ML17262B120

Re Ginna Boric Acid Storage Tank Boron Concentration Reduction Study.

ML17262B120
Person / Time
Site:	Ginna
Issue date:	11/30/1992
From:	MCHUGH C J, SPRYSHAK J J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ATTACHMENT C YENDOR'S DOCUMENT REVIEVI I g~Mfg sexy proceed 2 Q Approved gobrctt gott doA.Mfg orty proceed 3 Q Approved except es noted Mete chsogcs ecd sobroit Bott dodh Mfg orey proceed ss cpprovcd C Q stot sptvovcd Correct eod resobea, 5 Q Revie>>oot rertvhcd Ilfg, rosy proceed Approvxl of this docorored docs ool regevo sreoher freer fvg ccophsrce with ooratrect or poreheso order reqrrtrecertL By Dxte ROCHESTER GAS 8 ELECTRIC CORP.ftOCHESTER, NY R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study November 1992 Prepared by: c..zW Christopher J.McH h J eph.Spryshak Transient Analysis II Westinghouse Electric Corp.Containment Design and BWR Technology Westinghouse Elec ric Corp.9', Pgq 9~8o 921Zao p , AD<<id OrOOogg4.PVR R.E.Glnna Boric Acid Storage Tank Boron Concentration Reduction Study.0 This report presents analyses of the R.E.Ginna plant steamline break (SLB)containment integrity and associated LOCA-related analyses, with a reduction of the boron concentration in the Boric Acid Storage Tanks (BASTs)from 20,000 ppm to 2,000 ppm.A boron reduction to this level will allow the removal of credit for the BASTs from the licensing basis accident analyses (and subsequently removal of the associated heat tracing required).

The BASTs will be retained for operation requirements and redundant flow paths as discussed in Technical Specifications.

'o~'R.E: Ginna currently must maintain 20,000 ppm boron in the BASTs, which requires heat tracing to prevent boron precipitation.

The BASTs and their heat tracing are part of the Safety Injection (SI)system and thus they must be maintained according to requirements which can impose operational restrictions.

The only accident analyses which are significantly affected by boron concentration reduction are the secondary side steamline break transients.

The core and the containment responses are affected by the steamline break transients and therefore were considered in the boron concentration reduction analysis.2.0 co-The SLB core response analysis is documented in Reference 1 and supports a reduction in the BASTs boron concentration to 2000 ppm.3.0 NTAINMENT INTE RITY ANALYSIS 3.1 Pur~seThe purpose of the Containment Integrity Steamline Break analysis is to demonstrate the acceptability of the Containment Safeguards Systems to mitigate the consequences of a hypothetical rupture of a steamline pipe.The impact of steamline mass and energy releases on containment pressure is addressed to ensure the containment pressure remains below its design pressure of 60 psig at the reduced boron concentration conditions.

I t>i 0 ,gr~i~n~Z'(p R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study 3.2 Relevant Acce tance riteria~:" The BASTs are components of the Safety Injection'System designed to mitigate'the consequences.

of postulated steamline break accidents by providing a high concentration of boric acid to the reactor coolant.A high concentration of boric acid causes a decrease in the post trip return core power level and subsequently a decrease in heat transferred to the secondary side fluid, which results in decreased containment pressures during a SLB.The containment pressures resulting from the mass and energy releases must remain below the design pressure of the containment building.For R.E.Ginna the containment design pressure is 60 psig.3.3 Evaluation

3.3.1 Methodology

Calculation of the steamline break containment response is a two step process.The LOFTRAN computer code (Reference 2)is first used to calculate the mass and energy released as a function of time.The releases are then used as input to the COCO code (Reference 3)to calculate containment pressures and temperatures as a function of time.Attachment 1 provides a brief description of the LOFTRAN and.COCO codes.1 The cases that were analyzed for peak containment pressures are listed in Table 1.The basic initial conditions, heat sink model, fan cooler data, and containment spray parameters for these cases are outlined in Tables 2 through 5.The following conservative assumptions are made for the mass and energy release analysis: 1.Maximum decay heat equivalent to the 1979 ANS decay heat+2e uncertainty.

2.No credit for water entrainment in the blowdown results.3.Conservatively high values for reverse steam generator heat transfer.4.Conservative moderator temperature coefficient for the rodded core at end-of-life.

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R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study 3.3.2 Spectrum of Break Assumptions A comprehensive set of break sizes, initial power levels, single failure assumptions, and off-site power availability must be considered so that there is reasonable assurance that the limiting cases have been covered.The complete set of steamline break cases that were addressed for the R.E.Ginna plant is listed in Table 1.The single failures considered in this analysis have been selected based upon their potential for increasing the amount of mass and energy released into containment or for reducing the amount of heat removed from containment.

The four postulated failures are as follows: Failure of the Main Steam Isolation Valve (MSIV)to close Failure of the Feedwater Control System (FCS)Failure of one containment spray pump (to operate)Failure of one diesel generator to start The breaks considered include 4.37 ft'ouble Ended Ruptures (DER)upstream of the flow restrictor, 1.4 ft'ER's downstream of the flow restrictor, and small breaks of 1.1 ft'r smaller.To determine the limiting break size for the small breaks, several cases were run with break sizes from 0.3 ft'o 1.1 ft'n 0.2 ft'ncrements.

After it had been sufficiently demonstrated that the two largest of the small breaks consistently resulted in higher break flows and limiting peak containment pressures, the remainder of the small break cases were run with only the two largest break sizes, 0.9 ft'nd 1.1 ft'.3.3.3 Consistent Off-Site Power Availability One of the conservative assumptions that has historically been made is with respect to the availability of off-site power.Under typical SLB containment analysis methodology, the mass and energy releases are generated assuming off-site power continues to be available for the duration of the transient.

This gives maximum primary-to-secondary heat transfer because of the forced reactor coolant flow from the Reactor Coolant Pumps (RCPs).The containment integrity calculation is then performed assuming that off-site power is not available, which extends the safeguards equipment startup delays due to diesel sequencing timing.These two assumptions contradict each other, but result in an analysis which bounds both with r I k g A3<<P'I+v l"'A'~at),

R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study.and without,off-site.

power, with one case.To remove this unnecessary conservatism the limiting cases were analyzed with a consistent off-site power availability assumption.

