarms are provided for returning the core to a safe condition. These alarms are described in detail in Chapters 7 and 16.

4.3.2-12 Amendment No. ]4

SUPP FSAR Errors in the ca culated relationship between detector flux and peak ro" power some distance from the measurement thimble.

The appropriate allowance for Category a above has been quantified by repetitive measurements made with several intercalibrated detectors by using the common thimble features of the incore detector system. This system allows more than one detector to access any thimble. Errors in Category b above are quantified to the extent possible, by using the fluxes measured at one thimble location to predict fluxes at another location which is also measured. Local power distribution predictions are verified in critical experiments on arrays of rods with simulated guide thimbles, control rods, burnable poisons. These critical experiments provide quantification of errors of Categories a and c above.

Reference 4.3.2-1 describes critical experiments performed at the Westinghouse Reactor Evaluation Center and measurements taken on two Westinghouse plants with incore systems of the same type as used in the SHNpp. The report concludes that the uncertainty associated with Fo (heat flux) is 4. 58 percent at the 95 percent confidence level with only 5 percent of the measurements greater than the inferred val'ue. This is the equivalent of a 1.645s limit on a normal distribution and is the uncertainty to be associated with a full core flux map with movable detectors reduced with a reasonable .set of input data incorporating the influence of burnup on the radial power distribution. The uncertainty is usually rounded up to 5 percent. '%4M RT In comparing measured power distributions (or detector currents) against the ca].culations for the same si.tuation, it is not possible to subtract out the detector reproducibility. Thus, a comparison between measured and predicted power distributions has to include some measurement error. Such a comparison is given in Figure 4.3.2-24 for one of the maps used in Reference4.3.2-1 ~

Since the first publication of the repozt, hundreds of maps have been taken on these and other reactors. The results confirm the adeqgycy of the 5 percent uncertainty allowance on the calculated FO. ~ a~4~

A~SaSker analysis for the uncertainty in FN AH (rod integral power) measurements results in an allowance of 3.63 percent at the equivalent of a

].645s confidence level. For historical reasons, an 8 percent uncertainty factor is al'owed in the nuclear design calculational basis; that is, the predicted rod integrals at full power must not exceed the design F less.

8 percent. This 8 percent may be reduced in final design to 4 percent to allow a wider range of acceptable axial power distributions in the DM analys's anc still mee: the des'gn bases o Section 4.3. }.3.

A recent measurement in the second cycle of a ]21 assembly, 12 ft. core is compared with a simplif'ed one-dimensional core average axial calculation in Figure 4.3.2-25. This calculation does not give explicit representation to the fuel grids.

'4.3. 2-13 Amendment No. 2

4.3.2-11 Ford, W. E., III, et.al., "A 218-Group Neutron Cross-Section Library in the AMPX Master Interface Format for Criticality Safety Studies," ORNL/CSD/TM-4 (July, 1976).

4.3.2-12 Greene, N. M., et.al., "AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B,"

'ORNL/TM-3706 (March 1976).

4.3 '-13 Petrie, L. M. and Cross, N. F., "KENO IV An Improved Monte Carlo Criticality Program." ORNL-4938 (November, 1975).

4.3.2-14 Bierman, S. R., et.al., "Critical Separation Between Subcritical Clusters of 2.35 vt Z 235U Enriched U02 Rods in Water vith Fixed Neutron Poisons," Battelle Pacific Northvest Laboratories PNL-2438 (October, 1977) ~

4.3.2-15 Bierman, S. R., et.al, "Critical Separation Between Subcritical Clusters of 4.29 vt Z 235U Rods in Water with Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2615 (March, 1978).

4.3+2-16 Thomas, J. T., "Critical Three-Dimensional Arrays of U (93.2)~etal Cylinders," Nuclear Science and Engineering, Volume 52, pp. 350-359 (1973).

4.3.2-17 Poncelet, C. G. and Christie, A. M., "Xenon-Induced Spatial

. Instabilities in Large PWRs," WCAP-3680-20 (EURaEC-1974),

March, 1968 Skogen, F. B. and McFarlane, .A. F., "Control Procedures for

'.3.2-18 Xenon-Induced X-Y Instabilities in Large PWRs," WCAP-3680-21 (EURAEC-2111), February, 1969.

4.3.2-19 Skogen, F. B. and McFarlane, A. F., "Xenon-Induced Spatial Instabilities in Three"Dimensions," WCAP-3680-22 (EURAEC-2116),

September, 1969.

4 ' '-20 Lee, J. C., et.al., "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964, June, 1971.

4.3.2-21 Barry, R. F., et.al., "The PANDA Code," WCAP-7048-P-A (Proprietary) and WCAP-7757-A (Non-Proprietary), January, 1975.

4.3.2-22 England, T. R., "CINDER - A One-Point Depletion and Fission Product Program," WAPD-Hi"334, August, 1962.

4.3.2-23 Eggleston, F ~ T., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program

)

4.3.3-1 Summaries-Minter 1976", WCAP-8768, Revision 1, June, 197?.

Poncelet, C. G., "LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS," WCAP-6073, April, 1966.

I

INSERT A Based on the application of Reference 4.3.2-24 (i.e., Statistical Evaluation of LOCA Heat Source Uncertainty) the factor of 1.05 for conservatism and 1.03 E

for F Q

are not included in the calculation.

INSERT (B However, based on the implementation of Reference 4.3.2-24, this is included in the uncertainty analysis for determination of the limiting value of FQ and need not be included in the measured value of FQ.

INSERT C)

However, Reference 4.3.2-24 shows that the Westinghouse ECCS evaluation models already account for the nuclear uncertai'nties and hence, the nuclear uncertainties and conservatisms can be eliminated from the nuclear design calculations of F Q'.

Z

~ ~

SHNPP FSAR I

I 6~ ei 4 ~ Safe System Failures 1 I v

/

ai lur of fain Fe ed wa te r Line Is o 1 ation Valve .- There are wo valves rood ter i and booth are designed topless within 7 sec. ~

iter the Ler i. ol L on etpoint>'is reefed (2 sec. i trumentati n respon e time 1 dinge jd la a 5 sec.,valve closing time).

I i ure of/ ne va ve folloying a steam liqe brea /would increase the

~nisi/a)'le ee I inc/igided i Lhe mass water li s e by the I

olume between the two'valves. This effecL nd energ release hgafysis by increasing the

~

g isol Led feed Ler lin volume o 717.7 ft.~ '

I I I I

Fai iur of Lhe xi liary eedwater Pu Runout frotection , The motor g~v n a~ixili ry feed,t'er pump are'quippe with safety grade fZow

>ht ollkrs. The ass ed auxil ary feedwat r flowrates are based on runout 10 from ~t Lurbine driven a xiliary fee ater pump. Therefore, no further on iderati n need h given L'o this failur I e

/ /

) Fai ure of N 'n Feedwa er, Pump Tr l No credit is taken for feedwpter mp trip and coast own in calculating f edwater addition pribr'to edwat]r line iso tion. T erefore, t is failurg has no effect on ther esults:~ resented" I

/

) w I allure of a Steam ne Isolat)'on Valve q Failure of a main s am line snlal/I n valve i cress s be volume f steam which empties into the.d onta ment Lo 3 32 ft. The Conta nment is designed to Iaccept th s iddi t al Steam I

/

I

) j oniain at gafeguards Train, pailure ,pailure of a connai ment~

I fe rds Lra results in minimuar heat removal capability correkponding I to operant'aloof two containment fan coolers and one wontainment spray pump.

6e2. l.5 Hinfmum Containment. Pressure Anal sis for Performance Ca abilit Studies of Emer enc Core Coolin S stem 8

The containment backpressure used for the limiting case CD~0.4, DECLG break for Lhe ECCS analysis presented in Section 15.6.5 is presented in Figure 6.2.l-302. The containment backpressure is calculated using the methods and assumptions described in "Westinghouse Emergency Core Cooling

.System Evalulation Hodel Summary," WCAP-8339, Appendix A. Input parameters including the containment initial conditions, net free containment. volume, passive heat sink materials, thicknesses, surface areas, starting time, and number of containment heaL removal systems used in the analysis are described helow.

The anaylsis was performed assuming the loss of offsite power as the most limiting condition. As indicated in WCAP-8471, the three loop plant limiting case break (CD 0.4 DECLG) yields lower calculated PCT values with offsite power available (reacLor coolant pumps run case) than if offsite power is lost (reactor coolant pumps trip case). This results from core thermal hydraulics during hlowdown and is true even though calculated containment pressure may be lower in the offsiLe power available case due to faster actuation of the engineered safeguards. The applicability of the generic conclusion regarding offsite po~er status to the Shearon Harris ECCS analysis is presented in detail below.

