ML18018B912

Forwards Addl Info Re SER License Condition 7 on Restricting Operations Above 90% Power Until Completion of Rod Drop Analyses,Per WCAP-10297-P.Draft FSAR Pages Will Be Incorporated in Future Amend

ML18018B912
Person / Time
Site:	Harris
Issue date:	05/07/1985
From:	ZIMMERMAN S R CAROLINA POWER & LIGHT CO.
To:	DENTON H R Office of Nuclear Reactor Regulation
References
MLS-85-119, NUDOCS 8505140405
Download: ML18018B912 (28)

Text

REGULA Y INFORMATION DISTRIBUTIO YSTEM (RIDS)ACCESSION NBR:8505140405 DOC~DATE!.85/05/07 NOTARIZED:

NO FACIL:50-400 Shearon Harris Nuclear Power Planti Unit ii Carolina AUTHRNAME AUTHOR AFFILIATION ZIMMERMANiS

~RE Carolina Power.8 Light Co, RECIP~NAME RECIPIENT AFFILIATION DKNTONiH,RR Office of Nuclear Reactor Regulationi Director DOCKET 05000400 SUBJECTS Forwards addi info re SKR License Condition 78 on restricting operations above 90K power until completion of rod drop analysesiper I'ICAP 10297 P,Draft FSAR pages will be'ncorporated in future.amend, DISTRIBUTION CODE.: B0010 COPIES RECEIVED:LTR L'NCL l'IIE'VP$TITLE'.Licensing Submittal:

PSAR/FSAR Amdts 8, Related Correspondence NOTESR RECIPIENT ID CODE/NAME NRR/DL/ADL NRR LB3 LA INTERNALS ACRS ELD/HDS1 IE/DEPKR/EPB 36.NRR ROK'gM~L NRR/DE/CEB 11 NRR/DE/EQB 13 NRR/DE/MEB 18 NRH/DE/SAB 24 NRR/DHFS/HFEB40 NRR/DHFS/PSRB NRR/DSI/AEB 26 NRR/DSI/CPS 10 NRR/DSI/IC8B 16 NRR/DSr/PSB NRA/DS I/RSB 231 RGfi2 EXTERNAL;BNL(AMDTS ONLY)LPUR 03 NSIC 05 COPIES LTTR ENCL 1 0 1 0 6 6 1 0 1 1 1 1 1 1 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 3 1 1 1 1 l 1 RECIPIENT ID CODE/NAME NRR LB3 BC BUCKLEYiB 01 ADM/LFMB IK FII E IE/DQAVT/QAB21 NRR/DE/AEAB NRR/DE/KHEB NRR/DE/GB 28 NRR/DE/MTKB 17 NRR/DK/SGKB 25 NRR/DHFS/LQB 32 NRR/DL/SSPB NRR/DS I/ASB NRR/DSI/CSB 09 NRR/DSI/METB 12 N'B 22 REG FI 04 AMI/MI B DMB/DSS,(AMDTS)

NRC-PDR 02 PNL GRUEL'gR COPIES LTTR ENCL*1 0 1 1 1 0 1 1 1-=1 0 1 1 2 2 1 1 1 1 1 1 1 0 1 1 1 1 1 1 1 1 0 1 1 1=1 1 TOTAL NUMBER OF COPIES REQUIRED;LTTR 51 ENCLt 4*4447 ita fr f,g ill h44>a f 14~*I i'I't>r 44~4 I k IJ r ll fl.>it a il, t'l It 4a a.lt 4 I q(a 4 ,I 14~h'-,.Jf,l h 4 a t~p~'~4~Il tti, 4 alger'Alt fp n4 r)4I tp'1 I l>v rf4,>I r~nO,I 4 4 4:r,i f C Jriili'l9wOa4~rtk!>V4~,f Ir,q).--N(e)AI<<h 4 Alp'1'4 t)'1 a f r calla~1 lt 4tt I C Jra'ial 4()pilir I')r 4 54 ag qa'a~ha g I gg4~Iratel'" rl f 4~l I ra$4 44'4~)tra r ee e e~)4~a4~h 4~f>: "I I V I I.')l,l th,J>4 tI')'fra f'4 te 8 14 , g I)frti.I4'"-l,l 5 4 4'1: I~I 3 r: 4>>~'>>r'rr~I4 r i I~<<c.aI>: jrrf4~4 I ag)t g tr~E<<g II'I II).,)I kl'I kt X Q I4 ll O'L it<<t a X-ir.XR~II I I X Qb.t 44++t',)4I4;X ICihta44tll (INES it~lr~"4 C 4eh J r<r'.I'XI.'ill"'41."..')X I 4,'>$~t>f,s ght IQ JI~hhi, I 4I$H 4)I%1')4<I'X-'4 I f a I I I r I r.T4>l:ll'4 4~Ct (Y I'gfT'ht4)-tv'I"'4'I" 3 4 4 Itl I I f I~I Ll tr I 4'" 4'iI ti I f