A small number of cases were analyzed with inconsistent assumptions and demonstrated a high margin to the pressure limit.3.3.4 Mass and Energy Calculation Assumptions 3.3.4.1 Main and Auxiliary Feedwater Flow as a Function of Steam Generator Pressure The cases presented in this study assumed a main feedwater flow rate as a function of both the steam generator pressure and the feedwater control valve position.The feedwater control valve (FCV)position varies with power level and postulated break location.The break location affects the FCV position in'hat a'steamline break results in an increase in steam flow and subsequently a steam flow/feed flow mismatch.In response to the mismatch, the feedwater control system is assumed to increase feed flow to match steam flow.The typical analysis assumption is to assume that the faulted loop FCV is wideHowever, with a break upstream of the steamline flow restrictor, no steam flow/feed flow mismatch would be present.As such, for cases with a break size larger than the flow area through the flow-restrictor, it is assumed that no mismatch signal is present and that the faulted loop FCV stays in its nominal pre-break-position.

-The intact loop FCV is conservatively-assumed to remain in its nominal pre=-break position until reactor trip.A turbine trip is assumed to occur at the same time as reactor trip and the intact loop FCV is assumed to close instantly in response to the decrease in steam demand.For steamline breaks located downstream of the flow restrictors and those breaks having a break area smaller than the flow restrictor, it is assumed that the FCV on the faulted loop goes wide open in response to the increased steam flow.As with the upstream breaks, the intact loop FCV is assumed to be in its nominal position initially and closes instantly, coincident with reactor trip.Auxiliary feedwater flow rates as a function of steam generator pressure were also assumed in the analyses.,Auxiliary feedwater flow rates varied depending on the availability of offsite power and the single failure being evaluated.

At HZP,the main feedwater pumps will not deliver feedwater to either steam generator.

Thus, none of e\~E~

R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study the zero power cases assume any main feedwater.

These cases assume auxiliary feedwater only, initiated at the time the steamline break occurs.3.3.4.2 Core Reactivity Coefficients LOFTRAN utilizes a point kinetics model, which uses reactivity feedback coefficients to calculate the kinetics conditions in the core.Steamline break transients initialized at hot-zero power.assume rodded-reactivity feedback coefficients with an allowance for the most reactive Rod Cluster Control Assembly (RCCA)stuck in its fully withdrawn position.Steamline break transients initiated with the reactor at power typically assume End-Of-Life (EOL)reactivity coefficients calculated assuming that all RCCAs are fully withdrawn.

However, for these analyses, since the majority of the transient is post reactor trip, rodded coefficients (again with an allowance for a stuck RCCA)were assumed.Confirmation of the conservatism of the overall reactivity model has been obtained by more detailed core neutronics calculations.

3.3.5 Containment

Integrity Assumptions The major containment integrity calculational assumptions used with COCO are as follows: 1.The mass and energy release to the containment is for-a break opening time of zero.2.The saturation temperature corresponding to the partial pressure of the containment vapor is used in calculating the condensing heat transfer to the passive heat sinks and the heat removal by containment fan coolers.3.The Westinghouse containment model utilizes the analytical approaches described in References 3 and 4 to calculate the condensate removal from the condensate film.A convective heat flux revaporization model is used for small breaks.100%revaporization is assumed for large breaks.4.The small steamline break containment analyses utilized the stagnant Tagami correlation, Reference 5.

4 4~rr R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study 5.The diesel failure conditions (minimum safeguards), that were modeled, assumed that there were 2 fan coolers and one containment.

spray-pump (1300 gpm)were operating;

-The time--delays that were assumed for initiation of containment sprays and fan coolers with a diesel failure are given in Table 3.3.4 Desi n-Basis Containment lnte rit Anal sis Results.Figures 1 and 2 provide the pressure and temperature transient curves for the 4.37 ft'ER upstream of the flow restrictor case producing the highest peak containment pressure of this type of break and all other breaks analyzed.This case represents a main steam isolation valve (MSIV)failure at 30%power with offsite power available.

The BASTs boron concentration of 2000 ppm was assumed in this case and all other cases identified in Table 1.The mass and energy releases for this case are shown in Figures 3 and 4.The limiting 1.4 ft downstream DER containment pressure and temperature transients are shown in Figures 5 and 6.This case represents the feedwater control system failure at 70%power without offsite power available.

Note that the peak pressure is lower for the 1.4 ft'reak than for the 4.37 ft~break.The smaller break area reduces the blowdown mass and energy release rate without significantly delaying actuation of protective functions and, therefore, results in a lower peak containment pressure than the 4.37 ft'ase.The mass and energy release rates for this case are included in Figures 7 and 8.The limiting small DER is a 1.1 ft'reak, resulting in the pressure and temperature transients shown in Figures 9 and 10.This case was analyzed assuming a diesel failure at 102%power, without offsite power available.

The mass and energy release rates for this case are included in Figures 11 and 12.The containment pressures reached by the limiting breaks with the boric acid storage tank concentration of 2000 ppm remain below the containment design limit of 60 psig.

3'i C JiE k~', I fI ax e R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study 4.0 EVALUATION OF LOCA-RELATED ANALYSES 4.1 Lar e Break LOCA The current Large Break Loss-Of-Coolant Accident (LBLOCA)analysis of record for R.E.Ginna was performed using the NRC-approved 1981 ECCS Evaluation Model, Reference 6.The proposed reduction in the boron concentration in the BASTs will not adversely affect the Large Break LOCA because the Evaluation Model codes used in analyzing the large break do not explicitly model boron concentration in the reactor coolant system.4.2 Small Break LOCA The current Small Break Loss-Of-Coolant Accident (SBLOCA)analysis of record for R.E.Ginna was performed using the NRC-approved Small Break LOCA ECCS Evaluation Model with WFLASH, Reference 7.The proposed reduction in the boron concentration in the BASTs will not adversely affect the Small Break LOCA because the Evaluation Model codes used in analyzing the small break do not explicitly model boron concentration in the reactor coolant system.4.3 Post-LOCA Lon Term Core Coolin Subcriticalit Re uirement The Westinghouse licensing position for satisfying the requirements of 10CFR 50.46 Paragraph (b)Item (5)"Long Term cooling" is defined in WCAP-8339, Reference 8.The Westinghouse commitment is that the reactor will remain shutdown by borated ECCS water residing in the sump following a LOCA, Reference 9.Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a concentration that, when mixed with other sources of borated and non-borated water, will result in the reactor core remaining subcritical assuming all control rods are out.The large reduction in boron concentration in the BASTs will have a significant effect on the Reactor Coolant System boron concentrations assumed for this calculation.