SHNPP FSAR bm2. I ~ 5 2 In I t ia 1 Con t a i nme nt ln t e ma 1 Cond i t i ons The following initial values were used in the analysis:

Containment pressure 14.7 psia Containment temperature 90 F RWST temperature 40 F Service water temperature 40 F Outside temperature -2 F Initial Relative Humidity 100 X The initial temperature condition that may be encountered under limiting normal operating conditions used in the ECCS performance analysis was ..

as assume d An evaluation determined that the containment cannot Eall below 80 F and thhe normal expected average containment temperature es tima t e d at i00 I'. i'he 90 F value was chosen because it was shown to uu ve ue consistent vith representative normal full pover operation of other o bea sa conservat ively 1

nuclear plants. The normal operating range for containment pressure is expected to be between negative 1 inch wg to positive 4 inch wg with the nominal pressure expected to be slightly positive. The value of 14.7 psia. was assumed for the ECCS performance analysis. The containment is the atmospheric type per item d oE SRP 3.8. The normal containment purge and makeup systems alu>>g with the containment cooling system will maintain the containment within the normal operating range. The Normal Containment Purge Exhaust is first n< justed to allow the system to draw down the containment atmosphere to a slight negative pressure (to prevent outleakage). When the containment pressure is reduced to -0.25 in. wg, one of the two 100 percent capacity makeup fans will automatica'Ily start. The static pressure controller will rut;ulate th>> respective supply fan inlet damper to modulate and maintain the c>>ntainmf nt pressure setpoint. The pressure transmitter for controlling this lower valui has a range of 0 to negative 1.0 in. wg. Safety grade pressure transmitters with a range of -5 to 0 to +5 in. wg provide coverage of the norm~i expected containment pressure range to include the negative pressure transient requiring initiation of the containment vacuum relief system.

Without continuous purging the containment pressure will buildup due to equipment air (instrument air) leakage. Purging vill be intermittent and vill b'imited to n duration consistent with NRC requests, plant operational requirements, and ALARA considerations.

6.2. I.5.3 Containment Volume The volume used in the analysis is 2.344 x 10 ft.

6.2. I~ 5.4 Active Heat Sinks fhe Containment Spray System and the containment fan coolers operate to remove liest from the Containment.

as<ot

's i

a i

t nt y r s e e t o h r ta m on S al

SHNPP FSAR p c r 1 ( ~

Pert. inenl. dal.a for t.hese systems which were used in the analysis are presented tn Tahle 6.2.1-62. The heat. removal capability of each fan cooler is presented in Figure 6 ~ 2 ~ 1-303 containment sump temperature was not used in the analysis because the

'lt(,

(na~imum peak cladding temperat.ure occurs prior to initiation of the recirculat.ion mode for Containment Spray System. In addition, heat transfer bet.ween the sump water and the containment vapor space 'was not considered in t.hc analysis.

6.2.1.'5.5 Steam-Water Hixing Wal.er spillage rates from the broken loop accumulator are determined as part.

of t.he core ref looding calculation and are included in the containment. (COCO) code ca 1 c(tin t. ion model.

6.?.1.5.6 Passive Heat. Sinks t)r(expert.ies,

'fhe passive heat sinks used in t;he analysis, wit:h their t.hermophysical are given in Table 6 ~ 2 ~ 1-63 ~

Concrete t.hermophysical properties utilized were taken directly from BTP CSB 6-1 ~ A carbon steel thermal conductivity value of 26Btu/hr-ft-F is specified for t.he t.emperature range of interest for Shearon Harris from Reference 6.2.5-5; likewise, a volumetric heat capacit:y value is obtained from that reference. The values shown in Table 6.2.1-63 were used in the analysis.

6.2.1.5.7 Heat Transfer to Passive Heat Sinks The condensing heat transfer coefficients used for heat transfer t:o the st.'eel containment str(tct.ures is given in Figure 6.2.1-304 for the limi.ting break.

The containment t:emperature transient for the limiting break in shown in Figure 6.2.1-305.

6.2.1.5.8 Containment Purging During a LOCA An analysis was performed to determine the reduction in containment. pressure resulting from containment purging during a LOCA for ECCS backpressure det.erminat.ion. This analysis was based upon the containment conditions (le f ined using the 1978 Westinghouse Evaluation Hodel. A containment isolat.ion signal is received in that analysis at 1.03 seconds after inception of the LOCA. Adding 1.5 seconds for signal delay, a calculation is performed for a cont.ainment purge system consisting of two 8-inch diameter lines and the following conservative assumptions:

a) A 3.5 second isolation valve closure time is assumed. During the 6.03 s('.cond period immediately following t: he LOCA, no credit is taken for the reduction in effective flow area which occurs while the valve is in the process of closing.

SHNPP FSAR b) The frictional resistance associat:ed with duct entrance and exit.

losses, f ilt.'ers, ducLwork bends and skin friction has not. been considered.

c) No fan cnastdown effects are considered.

d) No inertia is considered. Steady state flow out. the purge system duct.s is esLablished immediately at the time of the LOCA.

~ A mixt.ure of steam and air wi11 be exhausted from the containment through the purge lines during t.he 6.03 seconds that t: he isolation valves are assumed to remain open. The effect of the composition of the gas being exhausted on containment pressure has been bounded by investigating t.he two extreme cases, air alone and st,earn alone. Mithin several seconds of the inception of. the LOCA, conLainment pressure will have increased t.o the point that critical flow wi11 occur in the purge lines. To bound the calculated gas mixt.ure exhausted Lhro<<gh t.he purge lines, the critical flow rat:es of steam and air were rale<<1aLed di)ring the first 6.03 seconds of the t.ransient.. Using t.hese flow rates, critical. flow was then conservatively

<<ssumed Lo be in effect from time zero. Equation 4.18 in Reference 6.2.1-13 was employed Lo calculate t.he critical flow rate of air through t: he purge lines. Fig<<re 14 of Reference 6.2 ~ 1-14 was applied to compute the crit. ical flow raLe of sLeam through Lhe purge lines. The total mass released during t.he 6.03 seconds that the valves are presumed open is calculated as 331 ibm air or 239 1bm st.earn. The impact on containment pressure at 6.03 seconds res<<lt.ing from t.his loss of air or steam is less than 0.05 psi in either case. The effect of a containment pressure reduction of this magnitude on the Therefore, there is no FQ penally and margin with respect t:o 10CFR50.46 PCT requirement:s would remain.

(i.2).1e S~ 9 OLher Parameters No ot.her parameters have a substantial effect on the minimum containment t)1 ess<<re analysis.

6~ 2.1 ~ 6 Test. in and Tns ection t

b ct.ural t:e rity sty and service rveilla ce requiWem ts are~

(i. 8. ttt'n ent-i3u 1 ing an /

io '3..2. 7 r s'e

t. he Cl ssafety co pone

". s enetr Sons, lo s, and h ches ) 'ont.ai t leak e testi g i dis ussed n Sectio 6.2.6. q, ti g a tinge raq XEALe l af em..and ineered oaf L) f alures haL af t;he func onal c bilit.y f e Contai me t are in Sect.i

'sc'usse n 6.2 an .6. P operati na ee ting e d enebbe] i

. ecLiqn 4 '%12 /

6.2.1.7, ns rumen tat io A ication t

Pressurd SenSing instrument monito,'the ontai e t

/

atmos here and kn iat,e, Lhe co ainmenL so ation, injection n co a ee ~pray act.uation signals accordin t:o the ltjgic discukqed in >Sec 4.on 7.3.: Radiation monitors mnnit,or cont.ainment atmosphe're and isblate c'ongainrt)e, t purge through the conLai.nment.-vent<k~t<on-+scgl kgnaZ ~s. discusse in Section 7.3. /

/

SHNPP FSAR TABLE 6 o 2. 1-59 SLOWDOWN MASS /ENERGY RELEASES DECLG C & 0. ~ 8 Time (sec.) Mass Flov lb. /sec. Ener Floe Btu/sec.

0.0 Q. O O. o

5. 7T. Q

~ 05 x 104 35&8< x 107 5 ~ 30k 3i II 4 2.0 8595 x 104

~~

+484- x 107 3.CXC 2ett~

4.0

~~

sh+934+N x 104 x 107 6.0

~i%7 $

x 104

/iCf3

~~444 x 107 8.0

/~

P V7t x 104 3

l~~

/ilZ5 7~ 8VS x 107 10m 0 7'4864 x 10~ %r&&&9897 x 106

12. 0

v. $ 7S 1~&rB x 10~~

5".

Ss+~~

leg x 106 g, F53

14. 0 ~Hr39i x 106,

g. o 'g2

16. 0 39994 x 106 3 3$ < I ~ /C 2.

18 ~ 0 3s&R~i& x 103 ~6&RR8 x 106 2,o/C 5'72 4

20. 0 ~<HE x 103 L44PP7-1 x 10~

7/. e9 S / 5i (35 RES'~ x 10$ 1 %4$ P84 x 10~

0~

SHNPP FSAR TABLE 6 ~ 2.1-60 REFLOOD MASS AND ENERGY RELEASE DECLG (C ~ Oe4)

Time (sec.) Mass Flow lb./sec. Ener Flow Btu/sec.

3Z e'9

~

%~9- 0. Oo 99 49 5'P< 1 Sfro AS c808$ 3 gQ. /o Sl,o7 4Am~ RQaaRQ A&8EH V8 Yo~ 2,o5'.9/ /. 73v /o ~

~554 .489C8 C ff'./a 2fP 7o / ~ S~ v/o+

2&gee g3. 1o 3/e 7/ / VEvro~

&H~ 4PRPFB

/o go 'c/o g2 ge7C ) QV )4/b

~

3eh~ ]esi4RW 3/ i /c7 gg oI5 / /. S rvro ~

~e584 3CS~ JeCI45IO'.