@MD Carolina Power gr Light Company MAY 0 7 1S85 SERIAL: NLS-85-119 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 ROD DROP ANALYSES

Dear Mr.Denton:

Carolina Power R Light Company hereby submits additional information on the Shearon Harris Nuclear Power Plant (SHNPP)Safety Evaluation Report (NUREG-1038)

License Condition No.7.This License Condition restricts operations above 9096 power until the rod drop analyses for SHNPP have been completed.

The analyses for SHNPP Cycle 1 have been completed using the methodology described in WCAP-10297-P,"Dropped Rod Methodology for Negative Flux Rate Trip Plants." The results of the analyses indicate that the thermal limits will not be exceeded.These results are included in the Draft FSAR pages included as Attachment 1.These Draft FSAR pages will be formally incorporated into the FSAR in a future amendment.

If you have any questions, please contact Mr.Gregg A.Sinders at (919)836-8168.GAS/ccc (1357GAS)S..merman Manager'uclear Licensing Section Cct Mr.B.C.Buckley (NRC)Mr.G.F.Maxwell (NRC-SHNPP)

Dr.3.Nelson Grace (NRC-RII)Mr.Travis Payne (KUDZU)Mr.Daniel F.Read (CHANGE/ELP)

Wake County Public Library Mr.Wells Eddleman Mr.3ohn D.Runkle Dr.Richard D.Wilson Mr.G.O.Bright (ASLB)Dr.3.H.Carpenter (ASLB)Mr.3.L.Kelley (ASLB)8505i40405 850507 PDR ADOCK 05000400 E PDR 411 Fayettevitte Street o P.O.Box 1551 o Raleigh, N.C, 27602

~~ts 4 1 TABLE 15.0.3-2 (Continued)

Faul ts 15.4 keactivity and Power Distr ibution Anomalies Reactivity Coefficientsa Assumed Moderator Moderator Computer Temperature Density Codes Utilized (Ak!F)(hklgmlcc)

Doppler Initial NSSS Thermal Power Out>put Assumedb (Mvt)Uncontrolled rod cluster control assembly bank withdraval from a sub-.critical or low power startup condition TMINKLE, FACTRAN, TNINC Refer to Section 15.4.1 0.43 lovers Uncontrolled rod cluster control assembly bank withdraval at power kod cluster control 8)I soPAg Tiolv LOF TRAN THING~%8ÃghR>>LOFTRAN~~9 0 and 0.43 lovers~and uppera i 2785 2785 Startup of an inactive reactor coolant loop at an incorrect temperature LOP TRAN, FAG TRAN TNINC 0.43 lovers 1671 Chemical and volume control system malfunction that results in a decree'se in the boron concentration in the reactor coolant NA NA NA 2785 Inadvertent loading and LEOPARD, operation of a fuel assembly TURTLE in an improper position NA NA 2785 I ROUE><b Ig RC FCRCNCE