The calculations for determining whether the reduction in the boron concentration in the BASTs will

• a l ,~J gp 1~g44 I R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study result in the core remaining subcritical was re-done with a new concentration of 2000 ppm in the BASTs.A new RCS boron concentration curve for the 2000 ppm value was.generated and used in the core design , process to.ensure-that the core will remain subcritical with a boron concentration-of 2000 ppm.---4.4 Boron Preci itation Durin Lon Term Coolin The post-LOCA boron precipitation long term core cooling requirement ensures no boron precipitation in the reactor vessel following boiling in the core.Since Ginna has simultaneous injection from the residual heat removal safety injection system into the upper plenum and the high head safety injection system into the cold legs, this requirement is met by requiring alternate injection within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after a LOCA.This time is dependent on power level, and the RCS, RWST, accumulator, and other water sources volumes and boron concentrations.

A reduction in the boron concentration in the BASTs will have no effect on the power level, or volumes assumed for the RCS, RWST, accumulators, and other water sources.Although the boron concentrations will be affected, it requires an increase in the concentration to adversely affect the boron precipitation.

Since the boron concentration would be decreasing with the proposed change, there will be no adverse effect on the post-LOCA alternate injection requirement of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> for the R.E.Ginna plant.4.5 Post-LOCA Lon Term ore Coolin Minimum Flow Post-LOCA long term core cooling minimum flow is determined to ensure adequate flow for large break and small break at the time of recirculation switchover.

A reduction of the boron concentration in the BASTs will have no effect on the inputs for this calculation.

Therefore, this change will have no effect on the post-LOCA long term core cooling minimum flow for the R.E.Ginna plant.4.6 LOCA Summa and Conclusions The effect of reducing the boron concentration in the BASTs on the LOCA-related analyses for R.E.Ginna has been evaluated by Westinghouse.

The potential effect of the change on the UFSAR analysis results for each of the LOCA-related accidents was evaluated and it was shown in all cases that the effect of the change did not result in exceeding any of the following design or regulatory limits: 1.-.The calculated peak-fuel element cladding temperature is below the requirements of 2200'F.

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R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study2.-.-The amount of fuel element cladding that.reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.3.-The cladding-temperature transient is terminated at a time when the core geometry is still.amenable to cooling.The localized cladding oxidation-limit-of 17 percent is-not exceeded during or after quenching.

4.The core remains amenable to cooling during and after the break.5.The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.Therefore, it is concluded that the proposed modification to reduce the boron concentration in the BASTs is acceptable from the standpoint of the UFSAR accident analyses discussed in this section.5.0 Conclusions A reduction of the BASTs boron concentration to 2000 ppm at the R.E.Ginna plant will be acceptable.from the standpoint of core response, steamline break containment integrity,and LOCA=,evaluation.-.

0~I'g c".3K i II>f tl f~.t C~h' R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study References 1.RG&E to NRC Request for Amendment to Technical Specifications dated Oct.16, 1985, purpose-Revise Containment Internal Pressure Limitations.

2.Burnett, T.W.T., et.al.,"LOFTRAN Code Description," WCAP-7909-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.3.Bordelon, F.M.and Murphy, E.T.,"Containment Pressure Analysis Code (COCO)," WCAP-8327, July 1974.4.Hsieh, T.et al,"Environmental Qualification Instrument Transmitter Temperature Transient Analysis", WCAP-8936, February 1977 (Proprietary) and WCAP-8937, February 1977 (Non-Proprietary) 5.Jens, W.H., and Lottes, P.A.,"Analysis of Heat Transfer, Burnout, Pressure Drop, and Density Data for High Pressure Water", USAEC Report ANL-4627, 1951.n 6.WCAP-9220-P-A(Proprietary), WCAP-9221(Non-Proprietary),Eicheldinger, C.,"Westinghouse ECCS Evaluation Model-1981 Version", Revision 1, 1981.7.WCAP-8200 (Proprietary),"WFLASH-A FORTRAN-IV Computer Program For Simulation Of Transients In A Multi-Loop PWR", Esposito, V.J., et al., July 1973.8.WCAP-8339 (Non-Proprietary), Bordelon, F.M., et.al.,"Westinghouse ECCS Evaluation Model-Summary", June 1974.9.Westinghouse Technical Bulletin NSID-TB-86-08,"Post-LOCA Long Term Cooling: Boron Requirements", October 31, 1986.10

~r Table 1: Containment Integrity Analysis-Steam Line Break Cases CS-Failure of One Containment Spray Pump to Operate DIESEL-Failure of one Diesel Generator to Start MSIV-Failure of Main Steam Isolation Valve FCS-Failure of Feedwater Control System Case Break T e Break Size ft~Power%Failure M&E Containment Offsite Power 1A UPSTREAM DER 4.37 102 CS AVAIL AVAIL 1B DIESEL NOT AVAIL NOT AVAIL 2A MSIV AVAIL AVAIL.2B MSIV NOT AVAIL NOT AVAIL 3A FCS AVAIL AVAIL 3B FCS NOT AVAIL NOT AVAIL 4A 70 AVAIL AVAIL 4B DIESEL NOT AVAIL NOT AVAIL SA MSIV AVAIL AVAIL 5B MSIV NOT AVAIL NOT AVAIL 6A FCS AVAIL AVAIL 6B FCS NOT AVAIL NOT AVAIL 7A 30 CS AVAIL AVAIL 7B DIESEL NOT AVAIL NOT AVAIL MSIV AVAIL AVAIL 8B MSIV NOT AVAIL NOT AVAIL 9A FCS AVAIL AVAIL 9B FCS NOT AVAIL NOT AVAIL 10A AVAIL AVAIL 10B DIESEL NOT AVAIL NOT AVAIL 11A MSIV AVAIL AVAIL 11B MSIV NOT AVAIL NOT AVAIL 12A FCS AVAIL AVAIL 12B FCS NOT AVAIL NOT AVAIL MQE Mass and eneray released into containment b 81~