/E4 go ~ 3HZ l.V 8 g~/o BBR~

SF 3o /o /7X /0 QQ 3 ~

sHdhfh&4 Lo 3~ bSHRBCc

SHNPP FSAR TABLE 6 ~ 2i 1 61 BROKEN LOOP ACCUMULATOR MASS AND ENERGY RELEASES TO CONTAINMENT (C ~ 0.4)

Time (sec.) Mass Flov lb./sec. Ener Flov Btu/sec.

0. 000 3TXUZr3 yC7r /5 2$ 6S%9P3'35 Q7 P fco '37

1. 010 3~F7 V/ao 5C 1%i%0~0 tace a.Cc/. 2 3

2. 010 3 ~ 010

~~3 57 vV.c3 0 3 I 5O.L>

18 ..

ll408t~l 2 tz>zz< fo Xog vr, 8'C I

4. 010 2XRFi"-RO 32IS So lkb&8~5 //r ro'7~0 +

5 ~ 010

6. 010 ZfiR~2 2 &&2i+&9 3o 2.0.t7 z ts'v.f1

&~~8 /roc CC.t/

14&~%%3 / 7d Xr V.1/

7 010 ~md% 9 27/5 IS ~

14CtXSPr966 r C r 7 8'S' 8'

~

8. 010

9. 010

~F5

~~US 3 3.Stf

</

20.010 }6RSM2 Zd E.cc/7 F4ÃRRH I o 7 "2n.tt.

21.010 ~~4

~RN

/772 S2 Ra~31 /@54 CQRk5 77Enthalpy of accumulator vater is 59.6 Btu/ibm.

IO X IO TO THK CKt4TIMKTKR IS X tS CSS.

KEUFFCL 0 cssER co. asasst w ssA.

a~ 46 1510 z e ei i: ;.:jl I '

i ~ ~

I'-

'ic r

OS tr zC I.':

'I:I

~ I 'Is AOI m w rX mrs I

~ eI.a

i ~

aa I~ '.I:

g XX ~o . 'll jai jj ~

I I i !II I I~

I A sss il O g7~

cn 3 '0 p it ~

I I~

yN m z O

~~~

i[ ",.'I I'

A ~

I e O ~

z z'

z 8

I O

A I

Cl A

O S

p

z I R CA C m ~

O rz O

mm Ol

~~o I

ra 2 2 wO

~ 8 OZ I 0+ m~

0 300 tn 3 cn P rv ZERO FOULING m ~

33 F WATER TEMPERATURE 0

I7 0 31,250 CFM AIR FLOW w 250 1500 GPM WATER FLOW 100 PERCENT SATURATION D

X I-200 0C I-I z 37 X

z 0 150 m

O m

Cl o

00A 100 I 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 HEAT REMOVAL RATE PER FAN (BTUIHR X 106) 4 C

Z Cl IM

IO X IO TO THE CENTIMETER Ib X & CM.

HE~E KEUFFKI d Essrrl co. ra< ~ trrsr. 46 1510

~ ~ ~ ~

yl' Ii'i I; II

~

I

(

'll

'I II

.~ ~h I

I

~

I

'I L'LL II jj

~ I

!I I'!3

[

SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Carolina CONTAINMENTWALL HEAT TRANSFER COEFFICIENT 6.2.1-304 Power III Light Company FINALSAFETY ANALYSIS REPORT

() 0 ~E )0 X 10 TO TNE CENTIMETER NCur)CI. d CSSKR CO. vsosw rsa IC x ts C)C 461510

~ i !I

~ ~

I I I I,il I Z I't

~ ~ ~

i)i t. ~ I 'tt C

! I .i I ~

.' I I ~ ~ ~

~ I ..  ! Ill ~ ~ I CO O os X

'.t. ll')s i I iill I'I i): I I mm

! ~

'ael II) .

'g)l I 'i!! JI ~ ~

I

~ ~

':;I' I~

22

) a) 'I: Ii j t)i

~ ~

-'. ~ ~ ~

ij!':

.s.

e I! ~~ rG' I;::

~ ~

mO

~ ~

I

'le ~ ~

~ I  :))I i:I:II ll. I'I:

rs 0 OZ i! ii.: all; .:I

~ ~

I) 00

~ ~

I I: ml(i li:i ::I i' I e

~~

I '! f gsr ttl

'ra I'i:

r

~ e I ~:I el'el I ala I) ".

O~

~

• I ~ I:I: l)e

.. .  !: :ii t;:

.'I e I:

~ e I:i': ill ll li,!

I<<II~(

i 'I: ~ CO

~ ~ ~ ~

! I I~ ~

iil I~

I, 'Iji liil ii

si.':

II::.:I:>> II  :, I. ~

ill II I a'. I li I!: ii) ~ I II:. .:;I :lii

~

ail I '.

I

~ ~

I

~ ~ ~

I( Igl )ill ~

I I illl (ITI Ii)I ~ I~

~ e I: )

IiI

~ ) III:) ~ I O

0Z I

00 ~

I ~

',il..l! I

~ I

~ I

~

\I

~ ~

I

.:.' II ~

'I:

)j~

~

I Z jl['I I I ~

I I

~

50 j lei ~

Cst ~ ~

m I I

~ ii:::le I '.: I

~ ~

~

I.I

~:I: ':i I I: ~

I:

I il:I

~ ~ ~ ~

I~ i.' '!

~ ~ I i:; 'I'

~

I: I 'I:.'::!

C 00

~ I li ~ ~

i

~ ~ ~

~ ~ ~ 'I m

~ ~ ~

I~ ~  ;)it::;I! ~ ~: !.Ii I

)

~ liI '.hl' O e ~

!ii m

~ ~ ~ ~

.'L! IJ!'" I e

"'l"  !!I I O

A

'i':

La i'." Il ) jl O

!:Il I)ii:il: '.

ll ~

I ~ I!i U III Is ql:(i '-,..;

~ Ie l ~ I a

C) 4eO is I; I I ): .!.': !iii :I ll:'I!::,'f I!

J!

II ~

!i Ill')): I:I I) ~  : I' s I ~

li '

I Iil,)'ll)

I li,j :I I I~ ~I

-'8 I iiil 0)iI i .2 tiiI ~

!)';I 0I!. 3 Gl ao Ctt I ~

I I. Ji: il)'.I: :I': 'I: ~ ~

~ ~

lV

' I. I 'I*; 'l l II~ ! ~ ~

li. ;: I. A It!, I C I'I T  ;=.'li' I'

~

II I

I

~ ~

~~

j it'I

!.I I;)

'l I~ f It ~

I I Ii,l

( EC

I il

~ ~

'I'4 ll.: !i Il.'

I~

~

~

I I-. )ii!.

~

Il i', I

~

I Il:. I ti ~ II ~ !I

SHMPP FSAR One branch injection line spills to containment backpressure. The branch injection line with minimum system resistance is selected to spill to minimize delivery to the core.

The flow delivered into the reactor through the reactor coolant pump seals is assumed to be lost and, therefore, seal injection is not included in the total core delivery.

Safety injection flows computed via this methodology are conservatively low for any postulated break location.

Description of Lar e LOCA Transient The sequence of events following a large break LOCA are presented in Figure 15.6.5-1.

Before the break occurs, the Unit is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. During blowdown, heat from fission product decay, hot internals and the vessel continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nu'cleate boiling.

Thereafter, the core heat transfer is based on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms.

The heat transfer between the Reactor Coolant System and the secondary syste may be in either direction depending on the relative temperatures. In the case of continued heat addition to the secondary, secondary system pressure increases, and the main steam safety valves may actuate to limit the pressure.

Make-up water to the secondary side is automatically provided by the Auxiliary Feedwater System. The safety injection actuation signal isolates the steam generators from normal feedwater flow and initiates emergency flow from the Auxiliary Feedwater System. The secondary flow aids in the reduction of reactor coolant system pressure.

When the Reactor Coolant System depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. Since the loss of off-site power is assumed, the reactor coolant pumps are assumed to trip at the inception of the accident. Previous sensitivity studies have demonstrated the conservatism of this assumption for large break LOCA analyses. The effects of pump coastdown are included in the blowdown analysis.

The blowdo se of the transient ends when the RCS pressure (initially assumed a 22/0 sia) falls to a value approaching that of the containment atmosphere. or to or at the end of the blowdown, some amount of injection water begins to enter the reactor vessel lower plentum. At 'this time (called end of bypass} refill of the reactor vessel lower plenum begins. Refill is complete when emergency core cooling water has filled the lower plenum of the reactor vessel which is bounded by .the bottom of the fuel rods (called bottom of core recovery time.)

The reflood phase of the transient is defined as the time period lasting from the end of refill until the reactor vessel has been filled with water to the

SHNPP FSAR extent that the core temperature rise has been terminated. From the later stage of blowdown and then the beginning of reflood, the safety in)ection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer. The downcomer water elevation head provides the driving force required for the ref looding of the reactor core. The RHR (low head) and charging (high head) pumps aid the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the reflooding process.

Continued operation of the ECCS pumps supplies water during long-term cooling.

Core temperatures have been reduced to long-term steady state levels associated with dissipation of residual heat generation. After the water, level of the refueling water storage tank (EST) reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching from the in)ection mode to the cold leg recirculation mode of operation in which spilled borated water is drawn from the containment sumps by the pumps and returned to the RCS cold legs. The Containment Spray System continues to operate to further reduce containment pressure. Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control the boric acid concentration in the reactor vessel.

Description of Small Break LOCA Transient As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for the small break LOCA there are only three characteristic stages, i.e., a gradual blowdown in which the decrease in water level is checked, core recovery, and long-term recirculation.

15.6.5.3 Core and System f Per ormance 15.6.5.3.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50 (Reference 15.6.5-1).

~ 6LO lua Cea- XA/Wu C 30 I

f Ap ndixX ~Qr.e.'s'o p ie i o 5 RQ t es el at q nsa f y v.a io del red ce'nal ntil 'e a

s on t ov mo ~np r c ed, sis is s o 80 e ced .,use ~ e ana eemed app~, Yijh~-

Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases:

1) blowdown, 2) refill, and 3) reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the Containment, and the fuel and clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.

S1KPP CESAR The description of the various aspects of the LOCA analysis methodology is given in WCAP-8339, Reference l5.6.5-3. This document, describes the ma)or phenomena modeled; the interfaces among the computer codes, and the features gpC,A ) of the codes which ensure compliance with the acceptance criteria. The SATAN- , WREFLOOD, COCO,4and LOCTA-IV codes, which are used in the LOCA analysis, are described in detail in References 15.6.5-4 through l5.6.5-7$

These codes are used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and reflood phases of the LOCA Th SATAN-V cood e analyzes )he thermal&ydraulic transient in the RCS during blowdown and the WREFLOODfcomputer codes C used to calculate this transient during the refill and reflood phases of the accident Thee COCO computer code is used to calculate the containment pressure transient duri LOCA analysis.

ll ur ng a t h ree p h ases of the Similiarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three phases.