'SHNPP FSAR 15@Oo 5 ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS*

The negative reactivity insertion following a reactor trip is a functioa of the acceleratfoa of the rod cluster control assemblies and the varfatfon fn rod worth as a" function of rod position.With respect to accident analyses, the critical parameter is the time of insertion up to the dashpot entry, or approximately 85 percent of the rod cluster travel.The rod cluster coatrol foa versus time assumed in accident analyses is shown in Fi ure 15.0.5-1.The rod cluster control assembl insertion time to dashoot g+~<entry is taken as 3.0 seconds unless otherwise aoted in the discussion The z<<niou use of such a long insertion time provides the most conservative results for all accidents aad fs intended to be applicable to all types of rod cluster R<<A control assemblies which may be used throughout plant life.Drop time testing equirements are dependent on the type of rod cluster control assemblies ctually used in the plant and are specified in the plant Technical Specifications 7~R<ch NisOpchw-.roe E~<A 2 2$o NS g7ig~7aM'igure 15.0.5-2shows the fraction of total negative reactfvity insertion d NP>>EA-versus normalfzed rod posftfon for a core where the axial power distribution fs skewed to the lower region of the core.-An axial distribution which is skewed to the lower region of the core can arise from an unbalanced xenon distribution This curve is used to compute the negative reactivity insertion versus time following a reactor trip.This negative reactivity insertion curve fs input to all point kinetics core models used in transient analyses.The bottom skewed power distribution itself is not an input into the point kinetics core model.There is inherent conservatism fn the use of Figure 15.0.5-2 in that it is based on a skewed flux distribution which would exist relatively infrequently.

For cases other than those associated with unbalanced xenon distributions, significantly greater negative reactivity wou1d have been inserted due to the more favorable axial distribution existing prior to the trip The normalized rod cluster control assembly negative reactivity insertion versus time is shown in Figure 15.0.5-3.The curve shown in this figure was obtained from Figures 15 0.5-1 and 15.0.5-2.A total negative reactivity insertion following a trip of 4 percent Ak is assumed fn the transient analyses except where specifically noted otherwise.

This assumption is conservative with respect to the calculated trip reactivity worth available as shown in Table 4.3 2-3 For Figures 15.0.5-1 and 15.0.5-2, the rod cluster control assembly drop time is normalized to 3.0 seconds, unless otherwise noted for a particular event, in order to provide a bounding analysis for all rod cluster contiol assemblies to be used in the SHNPP cores, as previously stated.The normalized rod cluster control assembly negative reactivity insertion versus time curve for an axial power distribution skewed to the bottom (Figure 15.0.5-3)fs used fn those transient ana1yses for which a point kinetics core model is used.Where special analyses require use of three*See 15 0 15.0.5-1 r I SHNPP FSAR dimensional or axial one dimensional core models, the negative reactivity insertion resulting from the reactor trip is calculated directly by the reactor kinetics code and is not separable from the other reactivity feedback effects.In this case, the rod cluster control assembly position versus time of Figure 15.0.5-1 is used as code input.15.0 5-2 SHNPP FSAR pq i Soph//Vi gnJ ie 15 4m3ml Identification of Causes and Accident Descri tion g,<~p R,RTyoH a)@~pe'>~~~Z c'SAm~~WoccP b)A dropped full length assembly bank.gc g~taticall

~nisali ned full 1~en th e~ssembl.d)Withdrawal of a single full length assembly.Each RCCA has a position indicator channel which displays position of the assembly.The displays of assembly positions are grouped for the operator's convenience Fully inserted assemblies are further indicated by a rod at bottom signal, which actuates a local alarm and a Control Room annunciator Group demand position is also indicated.

Full length RCCAs are always moved in preselected banks, and the banks are always moved in the same preselected sequence Each bank of RCCAs is divided into two groups The rods comprising a group operate in parallel through multiplexing thyristors

+e two groups in a bank move sequentially such that the first group is always within one step of the second group in the bank.A definite schedule of actuation (or deactuation of the'stationary gripper, movable gripper, and lift coils of a mechanism) is required to withdraw the RCCA attached to the mechanism.

Since the stationary gripper, movable gripper, and lift coils associated with the RCCAs of a rod group are driven in parallel, any single failure which would cause rod withdrawal would affect a minimum of one group Mechanical failures are in the direction of, insertion, or immobility.