Table l continued Case Break e Break Size ft'ower%Failure M&E Containment Offsite Power 13A DWNSTRM DER 1.4 102 CS AVAIL AVAIL 13B DIESEL NOT AVAIL NOT AVAIL 14A MSIV AVAIL AVAIL 14B 102 MSIV NOT AVAIL NOT AVAIL 15A FCS AVAIL AVAIL 15B FCS NOT AVAIL NOT AVAIL 16A 70 CS AVAIL AVAIL 16B DIESEL NOT AVAIL NOT AVAIL 17A MSIV AVAIL AVAIL 17B MSIV NOT AVAIL NOT AVAIL 18A FCS AVAIL AVAIL 18B FCS NOT AVAIL NOT AVAIL 19A 30 CS AVAIL AVAIL 19B DIESEL NOT AVAIL NOT AVAIL 20A MSIV AVAIL AVAIL 20B MSIV NOT AVAIL NOT AVAIL 21A FCS AVAIL AVAIL 21B FCS NOT AVAIL NOT AVAIL 22A AVAIL AVAIL 22C DIESEL AVAIL NOT AVAIL 23A MSIV AVAIL AVAIL 23B MSIV NOT AVAIL NOT AVAIL 24A FCS AVAIL AVAIL 24C FCS AVAIL NOT AVAIL 25A1 25A2 25A3 25A4 SMALL DER 0.3 0.5 0.7 0.9 102 CS CS CS CS AVAIL AVAIL AVAIL AVAIL AVAIL AVAIL AVAIL AVAIL S~~a Table 1 continued Case 25A5 25B1 25B2 25B3 Break T e.SMALL DER 1.1 102 CS 0.3 102 DIESEL 0.5 0.7 DIESEL DIESEL Break Size ft'ower%Failure Offsite Power M&E AVAIL Containment AVAIL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 25B4 25BS 0.9 DIESEL DIESEL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 26A4 26AS 0.9 102 FCS FCS AVAIL AVAIL AVAIL AVAIL 26B4 26B5 0.9 102 NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 27A4 27AS 0.9 70 AVAIL AVAIL AVAIL AVAIL 27B3 27B4 27B5 0.7 0.9 70 DIESEL DIESEL DIESEL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 28A4 28AS 0.9 70 FCS FCS AVAIL AVAIL AVAIL AVAIL 28B4 28BS 0.9 70 FCS FCS NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 29A4 29A5 0.9 30 AVAIL AVAIL AVAIL AVAIL 29B4 29BS 0.9 30 DIESEL DIESEL NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 30A4 30AS 0.9 30 FCS FCS AVAIL AVAIL AVAIL AVAIL 30B4 30BS 0.9 30 FCS FCS NOT AVAIL NOT AVAIL NOT AVAIL NOT AVAIL 31A1 0.3 CS AVAIL AVAIL 13 II~I A t,t49.0 40.2 28.4" 19.5 Nominal Steam Generator Water Level (%NRS)52.0 52.0 52.0 52.0'The actual steam generator level at zero power is 39%NRS+uncertainties.

52%NRS+uncertainties was conservatively assumed in the analyses.r.<<"'v>,'NRS-=

Narrow Range.Span I Initial Condition Uncertainties Average RCS Temperature

=4'F Pressurizer Water Level=5%NRS Steam Generator Water Level=3.5%NRS (Some cases assumed 5%NRS) 0/ly A Table 3: Major Containment Assumptions Initial Pressure Initial Temperature Initial Humidity Containment Volume Containment Fan Coolers High-1 Setpoint Used Actual Setpoint Instrument Uncertainty Initiate on Heat Removal Rates With off-site power available Number of Fan Coolers Delay Without off-site power available Number of Fan,Coolers without Diesel Failure with Diesel Failure Delay Containment Sprays Flowrate per Spray Pump~RWST Water Temperature Pressure Setpoint Used Actual Setpoint Pressure Instrument Uncertainties 15.7 psi 120'F 20%1.0 E+06 k';~'.0 psig 4.0 psig 2 psl SI (or High-1 signal if earlier)Table 5 34.0 sec 44.0 sec 1300 gpm 80'F 32.5 psig 28.0 psig 4.5 psig 17 4 II'l I Table 3 (continued):

Major Containment Assumptions

.With Off-site Power Available Number of Spray Pumps Operating without containment spray failure with containment spray failure Delay without containment spray failure with containment spray failure Without Off-site Power Available Number of Spray Pumps operating without diesel failure with diesel failure Delay 27.3 sec 28.5 sec 45.5 sec Heat Sinks Table 4 18 Table 4: PASSIVE HEAT SINKS Wall Description Heat Transfer Area 2 Material Thickness ft 1.Insulated portion of dome and containment wall 36285.0 stainless steel insulation steel concrete 0.00158 0.1042 0.03125 3.364 2.Uninsulated portion of dome and containment wall 12370.0 steel concrete 0.03125 2.5 3.Basement floor 6576.0 concrete steel concrete 2.0 0.0208 2.0 4.Walls of sump A under sump level 8.24 steel concrete 0.0208 3.0 5.Wall of sump A over sump level 2052.75 steel concrete 0.0208 3.0 6.Floor of sumps A and B 366.0 concrete steel concrete 2.0 0.0208 1.0 7.Walls of sump B 189.0 concrete steel concrete 2.0 0.0208 1.0 8.Outer refueling cavity wall 9.Inner refueling cavity wall 5870.0 5870.0 concrete stainless steel concrete 0.0208 2.0 10.Bottom of refueling cavity 1143.0 11.Loop compartments (Loops A and B)'8846.0 stainless steel concrete concrete 0.0208 4.0 1.4115 12.Floor of intermediate level'672.0 concrete 0.25 13.Operating floor and structure on operating floor'5570.0 concrete 1.0 Table 4 (continued):

PASSIVE HEAT SINKS Wall Description Heat Transfer Area fthm Material Thickness ft 16 I-beam'592.0 17.I-beam, cylindrical supports for S.G.5536.0 and RCPs, and containment crane rectangular support columns 14.I-beam and beams for crane structure'120.0 15.I-beam and-beams for crane structure'458.0 steel steel steel steel 0.0625 0.03455 0.0217 0.0586 18.Containment crane rectangular support 342.0 columns steel 0.167 19.Beams for crane structure 236.0 steel 0.12 14000.0 20.Grating, stairs, misc.steel'teel

'-Area accounts for both sides of heat sink walls, thickness is half of actual thickness 0.0625 Thermo h sical Pro erties of Containment Heat Sinks Insulation Concrete Steel Stainless Steel Thermal Conductivity 0.0208 0.81 28.0 10.9 Volumetric Heat Capacity TU/ft~'F 2.0 31.5 54.4 60.0 f'II Table 5: Containment Fan Cooler Heat Removal Rates Containment Temperature deg F>>200 210 220 230 240 250 260 270 280 287 290 300 Group A: with Offsite Power Available BTU/hr(~10')