SATAN-VI is used to calculate the RCS pressure, enthalpy, density and the mass and energy flow rates in the RCS, as well as energy transfer between the primary and steam generator secondary systems as a function of time during the fll blowdown phase of the LOCA. SATAN-VI also ca calculate t cu a es th e accumulator mass and i nternal pressure and the pipe break mass and inter erna 1 energy ow rates that are assumed to be vented to the Containment during blowdown. During blowdown, no credit is taken for rod insertion or boron conte t f th ij T hee core will shutdown due to void formation. At the end of the blowdown p ase, these data are transferred to the WREFLOOD ccodee. Als tt f so at th e endd of f

( bllowdown, the mass and energy release rates during bl owwd own are transferred to e

during this co e or use in the determination of the containment pressure response first phase of the LOCA. Additional SATAN-VI output data ata from theth end of blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA"IV code IIA ¹¹¹¹¹ 1 1

¹¹ ¹ ECC replac1ng WREFLOOD to prov1de a more real1st1c thermal/hydraul1c s1mulat1on of the reactor core and RCS dur1ng the ref lood phase of a LOCA. F1gurea 4~~ ~'~'+

44 1llustrates how BASH w111 be subst1tuted for gREFLOOO 1n calculat1ng tr>>-

s1ent values of core 1nlet flow, enthalpy, and pressure for the deta1led <<el rod model, LOCTA. Instantaneous values of accumulator flow, safety 1n)ect1on flow, conta1nment amb1ent temperature and pressure, and the t1me of complet1on of RV lo~er plenum ref111 w111 be prov1ded to BASH by MREFLOOD/COCO, wh1ch, 1n the proposed ECCS model, has been relegated solely to prov1d1ng these requ1red boundary cond1t1ons.

~a4cAQ ¹' cwP- F/¹2CC ~ (Arfcic¹ce >~< 5 -32)

0 The 'WREFLOOD code prov1des mass and energy discharge rates from the reactor coolant system to the containment during a core reflood trans1ent. A brief overview is presented here. A complete descr1pt1on of the code 1s available 1n WCAP-9170,~ QArfc'rc'n4'C /S C.S- C)

The bas1c goemetr1c configurat1on in WREFLOOD divides the pr1mary coolant system 1nto three sect1ons; the reactor vessel, the broken loop, and a second loop wh1ch comb1nes all unbroken loops. The reactor vessel region 1s further d1v1ded 1nto a downcomer, lower plenum, and core. Us1ng the in)ection charac-terist1cs of the ECCS as 1nput, the code calculates the downcomer and core water levels as the reflood trans1ent cont1nues. Other bas1c 1nput to WREFLOOD 1ncludes geometric data and 1nit1al and boundary conditions 1n the core, steam generators, and conta1nment.

The COCO code 1s a mathemat1cal model of the containment. Select1on of var1ous opt1ons 1n the code allows the creat1on of models of part1cular con-tatnment belldlngs.'OCO ls descrtbed ln detail ln NCAP-8327. (l(rfcir~re rs t ~ 8)

COCO 1s run s1multaneously with WREFLOOD, which prov1des the necessary mass apd energy 1nputs to the contain-ls 4 A4('f $ 4$

ment on a cont1nuous basis. In though, WREFLOOD 1s only a subsidiary code, running parallel to the main trans1ent analys1s code, BASH.

Dur1ng ref lood, the WREFLOOD/COCO system 1s used only to prov1de containment boundary cond1tions required by BASH.

The LOCTA code is a computer program that evaluates fuel, cladding and coolant here can be found in WCAp-B3Ol ~

temperatures dur1ng a LOCA. A more complete descr1pt1on than

(~d'g~

1s presented

SHNPP FSAR

~

Qur1ng ref111 and ref lood, the yg~~ ECCS model uses a mod1f1ed vers1on of LOCTA to y1eld a s1gn1f1cant 1mprovement 1n fuel rod behav1or pred1ct1on.

I <</

local heat transfer coeff1c1ents, the emp1r1cal FLECHT correlat1on 1s replaced by the BART code. BART employs r1gorous mechan1st1c models to generate heat transfer coeff1c1ents appropr1ate to the actual flow and heat transfer reg1mes exper1enced by the LOCTA fuel rods. Th1s 1s cons1dered a more flex1hle, real1st1c approach than rely1ng on a stat1c emp1r1cal correlat1on.

Small Break LOCA Evaluation Model The WFLASH program used in the analysis of the small break loss of coolant accidenr. is an extension of the FLASH-4 (Reference 15.6.5-15) code developed at the Westinghouse Bettis Atomic Power Laboratory. Th WFLASH (R f 15.6 .5-16) Program permits a detailed spatial representation of the Reactor Coolant System (RCS).

The RCS is nodalized into volumes interconnected by flow th Th b k oop s modeled explicitly with the intact loops lumped into a second loop.

The transient behavior of the system is determined ne from t rom thee governing conservatation equations of mass, energy and momentum applied e th roug o h out h et e ailed description of WFLASH is given in Reference 15.6.5-16.

s y stem,, A det The use of WFLASH in the analysis involves, among other thin s, the representation of the reactor core as a heated control volume with the associated bubble rise model to permit a transient mixture height calculation.

The multi-node capability of the program enables an explicit and detailed spatial representation of various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant transient.

Clad thermal analyses are performed with the LOCTA IU C o d e (R e ference 15.6.5-7) which w c use the RCS pressure, fuel rod power history, steam flow past the uses uncovered part of the core and mixture height history from rom thee WFLASH h ydraulic calculations as input.

Figure 15,6.5-44 gives the safety in)ection flowrate for th e analysis.

sma ll b rea k Figure 15.6.5-45 presents the hot rod power shape utilized e too per orm tthee erform small break anal ysiss presented here. This power shape was chosen beecause it i

P rovides an appropriate distribution of power versus core h e g h t an d also i

local p ower s maximized in the upper regions of the reactor core (10 f t. to 12 ft..).. Thiss power ow shape is skewed to the top of the core with the peak local power occurring at the 10.0 ft. core elevation.

SIKPP FSAR The- bases used to select the numerical values that are input pu parameters para to the analysis have been conservatively determined from extensive sensiti.

ve sens t vity studies (References 15.6.5"19 through 15.6.5-22}. In addition, the requirements of. Appendix K regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core peaking factors the co t i tthee performance erf of the ECCS. Decay heat generated throughout the transient. is also conservatively ca)culated.

The pressurizer heaters are not assumed to operate during ing th e 1 arge and small break LOCA analyses. During the blovdovn depressurization phase of the large break LOCA transient, liquid flashing in the pressurizer forces a very rapid surge of mass out of the pressurizer, leaving only steam vithin a few tens of seconds. The effect of the pressurizer heaters being energized would be to transfer some heat to the fluid surging out,of the pressurizer. Higher enthalpy fluid from the pressurizer mixing vith the broken loop hot let fluid vill result,in a very small pressure increase in the RCS d i th i d t im me that fluid is surging from the pressurizer. The impact of modelling pressurizer heaters during the large break would be to extend the end of blowdovn time by a very small amount. The integrated heat transfer from energized pressurizer heaters over the period of depressurization blowdown is on the order of one sixth of the metal heat release from the upper head. Fuel heat release is significantly greater than the heat input from pressurizer heaters. Additionally, the break flow controls the rate of system depressurization. Overall, pressurizer heaters should have a negligible effect on the large break transient results.

The loss of off-site power assumption has been shown in Reference 15.2.8-1 to result in more limiting peak cladding temperatures since the reactor coolant pumps lose po~er. The pressurizer heaters would not be energized during the loss of off site power.

Small break LOCA's result in a slow system depressurization characterized by distinct mixture levels within the system. All small break LOCA's rely upon the steam generators for some decay heat removal. In fact, the primary RCS pressure vill depend upon a balance between the amount of energy put into the system from core decay heat and metal heat

~ and the amount of energy that is