The dropped assembly, dropped assembly bank, and statically misaligned assembly events are classified as ANS Condition II incidents (faults of moderate frequency) as defined in Section 15.0.1~The single RCCA withdrawal incident is classified as an ANS Condition III event, as discussed below.No single electrical or mechanical failure in the Rod Control System could cause the accidental withdrawal of a single RCCA from the inserted bank at full power operation The operator could deliberately withdraw a single RCCA in the control bank since this feature is necessary in order to retrieve an assembly should one be accidentally droppeds The event analyzed must result from multiple wiring failures (probability for single random failure is on the order of LO 4/year (refer to'Se'ction 7~7~2.2)or multiple operator actions and subsequent and repeated operator disregard of event indication.

The probability of such a combination of conditions is low;however, the limiting consequences may include slight duel dasage.~Asia~,:."R" gc.cokDAHC 5 H~~lhiji'palectM'jjjj'ng criterion is~in ejdogmcdsni eid General Design Criterion jC)?d5 uhi'cji'tates; lhe protection system shall be designed to assure that".~p'ecified acceptable fuel design limits are not exceeded for'any single'malfunction of the reactivity control systems, such as accident withdrawal sl)lilt15.4.3-1 0

zl Thus, consistent with the philosophy and format of ANSI N18.2, the event is c3assified as a Condition III event.By definition"Condition III occurrences include incidents, any one of which may occur during the lifetime of a particular plant", and"shall not cause more than a small-fraction of fuel elements in the reactor to be damaged..." er<E P~<P~=r~'~-I A/0 CARVE~SHNPP PSAR (not e)ection or dropout)of control rods." It has been shown that single failures resuLting in RCCA bank withdrawals do not violate specified fuel design limits Moreover, no single malfunction can result in the withdrawal of a single RCCA.Thus, it is concluded that criterion established for the single rod withdrawal at power is appropriate and in accordance with GDC 25.A dropped assembly or assembly bank is detected by: a)Sudden drop in the core power level as seen by the Nuclear Instrumentation System.b)Sspnnetric poser distribution as seen on out of-co-re neutron detectors or core exit thermocouples c)Rod at bottom signal.d)Rod deviation alarms e)Rod position indication.

Misaligned assemblies are detected by: a)Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples b)Rod deviation alarm.c)Rod position indicators The deviation alarm alerts the operator to rod-to-rod deviations within the same bank in excess of 12n5 inches.If the rod deviation alarm is not operable, the operator is required to take action as required by the Technical Specifications.

If one or more rod position indicator channels should be out of service, operating instructions shall be followed to assure the alignment of the nonindicated assemblies.

The operator is also required to take action as required by the Technical Specifications.

The operating instructions require selected pairs of core exit thermocouples to be monitored in a prescribed time sequence and following significant motion of the nonindicated assemblies.

The operating instructions also call for the use of movable in-core neutron detectors to confirm core exit thermocouple indication of assembly misalignment In the extremely unlikely event of simultaneous electrical failures which could result in single RCCA withdrawal, rod deviation and rod control urgent failure would both be displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the assemblies in the bank<,, We urgent failure alarm also inhibits automatic rod motion in the group 4n which fc occurs'ithdrawal of a single RCCA by operator action, w er.delibeffte,or by a combination of errors, would result in activation of" e'same alarm and the same visual indications Withdrawal of a single RCCA results in both positive reactivity insertion tending to increase core 15.4.3-2 0 gf, SHNPP FSAR powc r, and an increase in local power density in the core area associated with thl SCCA.Automatic protection for this event is provided by the overtemperature bT reactor trip, although due to the increase in local power density it is not possible in all cases to provide assurance that the core safety limits will not be violated.Plant systems and equipment which are available to mitigate the effects of the various control rod misoperations are discussed in Section 15'.8 and listed kn Table 15 0.8-1.No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.15 4~3e2 Anal sis of Effects and Conse uences The dropped rod event is an ongoing generic issue with the Nuclaar~egulatory Commission..Reanalysis of this event will a~once the core design is finalized.