15.90 17.40 20.70 25.80 30.60 34.50 38.10 41.70 45.00 47.00 48.30 50.70 Group B: without Offsite Power Available BTU/hr(*10')

15.22 16.66 19.82'A.70 29.30 33.03 36.48 39.93 43.09 45.00 46.24 48.54 t~Y 7~k Table 6-Sequence of Events Accident Event Time sec 1.Main Steamline Break a.30%Power b.MSIV Failure c.4.36 ft break d.Offsite Power Available Steamline Break Occurs 0.0 Rod Motion Starts Steamline Isolation Occurs Feedwater Isolation Occurs Auxiliary Feedwater Starts Containment Sprays Start Fan Coolers Start 2.4 7.4 14.4 25.0 34.5 42.0 High 1 Containment Pressure Setpoint 1.0 (6.0 psig)is Reached Peak Containment Pressure is Reached 149 Auxiliary Feedwater is Terminated 600.0 Faulted Steam Generator Dries Out (i.e., mass releases stop)-610.0 22 I C F p V I f),

Table 6-Sequence of Events (continued)

Accident 2.Main Steamline Break a.70%Power b.FCS Failure c.1.40 ft'reak d.Offsite Power Not Available Event Steamline Break Occurs Time sec 0.0.SIS Low Steam.Pressure Setpoint (372.7 psia)reached 2.7 High 1 Containment Pressure Setpoint 3.8 (6.0 psig)is Reached Rod Motion Starts Feedwater Isolation Occurs Auxiliary Feedwater Starts Containment Sprays Start Fan Coolers Start 4.7 25.0 47.8 126.8 Peak Containment Pressure is Reached 569 Auxiliary Feedwater is Terminated 600.0 Faulted Steam Generator Dries Out (i.e., mass releases stop)-625.0 23 I~I A.l~k Table 6-Sequence of Events (continued)

Accident Event Time sec 3.Main Steamline Break a.102%Power b.CS Failure c.1.10 ft'reak d.Offsite Power Not Available Steamline Break Occurs High 1 Containment Pressure Setpoint (6.0 psig)is Reached Rod Motion Starts Auxiliary Feedwater Starts Feedwater Isolation Occurs Containment Sprays Start Fan Coolers Start Auxiliary Feedwater is Terminated 0.0 4.7 7.0 25.0 27.0 48.8 128.1 600.0 Faulted Steam Generator Dries Out (i.e., mass releases stop)-760.0 Peak Containment Pressure is Reached 762 24 r, P~P<a4%1 Containment Pr essur e (psigjC3 o~CO C)fU C)C)bJ C)C)C)C)ll tCI C CD 13 t: N CD C)CD o 0 I B tC)C)CO C)CD B C)C)CO C)25

'4*\+'t P(%11~:

Containment Temper atur e (degrees Fj A3 CD oo C)0 C)CA C)C)CD C)C)P3 C)C)C)0 0 fV CT)C)C)fV CD C)C)ll LD C CD o O CD o n O C)CO C)C)C)CO CO 26 Mass Release Rate (ibm/sec)O O 0 A3 O O O O O O CD O Ql O O O O CD O O O O O O O O O CV O O O O Tl ECI CD 43 O O O A3 O O O Ck b3 O rn CA Pl n O O O U)Ul O O O B CD CA O O O O O O 27 6 Energy Release Rate (E6 Btu/~<<<=j-o

>O OO fV O O CA O CD O O O fU O rl tlat 3 CD O O R3 O O O bJ O O rn M Pl O n 0'O Ul O O O V<CD Gl O O O O O O 28 Containment Pressure (psigj C)oo fU C)C)bJ C)C)CO C)Ul C)C)CT)C)C)3 CU o CD C)CD o 0 I (Cl C7 O U)CU C)CO C)

L'E I g'P V I%Mt l+E P Containment Temper atur e (degr ees Fj fU o oo Jh.o o CD o o 03 o o A3 o o o A3 A3 o o fV o o fV CD o o CV CD o o b3 o o o rl U3 C CD CD CD CU C CD (B O 3 I EC)C7 O CD CD o CD o 0 o o o o o o o

~l 4'Cl I V~~f%

Mass Release Rate{ibm/sec)O oo Ul O O O O C)O O Ul O O O fU O O O fU Ul O O 4 O ll tC)C: O O O fU O O O 43 O O rn CO Pl O 0 O Ul O CA O O O O O O 31 Energy Release Rate (E6 Btu/sec)O OO O Ul O O fV O ll tlat'CD O O O fO O O O.b3 O O rn M Ol O O O Ul O O O CD, Ol O O O O O O 32 Containment Pressur e (psigl o oo o o fV o o QJ o o o o Ul o o Ql o o ll ECI CD M CD o CD o C7 o o o o o o o 33 III 4 i E" A I'g1 ii f Containment Temperatur e{degrees Fj'C)C)oo (Sl C)C)P3 CO C)C)A3 Ul C)C)CO C)C)43 Ul CO C)ll tel C (D CD 13 CD CU CD i CD CD C)o CD A o I tlat C)CO C)CO C)CO C)34 Mass Release Rate{ibm/sec)O OO fV O O O O O O gl~O O O CD O O O O O O O A3 O O O O O Cn O O O tl~o tCI CD O O O A3 O O O Oi: bJ O rn CA Cll O O O Ul O O O I CD CA O O O O O O 35 Energy Release Rate (F6 Btu/sec)O OO O Ul O O A3 O rl LCI 3 O O O A3 O O O M O O 0)3 UD CA Pal O O O Ul O O CA O O O O O O 36 f i(

R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction StudyAttachment 1'om uter des sed for ntainment Inte ri Anal sis The following is a general description of each of the computer codes used in this analysis.LOFTRAN The LOFTRAN program is used for studies of transient response of a PWR system to specified perturbations in process parameters.

LOFTRAN simulates a multiloop system by a model containing a reactor vessel, hot and cold leg piping, steam generator (tube and shell sides)and the pressurizer.