-removed from the primary by heat transfer through the steam generators and energy removal through the break. If an additional heat source vere present, the system vould respond by resulting in a very slightly higher RCS pressure during that time that the heat source was active. The higher r resultt in a higher break flow during the time the heat source is active. This resu may'wesuH" in an earlier core uncovery and higher peak ccladdin a t ng temperature althouug h thee system s tends to compensate for changes in the mass flow. A higher break mass flow rate results in more mass removal and an earli er core e y. This also results in an earlier transition to two ase an d steam o p has break flow vhi ch causes earlier accumulator in)ection for a short or er d uration core uncoverry. The earlier core uncovery at higher decay heat. 1 eve 1 s tends to i of increase p eak cladding temperatures while the shorter duration of uncovery tends to decrease peak cladding temperature.

SINPP FSAR The loss of off site power has eak cladding temperature for also been determined to be li iti i small break LOCA's. Pressurizer heaters are not pea operable during the loss of off site power. Other systems which are not operable for the loss of off site po~er are the reactor eac or coocoolsantt pumps andd the steam dump control system. FSAR small break LOCAs result in the secondary pressure rising to the secondary safety valve setpoints. If the steam dump control system were operable, the RCS pressure would be si ifi 1 g ivving a considerable benefit in terms of peak cladding temperature.

So the effect of pressurizer heaters being energized during a large break or small break LOCA would have a small effect on thee resultresu ts an d the h more limiting assumption of the loss of offsite power precludes operability.

15.6.5.3.3 Results Large Break Results Based on the results of the LOCA sensitivity s t u di es (R e ferences 15.6.5-20 through 15.6.5-22), the limiting large break was as fou t b e the oun d to h doublemnded d

cold leg guillotine (DECLG). Therefore, only the DECLG bre k 1 thee lar argee break ECCS performance analysis. Calculations were performed for a range of Moody break discharge coefficients. The results of these calculations are summarized in Tables 15.6.5-1 and 15.6.5-3.

Factors affecting break flow in a Westinghouse PWR and the lower limit of break discharge coefficient based on experimental data are discussed in Reference 15.6.5-21. Conclusions in that report are th t b oof the Moody Moo discharge coefficient is about 0.6 and that varying the discharge coefficients from a maximum of 1.0 to a minimum of 0.4 covers all uncertainties associated with the prediction of the br k fl ot ne ttype severance of a cold leg pipe. The position to limit the break guuillotine discharge coefficient to that range has been reviewed and appro ed b th Therefore, analyzing a LOCA for break discharge coefficients less than 0.4 is not consistent with experimental data or with the established procedure for a 10CFR50 Appendix K evaluation of ECCS performance.

The mass and energy release data for the break resulting in the highest calculated peak clad temperature are presented in Section 6.2 '.5.

Figures 15.6.5-4 through 15.6.5-30 present the parameters of principal interest from the large break ECCS analyses. For all cases analyzed transients of the following parameters are presented:

a) Hot spot clad temperature.

b) Coolant pressure in the reactor core.

c) Water level in the core and downcomer during reflood.

d) Core reflooding rate.

INSERT A A comparison of the core ref lood transients (Figures 15.6.5-6, 21, and 27) and core average rod temperature transients (Figures 15.6.5-31, 32, and 33) shows that the 0.0 DECLG leads to the most severe ref lood transient due to the high initial clad temperature at the beginning of ref lood. The hot assembly calculation also reflects the same variation of initial clad temperature with break size. However, the variation of peak clad temperature for the hot assembly calculation is less severe. In fact, the 0.8 DECLG becomes slightly higher due to higher blockage and higher power at turnaround. Since the sensitivity of peak clad temperature is very weak, and since the ref lood transient shows definite improvement with larger break size, it is felt that the analysis of the three breaks presented is sufficient. In addition, sufficient margin exists to account for the small variation exhibited by the hot rod with break size.

(1347GAS/ccc )

SHNPP FSAR e) Thermal power during blowdown.

f) Containment pressure.

For the limiting break analyzed, the following additional transient parameters are presented:

a) Core flow during blowdown (inlet and outlet).

b) Core heat transfer coefficients.

c) Hot spot fluid temperature.

d) Mass released to Containment during blowdown.

e) Energy released to Containment during blowdown.

f) Fluid quality in the hot assembly during blowdown.

g) Mass velocity during blowdown.

h) Accumulator water flow rate during blowdown.

i) Pumped safety injection water flow rate during reflood.

The maximum iso clad temperature calculated for a large break is SHAN F which is less than the acceptance criteria limit of 2200 F of 10 CFR 50.46. The maximum local metal water reaction is AfP percent which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46. The total core metal~ater reaction is less than 0.3 percent for all breaks, as compared with the I percent criterion of 10 CRF 50.46, and 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. Full ECCS flow assumed in the LOCA analysis has not been noted to result in a higher peak clad temperature in 3 loop plants. Thus, the analysis presented remains conservative Reference 15.6.5-27 provides a more detailed discussion.

SHNPP FSAR TABLE 15.6.5-1 LARGE BREAK - TIME SE UENCE OF EVENTS OCCURRENCE TIME (SECONDS)

DECLG, CD ~ 0.4 DECLG, CD ~ Oob DECLG, CD ~ 0+8 Accident Initiation Oo0 0.0 Oo0 Reactor Trip Signal A4AC4 . Vyg ~, 9'33 ~,927 Safety In)ection 1~ 03 ~ 840 ~ 74 Actuation Signal Start Accumulator Injection 83+0 /X 4 4gss // W 9 /Z.

End of ECC Bypass B4+Q Zo.o2. k$ 40Q f 9 S$

2/

End of Bio@down gQ4k 3o./s'fdfpl 8Q.C g l~ Z./

I'5'4AM Ff Bottom of Core Recovery V2 olg @A4Q& 9C. 927 Q~k 33. Pgo Accumulators Empty 40@4&4 6 o 5 f'0 maaa ~2 <

7 closed signal; turbine trip signal

2) SI signal (on high containment pressure) (19.2 psia) ~ 7K

3) Accumulator injection

4) Safety injection begins ZS.7<

5) Containment fan coolers begin 33. 33

6) Containment spray begins 46.03 Reactor Trip Si al occurs on compensated pressurizer pressure signal 1860 psia)

Accumulator Injection - injection begins when RCS pressure drops to 600 psia. No failures assumed. Safety Injection Si al - occurs on high containment pressure (19.2 psia) Safety Injection - There is a 25 second delay before injection begins. Delay 2.0 sec signal generation time 14.5 begin charging flow 19.5 begin full SI flow 24.5 begin RHR flow J ~1 See attached curve for SI flow during worst large break transient. (Figure 15.6.5-18) Accumulator Injection when pressure in RCS reaches 600 psia. See attached curve for accumulator injection flow no failures. (Figure 15.6.5-16) SHNPP FSAR TABLE 15.6.5-1A (Cont'd) Containment Heat Removal System: Fan Coolers - 4 fan coolers operate containment pressure signal at 1.03 sec Delay of 32.3 sec includes: 7.3 sec delay to start fan coolers 25.0 delay to get power up 32.3 sec 10 sec diesel startup 5 sec sequences 10 sec fan coolers to reach full speed Fan coolers cooled by service water at 40 F. HEAT REMOVAL TABLE Temp ( F) 150 180 220 258 Q (BTU/sec) 7208.3 12355.6 20930.6 29555. 6 Containment Spray: Flow 3641 GPM Temp ~ 40 F Actuated on Hi-3 containment pressure (12 psig) Occurs at 4.0 sec Spray has 42.03 sec delay. Spray starts at 46.03 sec. SHNPP FSAR TABLE 15 6.5-2 INPUT PARAMETERS USED IN THE EGGS ANALYSIS Gore Power* 2775 Mwt (p. l3C Peak Linear Power (Includes 102Z factor) iks40$ kM/ft. Total Peaking Factor, Fq SaCk 2~ C ~ hxial Peaking Factor, FZ Power Shape Large break-chopped cosine Small Break-See Figure 15.6i5%5 Full Assembly Array 17x 17 /r5d Accuuulator Mater Volume (nominal) kOQ8 ft. ccauulator Accanulator Tank Volume (nominal) 1450 ft /accunulator Accumulator Gas Pressure (minimum) 600 psia Safety In5ection Pumped Flow See Figures 15.6.5-18 and 15o6o5%4 Containment Parameters See Tables 6.2.1&2, 6.2.1&3, and Figures 6i2il 303 and 6.2.1-304 Initial Loop Flow Vessel Inlet Temperature SK4~~ ~ 5 5 ~ l Vessel Outlet Temperature 4,2.Z. o 'F ,. Reactor Coolant Pressure 2280 psia Steam Pressure ~Ma- 0 S2.c's' Steam Generator Tube Plugging Level ec%%u-

• 2X is added to this power level to account for calorimetric error.

SHNPP FSAR TABLE 15s6s5-3 Results DECLG, CD ~ Os4 DECLG, CD ~ Os6 DECLG, CD ~ Os 8 Peak clad temperature Location (ft.) (F) ~ Qggr stot 7i CR 7f4i. ~ 7r4S iE'2< 7 25 Maximum local clad(va ter reaction (X) Location (ft.) PlgjS CsO ~ Ae or 3.2K Total core clad/vater <~3 < 3 <,3 reaction (X) Hot rod burst time (seconds) Location (fthm) ~ bio 6.0 4aa 7Z 4z