SH onsiders this to be a confirmatory item.14 a)Dropped assembly, dropped assembly bank, and statically misaligned assembly Hethod Anal sis state power distributions are analyzed using the TURTLE~(Reference 3-1).The LOFTRAN Code (Reference 15m%~2"I)is used for the transient response to rapped RCCA or RCCA nk.The code simulates the neutron kinetics, RCS, pressurize

, s urizer power operated relief and safety valves, pressurize~r s ra , steam ge or, and steam generator safety valves'he code computes pertinent plant variable~eluding temperatures,: pressures, and er level.The system transient responsee tom LOFTRAN, along w'l.th+h"'a ing factors from TURTLE, are then used as input to ttie INC Code (S on 4')which calculates the DNBR.s v~~~i J S g<~<<sr Bs.PlK bW)Dropped RCCA Qc4ey-A dropped RCCA~typically results in a reactivity insertion which will be detected by the po~er Qg+47$ppgprangenegativeneutron f lux rat'e trip circuitry.The reactor is tripped within approximately'.2 5 seconds following the drop of a RCCAy88pJA', The core is not adversely affected during this period, since power is ,'$00~m decreasing rapidlye'ollowing reactor trip, normal shutdown procedures may subsequently be followed to further cool down the plant.Z'~~7""Q"~eQ Statically Hisaligned RCCA-The most severe misalignment situations wit/,pgsggcg to DgBg at significant power levels arise from cases in whic~anK 5 ih fuTiy fnserle8.with one'RCCA fully withdrawn; Hultiple independent alarms, including a bank insertion limit alarm, alert the operator well before the postulated conditions are approached.

The bank can be"inserted to its insertion limit with any one assembly fully withdrawn..~-'.,"."without the DNBR falling below~R4 f'mp'g vc/~e, The insertkon limits in the Technical Specif'ications may vary from time to time depending on a number of limiting criteria.It is preferable, therefore, to analyze the misaligned RCCA case at full power for a a.One or more dropped RCCAs from the same group For evaluation of the dropped RCCA event, the transient system response is calculated using th LOFTRA e N code.The code simulates the neutron kinet'e scs, eactor Coolant System, pressurizer, pressurizer relief d f r'an sa ety valves, pressur-izer s ra ste p y, am generator, and steam generator safety valves.The code corn computes pertinent plant variables includ-ing temperatures, pressures, and power level.II CoN 7/II 0'v HCXt IWC

~/I t~F7 State ints are calculated and nuclear models are used to obtain a hot channel factor consi stent wi th the primary system conditions and reactor power.By incorporating the primary conditions from the transient and the hot channel factor from the nuclear analysis, the DNB design basis is shown to be met using the THING code.The transient response, nuclear peaking factor ana'lysis, and DNB design basis confirmation are performed in accordance with the methodology described in Reference l5.f.8-5.b.Statically Misaligned RCCA Steady state power distribution are analyzed using the com-puter codes as described in Table 4.1-2.The peaking ,factors are then used as input to the THINC code to calcu-late the DNBR.Results a.One or more Dropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion which may be detected by the power range negative neutron flux rate trip circuitry.

If detected, the reactor is tripped within approximately

2.5 seconds

following the drop of the RCCAs.The core is not adversely affected during this period, since I power is decreasing rapidly.Following reactor trip, normal shutdown procedures are followed.The operator may manually retrieve the RCCA by following approved operating procedures.

For those dropped RCCAs which do not result in a reactor trip, power may be,reestablished either by reactivity feed-back or control bank withdrawal.

Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition.

The equilibrium process without gt pg gp g cQfgp//vgJ QpJ N6x7%s'C 2636Q:I 0~'I'I

~~(i/I ggS 6'QT (V~t~~~4*t~~I~y~'ontrol system interaction is monotonic, thus removing power overshoot as a concern, and establishing the automatic rod control mode of operation as the limiting case.I For a dropped RCCA event in the automatic rod control mode, the Rod Control System detects the drop in power and ini-tiates control bank withdrawal.

Power overshoot may occur due to this action by the automatic rod controller after which the control system will insert the control bank to restore nominal power.Figure 15.4.3-1 shows a typical transient response to a dropped RCCA (or RCCAs)in automatic control.Uncertainties in the initial,condition are included in the DNB evaluation as described in Reference)Q,It,e-+In all cases, the minimum DNBR remains above the limit value.t(Tg&eRT Any action e p ant in a stabll required of the operator to maintain th ized condition will be i min be in a time frame in excess of t'nutes following the incident.be i'o en SHNPP CESAR position of the control bank as deeply inserted as the criteria on minimum DNBR and power peaking factor (see Section 4')vill allow.The.full power insertion limits on control bank D are then chosen to be above that position and vill usually be dictated by other criteria.Detailed results will vary from cycle to cycle depending on fuel arrangements.