The pressurizer heaters, spray, relief, and safety valves are also considered in the program.Point neutron kinetics model, and reactivity effects of the moderator, fuel, boron, and rods are included.The secondary side of the steam generator utilizes a homogeneous, saturated mixture for a thermal transients and a water level correlation for indication and control.The Reactor Protection System is simulated to include reactor trips on high neutron flux, Overtemperature hT, Overpower hT, high and low pressure, low flow, and high pressurizer level.Control systems are also simulated including rod control, steam dump, feedwater control, and pressurizer pressure control.The Emergency Core Cooling System, including the accumulators and upper;head injection, is also modeled.LOFTRAN is discussed further in Reference A.COCO The COCO computer code (Reference B)is used to analyze the containment pressure transient response following a main steam line break accident.COCO is a mathematical model of a generalized containment; the proper selection of various options in the code allows the creation of a specific model for the particular containment design.The values used in the specific model for the different aspects of the containment are derived from plant-specific input data.The COCO computer code consists of time-dependent conservation equations of mass and energy, together with steam tables, equations of state and other auxiliary relationships.

Transient conditions are determined for both the containment steam-air mixture and the sump water.The energy equation is applied to the containment shell to obtain transient temperature gradients as well as heat stored in and conducted through the structure.

Heat removal by means of energy storage in equipment within the containment, internal sprays, emergency containment coolers, and sump water recirculation cooling system is considered.

The containment air-steam-water mixture is separated into two distinct systems.The first system consists e 1

R.E>>.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study of the air-steam phase, while the second system is the water phase in the containment sump.This division permits more accurate representation of the distinct physical phenomena occurring in each system.The steam-air mixture and water phase are assumed to have uniform properties.

In addition, temperature equilibrium between the air and steam is assumed.However, this does not imply continual thermal equilibrium between the steam-air mixture and water phase.Sufficien'relationships to solve the problem independent of this restriction are provided by the equations of conservation of mass and energy as applied to each system, together with appropriate equations of state and heat transfer boundary conditions.

Air inside the containment is treated as an ideal gas.Thermodynamic properties of water and steam are derived from compressed water and steam tables.Pe Heat transfer through, and heat storage in, interior and exterior walls of the containment structure are considered.

Structural heat sinks, consisting of steel and concrete, are modeled as slabs having specific areas and layers of varying thicknesses.

The thermal conductivity, density and specific heat of each layer are specified at an initial temperature.

Discharge mass and energy flow rates through the rupture are established by separate analyses of the steam generator transient.

This information is supplied as time-dependent data to the code.For the larger steam line break cases, the calculation assumes the Tagami condensation heat transfer correlation and the revaporization model.The revaporization model assumes that an equilibrium condition exists between the condensate on the containment structures and the containment steam-air atmosphere.

At each time step, the conservation equations (mass, energy, and state)are solved simultaneously to deternjne a new containment air-steam-condensate condition.

If the calculated condition is a saturated state, water mass (condensate) forms and is assumed to fall instantly into the sump.If the condition is a super-heated state, the water mass would not form at that time step.The condensate which is at a saturated state based on the interfacial temperature at a previous time step may re-evaporate under the exposure to a rapidly increasing super-heated atmosphere.

The COCO code has been benchmarked against the CVTR tests (Reference C).The CVTR tests were super-heated steam blowdown tests.The containment free volume is about one-eighth of a typical three loop PWR containment.

The blowdown steam enthalpy was 1195 BTU/ibm, which is about the same as R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study that for a postulated-steam line break with no moisture carry-over.

~The COCO calculation showed good agreement with the test data when the revaporization model was-used.When no revaporization was assumed, the COCO calculation predicted a much higher temperature

.than the test.In both cases, COCO over-predicted the containment atmosphere pressure.For small steam line breaks, the condensation heat transfer is based on stagnant conditions and the wall condensate is assumed to fall to the sump with no revaporization.

The approved mass and energy release model assumes no entrainment, i.e., dry steam blowdown.The NRC staff has approved the use of the revaporization model, on previous plant-specific applications, for break sizes which would have entrainment (Reference D).The use of the revaporization model has been approved for large steam line breaks in the LOTIC-3 code used for ice condenser plants (Reference E).

Oi e R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction StudyReferences for Attachment 1 A.Burnett, T.W.T., et.al.,"LOFTRAN Code Description," WCAP-7909-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.B.Bordelon, F.M.and Murphy, E.T.,"Containment Pressure Analysis Code (COCO)," WCAP-8327, June 1974.C.Schmitt, R.C., Bingham, G.E., and Norberg, J.A.,"Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment

-Final Report," IN-1403, Idaho Nuclear Corporation, December, 1970.D."Diablo Canyon Safety Evaluation Report," NUREG-0675, June 1980.E.Hsieh, T.and Liparulo, N.J.,"Westinghouse Long Term Ice Condenser Containment Code-LOTIC-3 Code," WCAP-8354-P-A, Supplement 2, February 1979.

p~lWa,-ll'lt, I ATTACHMENT D Comparison of Existing to Proposed Technical Specifications Proposed verbage in bold print Deleted Verbage Crossed out 0>C IV 1 Dose E uivalent I-131 The dose equivalent I-131 shall be that concentration of I-131 which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.The dose conversion factors used for this calculation shall be those for the adult thyroid dose via inhalation, contained in NRC Regulatory Guide 1.109 Rev.1 October 1977.Re ortable Event A Reportable Event shall be any of those conditions specified 1.20 in Section 50.73 to 10CFR.Part 50.Canisters Containin Consolidated Fuel Rods Canisters containing consolidated fuel rods are stainless steel canisters containing the fuel rods of no more than two fuel assemblies which have decayed at least five years and are capable of being stored in a storage cell of the spent fuel 1 21 pool~Shutdown Mar in Shutdown margin shall be the amount of reactivity by which the reactor is subcritical, or would be subcritical from its present condition assuming all rod cluster control assemblies (shutdown and control)are fully inserted except for the single rod cluster control assembly of highest reactivity worth which is assumed to be fully withdrawn, and assuming no changes in xenon or boron concentration.

Amendment No.12 1-8 Proposed 4 f, 4 Chemical and Volume Control S stem A licabilit Applies to the operational status of the chemical and volume control system.To define those conditions of'the chemical and volume control system necessary to assure safe reactor operation.

S ecification During cold shutdown or refueling with fuel in the reactor there shall be at least one flow path to the core for boric acid injection.