4. Z.g

REFERENCES SECTION 15 ' Burnett, T. W. '5.6.1-1 T., et al., "LOFTRAN Code Description," t WCAP>>7907 June 1972. 15.6.5-1 "Acceptance Criteria for Emergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50. Federal Register, Volume 39, Number 3, January 4, 1974. 15.6.5-2 "Reactor Safety Study - hn Assessment of Accident Risks in U. ST Commercial Nuclear Power Plants," WASH-1400, NUREG 75/014, October, 1975o 15.6 '-3 Bordelon, F. M., Massie, H. W. and Zordan, T. A., "Westinghouse ECCS Evaluation Model - Summary," WCAP-8339 (Non-Proprietar ) July, 1974. 15.6.5-4 Bordelon, F. M., et al., "SATAN-VI Program.'omprehensive Space-Time Dependent Analysis of Loss of Coolant," WCAP-8302, June, 1974 (Proprietary) and WCAP-8306, June, 1974 (Non-Proprietary)i 15.6.5-5 Kelly, R. D. et al., "Calculational Model for Core Ref looding After a Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170, June, 1974 (Proprietary) and WCAP-8171, June, 1974 (Non-Proprietary). 15.6 ~ 5-6 Bordelon, F. M. and Murphy, E. T.> "Containment Pressure Analysis Code (COCO)," WCAP>>8327, June, 1974 (Proprietary) and WCAP-8326, June 1974 (Non-Proprietary). 15 ~ 6 ~ 5-7 Bordelon, F. M., et el ~ , "LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301, June, 1974 (Proprietary) and WCAP-8305, June, 1974 (Non-Proprietary). 15.6 '-8 PWR FLECHT Final Report, WCAP-7931, October 1972. 15.6.5-9 Bordelon, F. M., et al ~ , "Westinghouse ECCS Evaluation Model-Supplementary Information," WCAP-8471-P-A, April, 1975 (Proprietary) and WCAP-8472-A, April, 1975 (Non>>Proprietary). 15.6 '-10 "Westinghouse ECCS Evaluation Model October 1975 Version," WCAP-8622, November 1975 (Proprietary), and WCAP-8623, November 1975 (Non-Proprietary). 15e6.5-ll Letter from C. Eicheldinger of Westinghouse Electr'ic Corporation to D. B. Vassallo" of the Nuclear Regulatory Commission, Letter Number NS-CE-924 dated January 23, 1976. 15.6 '-12 Rahe, E. P., 'westinghouse ECCS Evaluation Model, 1981 (Proprietary Version), RECAP-9221-P-A (Non-Proprietary Vers)on,'lCAP-9220-P-A Version), Rev1s1on 1, 1981. SHNPP FSAR REFERENCES SECTION 15.6 (Cont'd) 15.6.5-13 Letter from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, Letter Number NS-TNA-1981, Savember 1, 1978. 15.6.5-14 Letter from T. M. Anderson of Westinghouse Electric Corporation to to John Stolz of the Nuclear Regulatory Commission, Letter Number NS-TMA-2014, December 11, 1978. 15.6.5-15 Porsching, T. A., Murphy, J. H., Redfield, J. A., and Davis, V. "FLASH-4: A Fully Implicit FORTRAN-IU Program for the Digital Simulation of Transients in a Reactor Plant," WAPD-TM-840; Bettis Atomic Power Laboratory (March, 1969). 15.6.5-16 Esposito, V. J., Kssavan, K. and Maul, B. A., "WFLASH-A FORTRAN IU Computer Program for Simulation of Transients in a Multi-Loop P R," WCAP-8200, Revision 2, July, 1974 (Proprietary) and WCAP-8261, Revision 1, July, 1974 (Non-Proprietary). 15.6.5-17 Skwarek, R., Johnson, W. Meyer, P., "Westinghouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8920-P-A (Proprietary) and WCAP-8971 (Non-Proprietary), April 1977. 15.6. 5" 18 Letter from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, Letter Number NS-TMA-2030, January, 1979.

15. 6. 5-19 "Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP-8341, July, 1924 (Proprietary), WCAP-8342, July, 1974 (Non-Proprietary).

15. 6. 5-20 Julian, H. V., Tabone, C. J., Thompson, C. M., "Westinhouse ECCS Three Loop Plant (17 x 17) Sensitivity Studies," WCAP-8853, September, 1976 (Non-Proprietary).

15. 6. 5"21 Salvatori, R., "Westinghouse ECCS - Plant Sensitivity Studies,"

WCAP-8340, July, 1974 (Proprietary) and WCAP-8356, July, 1974 (Non-Proprietary). 15.6.5-22 Buterbaugh, T. L., Julian, H. U., Tome, A. E., "Westinghouse ECCS Three Loop Plant (17 x 17) Sensitivity Studies," WCAP-8572, July, 1975 (Proprietary) and WCAP-8573, July, 1975 (Non-Proprietary). 15.6.5-23 Murphy, K. G., Compe, K. M., Nuclear Power Plant Control Room Ventilation System Design For Meeting General Criterion, 13TH AEC Air Cleaning Conference (1973).

15. 6. 5-24 Stolz, J. F., Letter to T. M. Anderson (Westinghouse), Transmitting Safety Evaluation Report of Westinghouse ECCS Evaluation Model, February, 1978 Version, August 29, 1978.

15.6.5-25 Stolz, J. F., Letter to T. M. Anderson (Westinghouse), Transmitting Safety Evaluat,ion of Westinghouse ECCS Small Break, October, 1975 Model, June 8, 1978. Amendment No. 5 SHNPP FSAR

REFERENCES:

SECTION 15.6 (Cont'd) 15.6.5-26 HcFetrtdge, R. H., D. C. Garner, "Study of Reactor Vessel Upper Head Region Fluid Temperature," WCAP-9404, Rev. 1, December 1978.

15.6.5-27 Letter from P. Rahe of Westinghouse Electric Corporation to R. Tedesco of the Nuclear Regulatory Commission, Letter Number NS-EPR-2538, December 12, 1981.

15.6.5-28 "Report on Small Break Accidents for Westinghouse NSSS," WCAP-9600.

15.6.5-29 MASH 1258 "Numerical Guides for Design Obgective and limiting Conditions for Operation to Hect the Criterion As Low As Practicable for Radioactive Haterial in Light-WaterWooled Nuclear Power Reactor Effluents." Volume 2, July 1973, U.S. Atomic Energy Commission.

15.6.5-30 ORNL-TH-212, Part IV, "Design Considerations of Reactor Containment Spray Systems. Calculation of Iodine-Water Partition Coefficients." L. F. Parsly, January 1970, U.S. Atomic Energy.

Commission.

Coll1er, G., et al., "BART-Al: A Computer Code for the Best-Est1mate t C. (, 5-3 I Analys1s of Ref lood Trans1ent," QCAP-9561, January 1980. (Qest1nghouse Propr1etary)

Amendment Nn lI

BREAK OCCURS 8 REACTOR TRIP ICOMPENSATED PRESSURIZER PRESSURE)

L PUMPED SAFETY INJECTION SIGNAL IHI-I CONT. PRESS. OR Lo PRESSURIZER PRESS.)

0 W PUMPED SAFETY INJECTION BEGINS IASSUMING OFFSITE POWER AVAILABLE) 0 0 ACCUMULATOR INJECTION W

CONTAINMENT HEAT REMOVAL SYSTEM INITIATION IASSUMING OFFSITE POWER AVAILABLE)

N f Nl) I)f BYPASS END OF BLOWOOWN PUMPED SAFETY INJECTION BEGINS IASSUMING LOSS OF OFFSITE POWER)

BOTTOM OF CORE RECOVERY R CONTAINMENT HEAT REMOVAL SYSTEM INITIATION IASSUMING LOSS OF OFFS'ITE PnwERI E

F ACCUMULATORS EMPTY L

0 0

D CORE QUENCHED L

0 N SWITCH To COLD LEG RECIRCULATION ON RWST LOW LEVEL ALARM IMANUAI. ACTION)

G T

E R

SWITCH TO LONG TERM RECIRCUI.ATION IMANUAL ACTION)

M C

0 0

L I

H G

SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Sequence of Events for Large Break 15.6.5.1.

( <<II I ) I Illwi:t 5 Li)lhl Ct>mluny FINAL SAFETY ANALYSIS REPORT

RLOMDOMH FHD OF BIOITDOITII (COB) REFILLS RCFLOUD I

'0 r 0 orZcn LUCIA BAR I /LUCIA m mm Qr LUCIA C f VIL ROO IHERHAL ~ Hf CIWIICA( TUEL ROO IN(RHAL ~ HECIWIICAL CONDITIONS I 0, wO CIWDIIICVCS DURING BLOMIÃNRI IAIRIIIG Rf fill~ R(fl000 2 T O OZ IRIT ASSIHDlv Af EOB I

I A Ot m+ AVERAGE RCO CONDITIONS cn O

3 r- fllCIA,Alf5HOI ASS(HOLT IRANSf(R CO(ffl'LITNI I

I%AT TRANSTER COfffICTIHT

'II cn g Br cn HEAT HOT ASSEHBLT 1Hf RHOHTORAUL IC m

'0

~ z CV%1110NS DURING REfl000 I

O I L

Hnl ASSTHBLT CORE INL(1 TLOM, ENIIIAI.I'T~ PR(55VRE HVSS TflOCIIV. OVhllIV. raf~SINE SAIAH s BASH I

RC'5 ~ CORE AT EOB IIUIRUIIP cR RCS COIOITIDNS DURING RETLOOD I

n RCS, TORE THERHOHTORAVLIC I cR CONDITIONS DURING BLONDOWI CORE IN(El fLOU EHIHALPT CORE OUTLET TLOM, ENIHAlPV RCS Al EOB I I I ONE THERHOHTORAUL IC COROT 1 IONS n I DURING RfflOOD I

D I 0

Rf fill TINE ACE(PC%RIOR B S.l. TIDNSe 01 CDNIAIIONHT PRESSINL HASS, EN(PGv Rf((ASE MRCf'LOUD/COCO I

MREFLOOD I CAlCVlATES BREAK HASS ~ fNERGY RELEASE I

HASS ~ ENERGT RELEASE CONTR INN(HI PRESSURE COCO CONIAIIRC(NI PRfSSURE Chl CIA AIES CON IA IIPCI Nl PRT SSIN(

LA(CD(ATE5 TTPIIAINIEHI f'Pl SSVRI I

L

~ 0

~ ~ ~ ~ '

~ S

COL 0.8 DECLG CLAD AVC.TEMP.HOT ROD 2000.0 C)

Cl 725' 1500.0 I

1000.0 500.00 0.0 Cl Cl C) ED CI ED C7 ED C) ED C7 CI CD C) C) lA IA CD lA CI CU TIME ISEC.I SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Cat olina Peak Clad Temperature DECLG (CD = 0.4) 15.6.5-4.

Power 5 Light Company r eao s ~ s a rrvv mala velc'cohttT

~

2SOO.O SHEARON HARRIS UNIT I ICOL) 0 8 OECLG C.OA'S.G ~ TU8E PLVGGlNG It/ACCUH TUNlNG PRESSURE CORE 80TTOH I ) TOP o I~I OOOO.O

$ SOO.O 1000.0 500.00 0.0 TAHE. ISECI SHEARON NARRIS FIGURE NUCLEAR POWER PLANT

( ~I I II I I I let Core Pressure DFCL(s (CD = Oaf 15.6.5 5.

I I le'I Rl L IIIIII ClllllILlllf FINAL SAFETY ANALYSIS R fPORT

WATER LEVEL (FT)

~ <<

<I<

n<<

~

~e SHE ARON HARRIS FIGURE NUCLEAR POWER PLANT

(:;<<<<Ii<<;< Ref food Transirnt - Core 8< Downconre<r W <Icr 15.6.b 6.

P<<<<<<.'< 5 L<<II<< G<<<<<IM<<y Levels DECLG tCD = 0.%)

FINAI SAFETY ANALYSIS REPORT 8