For the RCCA misalignment with Bank D inserted to its full pover insertion limit and one RCCA fully vithdravn, DNBR does not fall below~TKis caS&'vas analyzed starting at 102 percent of full power,'ominal RCS pressure-30 psia, nominal RCS cells.M uu increased radial peaking factor associated with the misaligned RCCA.DNB calculations have not been performed specifically for assemblies missing from other banks;however, power shape calculations.have been done as required for the RCCA egection analysis.Inspection of the power shapes shows that the DNB and peak kW/ft.situation is less severe than the Bank D case discussed above assuming insertion limits~on the other banks equivalent to a Bank D full-in insertion limits 7¹'<R r'D" sw~~r E.'~DNB does'not occur for the RCCA misalignment incidenp The peak fuel temperature corresponds to a linear heat generation rate based on the radial peaking factor penalty associated with the.misaligned RCCA~nd the design axial pover distribution.

The resulting linear heat generation is veil below that vhich would cause fuel melting.Following the identification of an RCCA misalignment condition by the operator, the operator is required to take action as required by the plant technical specifications and operating instructions.

b)Single RCCA vithdrawal Hethod Anal sis w~r distributions within the core are calculated by the TURTLE (Referen.3-1)based on macroscopic cross sectio en@i'ated by LEOPARD (Reference 15.4.3-~~~

aking factor~a ated by TURTLE are then used by THINC to calculate the mini~urn BKggr the event.The case of the worst rod vithdrawn from Bank~nserted at the in limit, with the reactor initially at~power, was analyzed.This incident med to occur at beginn'ing-of-life since this results in the minimum value of mo e'ra temperature to flatten th i For the single rod vithdrawal event, tvo cases have been considered as'o1.1ows: Ifhe reactor is in the manual control mode, continuous withdrawal of a single RCCA results in both an increase in core power and reactor coolant temperature, and an increase in the local hot channel factor in the area of the withdraving RCCA.In terms of the 15.4.3-4 Amendment No.14 For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value.This case is analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values, including uncertainties (as given in Table 15.0-3)but with the increased radial peaking factor a8sociated with the mis-aligned RCCA.and thus the ability of the primary coolant to remove heat from the fuel rod is not reduced.+N5cR 7 Power distributions within the core are calculated using the computer codes as described in Table 4.1-2.The peaking factors are then-used by THINC to calculate the ONBR for the event.The case of the worst rod withdrawn from bank D inserted at the insertion limit, with the reactor initially at full power, was analyzed.This incident is assumed to occur at beginning-of-life since this results in the minimum value of moderator tem-perature coefficient.

This assumption maximizes the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution.

SHNPP FSAR overalI.system response, this case is similar to those presented in Section 15',2;however, the increased local power peaking in the area of the withdrawn RCCA results in lower minimum DNBRs than for the withdrawn bank cases.Depending on initial bank insertion and location of the withdrawn RCCA, automatic reactor trip may not occur s>>fficiently fast to prevent the minimum core DNBR from falling below<~<Li~TVai~&~~