The minimum capability for boric acid injection shall be equivalent to that supplied from the refueling water storage tank.3.2.1.1 Mith this flow path unavailable, immediately suspend all operations involving core alterations or positive reactivity changes and return a flow path to operable status as soon as possible.3.2.2 I 4 g fs l\~R a s%I When the reactor is above cold shutdown, two boron injection flow paths shall be operable with one operable charging pump for each operable flow path, and one operable boric acid transfer pump for each operable flow path from the boric acid storage tank(s).~~4~Zf required by specification 3.2.2 above, the Boric Acid Storage Tank(s)shall satisfy the concentration, minimum volume and solution temperature recpxirements of Table 3.2-1.Amendment No.33 3~2 1 Proposed

~+~4 h'N~I IPI 3.2'With only one of the required boron injection flow paths to the RCS operable, restore at least two boron injection flow paths to the RCS to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be"'in at least hot shutdown and borated to a shutdown margin equivalent to at least 2.45%delta k/k at cold, no xenon conditions.

If the requirements of.3.2.2 are not satisfied within an additional 7 days, then be in cold shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.Whenever the RCS temperature is greater than 200'F and is being cooled by the RHR system and the over-pressure protec-tion system is not operable, at least one charging pump shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that.the control switch is in the pull-stop posi-tion.Amendment No.3'2 Proposed

~~, C4'A4~0 pj+,

Table 3.2-1 Boric Acid Storage Tank(s)Minimum-Volume-Temperature-Concentration+

Concentration ppm boron Minimum Volume gal.Minimum Solution Temperature F 4700 to 5000 to 6000 to 7000 to 8000 to 9000 to 10000 to~~-11000',to 12000 to 13000 to 14000 to 15000 to 0 16000 to 17000 to 18000 to 19000 to 20000 to 21000 to 22000 to less than 5000 less than 6OOO less than 7000 less than 8000 less than 9000 less than 10000 less than 11000 less;than.12000 less than 13000 less than 14000 less than 15000 less than 16000 less than 17000 less than 18000 less than 19000 less than 20000 less than 21000 less than 22000 less than 23000 8400 7800 6400 5400 4700 4200 3800 3500 3200 3000 2700 2500 2400 2200 2100 2000 1900 1800 1800 40 52 62 70 78 85 91 97 103 108 113 118 123 127 131 137 140 143 145 Amendment No.3~2 2a Proposed llDJf (~gL~~l 4

~~Basis The chemical and volume control system provides control of the reactor coolant system boron inventory.+

4e~~Q)

~'r%46'~g~lf~t'h This is normally accomplished by using one or more charging pumps in~~=-series with one of the two boric acid transfer pumps.Above cold shutdown conditions, a minimum of two of four boron injection flowpaths are required to insure single functional capability in the event..that an assumed single active, failure.renders one of the flow paths inoperable.

The boration volume available through any flow path is sufficient to provide the required shutdown margin at cold conditions from any expected operating condition and to compensate for shrinkage of the primary coolant from the cooldown process.The maximum volume*.;--~~'.recpirement-is.associated with boration from just critical, hot zero power, peak xenon with control rods at the insertion limit, to cold shutdown with single reactor coolant loop operation.

This requires 26q000@gallons of 2000 ppm borated water from the refueling water storage tank or the concentrations and volumes of borated water specified in Table 3.2-1 from the boric acid storage tanks.Two boric acid storage tanks are available.

One of, the two tanks may be, out of service provided the required volume of boric acid is available to the operable flow paths.Above cold shutdown, two of the following four flow paths must be operable with one operable charging pump for each operable flow path, and one operable boric acid transfer pump for each operable flow pathfrom the boric acid storage tanks.Boric acid storage tanks via one boric acid transfer pump through the normal makeup (FCV 110A)flow path to the suction of the charging pumps.Boric acid storage tanks via one boric acid transfer pump through the emergency boration flow path (MOV 350)to the j f I

(~)suction of the charging pumps.Refueling water storage tank via gravity feed through AOV 112B to the suction of the charging pumps.--Amendment No.24 3~2 3 Proposed

~l>~~fi~a~i~~1 5 4~ih 1$'5 C<<+'~l 0 (4)Refueling water storage tank via gravity feed through manual bypass valve 358 to the suction of the charging pumps.'Available flow paths from the charging pumps to the reactor coolant system include the following:

(1)Charging flow path through AOV 392A to the RCS Loop B hot leg.(2)Charging flow path through AOV 294 to the RCS Loop B cold leg.(3)Seal injection flow path to the reactor coolant pumps.The.rate of boric acid~injection-must.be sufficient.to offset the maximum addition of positive reactivity from the decay of xenon after a trip from full power.This can be accomplished through the operation of one charging pump at minimum speed with suction from the refueling water storage tank.Also the time required for boric acid injection allows for the local alignment of manual valves to provide the necessary flow paths.The quantity of boric acid specified in Table 3.2-1 for each concentra-tion is sufficient at any time in core life to borate the reactor coolant to the required cold shutdown concentration and provide makeup to maintain RCS inventory during the cooldown.The temperature limits specified on Table 3.2-1 are required to maintain solution solubility at the upper concentration in each range.The temperatures listed on Table 3.2-1 are taken from Reference (4).An arbitrary 5'F is added to the Reference (4)for margin.Heat tracing may be used to maintain solution temperature at or above the Table 3.2-1 limits.If the solution temperature of either the flow path or the borated water source is not maintained at or above the minimum temperature specified, the affected flow path must be declared inoperable and the appropriate actions specified in 3.2.4 followed.Placing a charging pump in pull-stop whenever the reactor coolant system temperature is>200 F and is being cooled by RHR without the over-pressure protection system operable will prevent inadvertent overpres-surization of the RHR system should letdown be terminated.">

References:

UFSAR Section 9.3.4.2 (2)(3)RG&E Design Analysis DA-NS-92-133-00"BAST Boron Concentration Reduction Technical Specification Values" dated Dec.14, 1992 L.D.White, Jr.letter A.Schwencer, NRC,

Subject:

Reactor Vessel Overpressurization, dated February 24, 1977 Amendment No.Proposed

<4)Kerr-McGee Chemical Corp.Bulletin 0151"Boric Acid-Techni-cal Grades" dated 5/84 Amendment No.3.2-5 Proposed 3.3 Emer enc Core Coolin S stem Auxiliar Coolin S stems Air Recirculation Fan Coolers Containment S ra and Charcoal HEPA Filters To define those conditions for operation that are neces-sary:(1)to remove decay heat from the core in emergency or normal shutdown situations, (2)to remove heat from contain-ment in normal operating and emergency situations, (3)to remove airborne iodine from the containment atmosphere following a postulated Design Basis Accident, and (4)to minimize containment leakage to the environment subsequent to a Design'Basis Accident.S ecification