VIi(FT/$)

m IA m

n

~ v SHE ARON I I A I I IIIS Nll(;I E All P()WI R I'I AN I II<!Ibxt<l I tt<ttsi<!<<l (:<w<! It<I<.l V<;1<x:tly

(:.x<>luw I <IWI!I 5 L<t)llt <<. till!Ll<lp DECI.G (CI) 0.4)

FINAL SAFETY ANALYSIS REPORT

SHE AROII HARRJ5 UIIlT I ICOL)

Owe OKCLG 6.0+5.G. TVBK PLVGGlNG VIACCUH TUIIlNG POVER

5. 7500 CI 4

l.5000 l o F500 l.0000

0. T500 0.5000 0.2500 0.0 CI CI TlNE ISKCI SHEARON HARRIS FIGURE NUCLFAR POWER PLANT I rilltllmd Core Power Transient DECLG (CD = O.i) 15.6.5-8.

I kiwi.i a Lion CiimIu<<y I INALSAFETY ANALYSIS R/PORT

HE- 10 X IO TO THE CENTIMETER II X tS CM.

C IltVttEL& ESMII CO. <<aOtieeta 46 1510 l"I

~

~ 'I I

IllI

~

~ ~

I I= ~ ~

~ ~

I':.:I

'II

P i'I

'I I~

I"/ Ii)!

Pt

'I I

~

I I:. I ~ II

! ii I ll I

~

~

I

~

I~ ~(

FIGURE SHE ARON HARRIS NUCLEAR POWER PLANT all I II Illa Containment Plessure DECLG (CD 0. P) 15.6.5 9 PIIVw.l Ãc Llqlll CIImtully libel FINAL SAFETY ANALYSIS REPORT

IOOO.O 5NKARON lrARRI5 UNEI I rCOLl 0.1 OECLG C.05 S.G. tUBE PLUGGING VIACCUat IUNlNG 7-fl QVRAIK CORK BQIIOH l l IO> ~ lal 5000.0 Gt Vl

~P 0.0

-2500.0

-5000.0 7000.0 Ct I{HE ISEC)

SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Carolina Core Flow {Top and Bottom) 15.6.5-10.

Power 8r Lrghl Company DECLG {CD = O.g)

FlNALSAFETY ANALYSIS REPORT

~

~$

~8~$

~S~E 0 ~

~$

~S~RIIW

~I~&'f%

Lt.,

5 ~ ' ' I ~ RISHI ~ ~ SW

~ IW filif~ I ~ l f IS% ll 3

~ ar ~,ri ~ ieeI Ilfreewe

~ I WP I I ~ faf ~ f ~ ffE% I IS% l:fl 1 iaIirlr ra.lir,ra I

iraI>> I fif I 5 ~ ~I W I I ~ f ~ I R f 0 f ~ ~ E W I Sar ~ ~ I Wf )fllNaI~ IRTa IN&'Q 5 firI fr tt%il fll I Llll tll lAlrl[ill (A l'lll II ~ ll~ I ~ I I SQM Rill;IllIlL'IIII IIIVlllrl%L m

~IWR I'jQlll'(J lf IS,I%ilkL'l'iR MIEH~'O'SRLl&lkrlJllfg SF& RklIfili I Il1llllNllbVItl1i%Ill

%II;III IItiflllftj'Lfl'ktllfikQ

fl Z

I Z In t) C

> C OIII I-m= X 2000.0 mm y I:." ~O OZ COL 0.8 OECLG a-=

I QI'lly

~x FLU10 TEHPERATURE c li50.0 III 3 ~ g)

III Q

~ 'J:

m Z 0

1500.0

~ 1250.0

'll C

CL 1000.0 ID la.

750.00 I

O Ill O

I Cl 500.00 A

0 II pboo 250.00 0.0 Ql

'll Cl ID C)

Cl C7 CD ED ED CD CJI C C7 C7 C) C) m Vl CD CD CU IA hl Ct AI Itl TlHE tSEC ~ l

I.OOE N5 SKKAROII IIARRl5 UlllT I ICOLI 0 ~ ~ OCCLG C ~ Oj% 5 ~ G ~ TUB K PLUGGllIG VilCCU< TUlllIIG BRClK fLOV LJ B.OXNl4 C.OOEHa CI i.OOE<a Z.OOMPH) ~

O.O CI Cr TlHE ISECI SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Carolina Break Flow Rate OECLG (CO = 0.4) 15.6.5-13.

Power & Light Company FINAL SAFETY ANALYSIS REPORT

'5.00f N)1 SHCiROII HARRIS UIIlt I ICOLI 0~ I OECLG C+0+ 5+Go tUBE PLUGGlIIG V/ACCUN TUktHG BREAa f WKRG Y 4J o.OX41 3.00EN) t 2.00E+1 1.00f N) 1 0.0 O

O SHEARON HARRIS NUCLEAR POWER PLANT CilI nl I I I~0 pg

~

- Ac 5'x COL OVALITV 0.8 OF OECLG FLulo O

cn 3 1.2500 en/ I

~ cn ffl gl p

1.0000 C

CL 0 O.V50O G

C I

Pt C

O I

0 fll 0.5000 O

I Gl 0O I

goy 0.2500 0.0 Ql fl 000 000 Cl C 00 0 0D 000 O 00 0 00C3 0 0O 00 0O O0 0 0 0 0 000 D

0 0 0 0 0 O 0 O 00 0 Ifl 0 0 0

~ ~

Ql m sn 0 O 0

~

~

0 0 D 0 0 0 0

~

0 0 0 Al Pl IIl cU m m TIHE (SEC.)

$ ,5X+gi SIIEAROII IIARRIS UIIlt I ICOL)

O.i OECLC C.O5C S.C. Tuel PLUCrtIIC ViACCUH VUIIlIIC ACCUH. FLOII LJ NN0.0 O

CO00.0 Lf LI F000,0 2'000.0 0.0 cr C7 SHEARON HARRIS FIGURE Tiara (Se<)

NUCLEAR POWER PLANT Carolina Accumulator Flow (Bfowdown) 15.6.5-16.

Power & Light Company DECLG (CD 04)

FINAL SAFETY ANAI.YSIS REPORT

Z I Z III C C 0.0 m

C

"=

rZ O

mm UI Cll Qo ~ D g) Al I COL 0.8 OECLG I

~

vO HA55 VKLOClTY OZ r o .- Nx m~ CD Xl r-CI g7 CO 3 II gy Ul

~ CO

-500.00 m Z O

~ - -f000.0

Ã

( an an C

x-1500.0 I

O m

A I

Gl O -2000.0 O

n 0

-2500.0 0 O O O O O D 00 0O 0D o0 00 O D oD 0D 00 00 00 O 0 O 0 CJI D 0AI O

08) O 0 D 0 0D 0D 00 0 D D O 0 D 0 O~ 0 00 00 00

~

D 0 0 00 0 ~

C) aA N O0 O Ql C 0 O 0 0 0 ~

AI Cn IA e 0AI 0 ~ ~

0 CV =

0 Pl lA v m TlME (5l:C.)

0 IO X 10 TO THE CENTIMETER IS X 25 CSI.

F

'fg~W KEUFFEI. 8 ASSER CO <<sot sm <<sa 46 1510

I I

+It "I:

~ ~ ~~

I>>

>> ~

I:

>> I

>>t'.

I

~ ~ I:It

~ ~

.I>> ~ ~

I - ~ ~

I

~

jii t >>II

.II:

~ >> t ~ >>

I II I L! jt:!

I ~~

II >>

II

~ ~

I

ii,l

~ I

I:

~ ~

I~ I~ I~

~

t I>>

I >>

SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Pumped ECCS Flow During Reffood (CD ~ 0.4)

Carotina Powel 5 Light Company 15.6.5-18.

EIKlttt eaVFTY ANALYSIS REPORT

II z

r cn <r.

z= Qz p vO CLAD AVG.TEHP.HOT ROD rO<<; %x m~

cn r r-y <<n 2000.0 m~

7ZS w I

1500.0 I g pre EP Ci D

III F000.0 CD C

CD U

m O

I 500.00 C)

O O

II o

Cl 0.0 CI Cl Cl Cl CI II CI Cl Cl CI CI CD CI0 A Irl CI Cl Ifl C CI CI <<Cl m

co TlH'E 1 SEC.)

t Z500.0 SttE ARON HARRIS UNIT I I(OLI 0 ~ 6 OE(LG 6 ~ 0+S.G TUBE PLUGGING II/A((UH TuttirtG I'RESSURE CORE BOIIOH I I TOP c I~ I Z000.0 I500.0 I000.0 500.00 0.0 CD CD CD CD CD CD CD CD CD CD CD CD CD CD AJ TIHE ISE()

SHEARON HARRIS FIGURE NUCI.EAR POlSER PLANT Carolina Core Pressure DECLG (CD = 0.6) 15.6.5-20.

Power 8r Light Company FINAL SAFETY ANALYSIS REPORT

1 ~.~

K

~ .I

~.~ ill.~ lie. ~

T I!If(SEC)

SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Ref lood Transient Core & Downcomer Water Levels 15.6.5-21.

C >rnlina Pnwr!r DECLG (CD = 0.6)

Sr Lriiln Cr>rniuny FINAL SAFETY ANALYSIS REPORT

1, ~

~ o~

~ Z.O

~.~ 1 ~ I. ~ 211. ~

T I IIE ( SK C)

SHEARON HARRIS FIGURE NUCLEAR POWER PLANT 15.6.5.22.

Ref lood Transient Core Inlet Velocity Carolina Power IIr Light Company DECLG (CD ~ 0.6)

FINAL SAFETY ANALYSIS REPORT

2.0000 SHEAROII HARR15 QII1T 1 ICOL) 0.6 OECLC 6.0j@S.G. TUBE PLVGGIHG M jACCUH TVHIHG PDVER I. 1500 1.5000 1.25UO 1.0000 0.: 590 a

O.SDr.O 0.25r."U 0.0 CI CI CI CI CI SHEARON HARRIS FIGURE NUCLEAR POWER PLANT Carolina Core Power Transient DECLG (CD ~ 0.6) 05.6.5-23.

Power 5 Light Company FINAL SAFETY ANALYSIS REPORT

iE IO X IO TO THF. CENTIMETEIT le X a CM.

ICEUFFEL 4h ESStR CO. eeeest w eSa. 46 1510 e

.I ile let

~ ~

I e ~

I I II ~

~

't I~ ~ I ~ ~

~ e

~

I :I ~

I e

~ ~

e iile I~ ~ I: II e

ts ~ ~

e

~

I

I,II e

I,'

-;i.i e

'i 'i.:,'ll I

~

I) et Is I::

e jl e

II e

~

II

~

~

I ~

Is i I '

~

I I

~

I I~

II et Ie SHEARON HARRIS FIGURE

.'I NUCLEAR POWER PLANT Col oIina Containment Pressure DECLG (CD = 0.6) 15.6.5-24.

~

llI Pt)wt'.I IIt t l<Illl CtttnILltlg I I

FINAL SAFETY ANALYSIS REPORT

2500.0 COL 0. i OL'C LG CLAO AVG.TEHP.HOT ROD 2000.0 O

O 7.dp l

O f500.0 I

1000.0 500.00 0.0 CD ED CD lD CI lD lD lD lD C7 Cl CD lD CD lD CD C7 lD lA lD V1 lD lA lD I/1 ED Lf7 fV V1 AJ SHEARON HARRIS FIGURE TIME CSEC.)

NUCLEAR POWER PLANT (I(

(.rll Ollllrl Peak Clad Tempera(urc - DECLG (CD = 04) 15.6.5-25.

I I IWIll III I IIIIII (rl llllILlllg FINAL SAFETY ANALYSIS REPORT

2500.0 SHKAROH HARRjS UHlT 'I ICOLI O.l OECLG 6.0~S.G. TUBE PLUGGlIIG Vl ACCUH TUHlHG PRE 5 SURE CORE BOTTOH I 1 TOP ~ I~ '1 2000.0 I500.0 l000.0 500.00 0.0 C)

CI Cl ro CD Ct Ay Co eee C> O SHEARON HARRIS NUCLEAR POWER PLANT FIGURE T(or(p(Q+C)

Ca(olina Core Pressure DECLG (CD = O.f) 15.6.5-26.

Power I('r L(ghi Company FINAL SAFETY ANALYSIS REPORT

I I~ . ~

K I

~.~

~.~ tia. ~ ?aa. ~

XIII'SEC)

SHEARON HARRIS FIGURE NUCLEAR I'OWER PLANT Rullootl Trttttsient - Core & Downcorner Weter Levels C4I OI I I IJ 15.6.5.27.

Powci & Light Company DECLG (CD = 0.8)

FINAL SAFETY ANALYSIS REPORT

4 tt ~

~ ad

2. ~

~.~ 1 ~ II. ~  ? II.I 7 I II'SE C)

SHEAR(>N HARRIS FIGURE NUCLEAR I'OWER PLANT Cvi olina Ref lood Transient Core Inlet Velocity 15.6.5.28.

Power & Light Company DECLG (CO = 0.0)

FINAL SAFETY ANALYSIS REPORT

SHEAHON HARRIS FIGURE NLICLEAR POKER I'LANT I'lslIllW Core Power TraIIsieIII DECLG (CD 0.0) 1 5.6.5 29.

PllWI'.I 5 Llgl I Gill IlJl'lf 1 1 1 FINAL SAFETY ANALYSIS REPORT 2.0000 5IIEAROII HARRI5 UNIT I ICPI.I 0.4 OEELG 6.0+ S.G. TUBE PLUGGING M/ACEUH TUIIIIIG POIIER I.1500 n

1.5000 I. 2500 I.0000 Q. T500 IL 0.5000 0.2500 0.0 Cl C) Cl Cl Cl Cl AJ TIRE I SEC I

~ V rj

~. e e I I

IO X lo To THE cENTIMETER Is x ss cai.

C KEUFFEL h ESSER EO. rara>>r raa 46 1510

'I:: II 'I '

~ ~ H!

I! I ':

~ ~

~ .'.i:

>>Ii'l: ,il :>>II

~ ~

I>> ~

~>>

~

I ~~ >>

I>> ~ ~

~ ~

UI I'I

~ I: ~ ~

jh I>>

I j I .I li l I

I~

I;. I ~

I I~

FIGURE SHEARON HARRIS NUCLEAR POWER PLANT (alolina Containment Pressure DECLG (CD = 00) 15.6.5-30.

Power 5 Light Company FIMhl 'chFCTY ANAI YSIS REPORT

IHI I o

~.~

~.~ $ ~ I, ~ ?I ~ . ~

TlIIE(SEC) t5',5'- ~f Core Aacrcrc loot Tee/or>ur o:

C ~ o, F

L'ni I 7 till.~

~4

~1III.~

~.~

~.~ I I I. ~

T1NE(SEC)

0 ggEq 4ow y~RR( 5 Jill el I /Ac 'W Q

I wl ill. ~

~ g I

~a~

~.~ Ill o ~ fii.~

l 1 IIE (SEC) o J g,~ fEJ~N P y c P4. ~ ~

WO p i5:a. S ~3 pe~.

E I