FValtZatiOn Of thiS Caae at the pOWer and COOlant COnditiOnS at which the overtemperature AT trip would be expected to trip the plant shows that an upper limit for the number of rods with a DNBR less than Ehe./i~i Qadi (id C9$8$is 5 oercent.2)lf the reactor is in the automatic control mode, the multiple failures that result in the withdrawal of a single RCCA will result in the immobility of the other RCCAs in the controlling bank.The transient will then proceed in the same manner as Case 1 described above.(T QnLLPG Q ftl I For the above cases a reactor trip will result, although not sufficiently fast in all instances to preventa minimum DNBR in the core of less than Following reactor trip, normal operating procedures may be 7ollowed to Barther cool down the plant 15.4.3.3 Conclusions 11 cases of dropped banks, the reacto~~~ed<~he-pew negative neu~nconse centi the DNBR desi ctitetion GS d in Section 4.4 is met.~MS'For all cases of any bank inserted to its rod insertion limits with any single RCCA in that bank fully withdrawn (static misalignment), the DNBR remains~greater thank<he.I'ws f vc-/~e.For the case of the accidental withdrawal of a single RCCA, with the reactor in the automatic or manual control mode and initially operating at full power with Bank D at the insertion limit, an upper bound of the number of fuel rod.experiencing DNB~Cggt is 5 percent of the total fuel rods in the core.~Lns s, J eel'5.4.3-5 Amendment No.14 For cases of dropped RCCAs or dropped banks, for which the reactor is tripped by the power range negative neutron flux rate trip, there is no reduction in the margin to core thermal limits, and consequently the DNB design basis is met.It is shown for all cases which do not result in reactor trip that the DNBR remains greater than the limit value and, therefore, the DNB design is met.

SHNPP FSAR~TABLE l5.4.3-1'INIMUM PREDICTED DNBR FOR CASES OF ROD CLUSTER CONTROL EMBLY~MISALIGNMENT AND DROPPED ROD CLUSTER CONTROL ASSEMBLY Bank D at insertion limit, D-12 fully vithdrawn (RCCA misalignment)

Dropped R H-12 dial Po Pitkin Za~or (FAH)<1 72<1.72 Minimum DNBR>1~3)1 3 15.4.3-6 I A P'l 1.2000 1.1000 1.0000.90000.80000.70000.60000.50000 Cl ID C)C)C)TiHE<SEC)1.2000 1~1000 1.0000.90000.80000.70000.60000.50000 Ct ED CI CI CD Q CI C)C)C7 TIHE (SEC) 600.00 LIJ UJ LC I CA I-CZ: LO D LCJ o o 580.00 560.00 510.00 520.00 500.00 CD C7 C7 C7 CI CI TIHE (SEC)2I00.0t 2300.0 UJ CC C/l Cjl LIJ CZ: Q IJJ hJ Q CA Ctl LIJ CL 2200.0'100.0 2000 0 1900.0 1800.0 CI CI C)C7 C)C7 C)TlHE (SEC)FIGURE 15.4.3-1 (Continued)

~E

REFERENCES:

SECTION 15'15.4.1-1 Risher, D.H., Jr.and Barry, R.F.;"TWINKLE>>A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-A (Pxoprietary) and WCAP-8028-A (Non-Proprietary), January 1975 15 4.1-2 15 4.2-1 Hargrove, H.G.,"FACTRAN-A Fortran-IV Code for Thermal Transients in a U02 Fuel Rod," WCAP-7908, June 1972.Burnett, T.W.T>>, et al.,"LOFTRAN Code Desex'iption," WCAP-7907, June 1972 15.4.2-2"Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974 15.4.3-1 Barry, R.F.and Altomare, S.,"The TURTLE 24.0 Diffusion Depletion Code," WCAP-7213"A (Proprietary) and WCAP-7758-A (Non-Proprietary), January 1975.15.4.3-2 Barry, R.F.,"LEOPARD-A Spectrum Dependent Non"Spatial Depletion Code for the IBM-7094," WCAP-3269-26, September 1963.15.4.8-1 Risher,.D H., Jr.,"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision l-A, January 1975 15 4 8-2 Taxelius, T G (Ed),"Annual Report-Spert Pro)ect, October, 1968, Septembex', 1969," Idaho Nucleax Corporation IN-1370, June 1970.15.4.8-3 Liimataninen, R.C.and Testa, F.J',"Studies in TREAT of Zircaloy-2"Clad, U02More Simulated Fuel Elements'NL 7225'anuary>>June 1966, P.177, November 1966.15.4.8-4 Bishop, A.A., Sandburg, R.O.and Tong, L.S."Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August 1965>>I5,".8+Morita, T., et.al.,"Dropped Rod Hethodology for Negative Flux Rate Tr ip Plants," MCAP-10297-P-A (Proprietary) and MCAP-10298-A (Non-Proprie-tary), June 1983.