3.3.1 Safet

In ection and Residual Heat Removal S stems a I 0 3.3.1.1 The reactor shall not be taken above the mode indicated unless the following conditions are met: a~b.Above cold shutdown, the refueling water storage tank contains not less than 300,000 gallons of water, with a boron concentration of at least 2000 ppm.Above a reactor coolant system pressure of 1600 psig, except during performance of RCS hydro test, each accumulator is pressurized to at least 700 psig with an indicated level of at least 50%and a maximum of 82~with a boron concentration of at least 1800 ppm.c~At or above a reactor coolant system temperature of 350oF three safety injection pumps are operable.Amendment No.24 3~3 1 Proposed 1 l N Ac At or above an RCS temperature of 350'F, two residual heat removal pumps are operable.At or above an RCS temperature of 350'F, two residual heat removal heat exchangers are operable.At the conditions required in a through e above, all valves, interlocks and piping associated with the above components which are required to function during accident conditions are operable.At or above an RCS temperature of 350'F, A.C.power shall be removed from the following valves with the valves in the open position: safety injection cold leg injection valves 878B and D.A.C.power shall be removed from safety injection hot leg injection valves 878A and C with the valves closed.D.C.control power shall be removed from refueling water storage tank delivery valves 896A, 896B and 856 with the valves open.At or above an RCS temperature of 350'F, check valves 853A, 853B, 867A, 867B, 878G, and 878J shall be operable with less than 5.0 gpm leakage each.The leakage requirements of Technical Specification 3.1.5.2.1 are still applicable.

Above a reactor coolant system pressure of 1600 psig, except during performance of RCS hydro test, A.C.power shall be removed from accumulator isolation valves 841 and 865 with the valves open.

At or above an RCS temperature of 350 F, A.C.power shall be removed from Safety Injection suction valves 825A and B with the valves in the open position, and from valves 826A, B, C, D with the valves in the closed position.Amendment No.42 3~3 2 Proposed

At or above an RCS temperature of 350o F, A.C.power shall be removed from Safety Injection suction valves 825A and B with the valves in the open position, and from valves 826A, B, C, D with the valves in the closed position.Amendment No.42 3~3 2 Proposed

.~*g 3.3.1.2 If the conditions of 3.3.1.1a are not met, then satisfy the~~~~~~~~~~condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be at hot shutdown in the next 6 3.3.1.3 3.3.1.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least cold shutdown within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.The requirements of 3.3.1.1b and 3.3.1.1i may be modified to allow one accumulator to be inoperable or isolated for up to one hour.If the accumulator is not operable or is still isolated after one hour, the reactor shall be placed in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and below a RCS pressure of 1600 psig within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The requirements of 3.3.1.1c may be modified to allow one safety injection pump to be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.If the pump is not operable after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and a4-aa 3.3.1.5 below a RCS temperature less than 350'F within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The requirements of 3.3.1.1d through h.may be modified to allow components to be inoperable at any one time.More-than one component may be inoperable at any one time provided that-one..train of the ECCS is operable.If the requirements of 3.3.1.1d through h.are not satisfied within the time period specified below, the reactor shall be placed in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at an RCS temperature less than 350 F in an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.a.One residual heat removal pump may be out.of service provided the pump is restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Amendment No.24 3'3 Proposed

t l 4 l II,.l~

b., One residual heat removal heat exchanger may be out of service for a period of no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.c.Any valve, interlock, or piping required for the func-tioning of one safety injection train and/or one low head safety injection train (RHR)may be inoperable provided repairs are completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (except as speci-fied in e.below).d.Power may be restored to any valve referenced in 3.3.1.1g for the purposes of valve testing provided no more than-~--one such valve has power restored and provided testing is completed and power removed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.e.Those check valves specified in 3.3.1.1h may be inopera-ble (greater than 5.0 gpm leakage)provided the inline MOVs are de-energized closed and repairs are completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.3.3.1.6 Deleted Amendment No.24, 33 3.3-4 Proposed WP~w'A'I(*

that the mass addition from the inadvertent operation of safety injection will not result in RHR system pressure exceeding design limits.The limitation on no safety injection pumps operable and the discharge lines isolated when overpressure protection is provided by the pressur-izer PORV's removes mass injection from inadvertent safety injection as an event for which this, configuration of overpressure protection must be designed to protect.Inoperability of a safety injection pump may be verified from the main control board with the pump control switch in pull stop, or the pump breaker in the test racked out position such that the pump could not start from an inadvertent safety injection signal.Isolation of a safety injection pump discharge path to the RCS may be verified from the main control board by the discharge MOV switch position indicating closed, or the discharge valve closed with A.C.power removed, or a manual discharge path isolation valve closed such that operation of the associated safety injection pump would not result in mass.injection to the RCS.Amendment No.48 3.3-14 Proposed ah dgr'g'>4 P',c High concentration boric acid is not needed to mitigate the consequences of a design basis accident.Reference (10).demonstrates.

that.the design basis accidents can be mitigated by safety injection flow of RWST concentration.

Therefore, SI pump suction is taken from the RWST.Requiring that the safety injection suction valves (825A and B, 826A, B, C-and.D)are.aligned.with A.C..power removed ensures that the safety injection system would not be exposed to high concentration boric acid and the assumptions of the accident analysis are satisfied.

Amendment No.'48 3.3-14 Proposed U

References (1)Deleted (2)UFSAR Section 6.3.3.1 (3)UFSAR Section 6.2.2.1 (4)UFSAR Section 15.6.4.3 (5)UFSAR Section 9.2.2.4 (6)UFSAR Section 9.2.2.4 (7)Deleted (8)UFSAR Section 9.2.1.2 (9)UFSAR Section 6.2.1.1 (Containment Integrity) and UFSAR Section 6.4 (CR Emergency Air Treatment)

(10)Westinghouse Report,"R.E.Ginna Boric Acid Storage Tank Boron Concentration Reduction Study" dated Nov.1992 by C.J.McHugh and J.J.Spryshak Amendment No.48 3.3.14a Proposed Ii A0 Cha Desc tion 10.Rod Position Bank Counters 11.Steam Generator Level 12.Charging Flow TABLE 4.Check S(1,2)N.A.Continued)

C ibrate Test RemarksV N.A.N.A.N.A.1)With rod position indication 2)Log rod position indications each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when rod deviation monitor is out of service 13.Residual Heat Removal Pump Flow N.A.N.A.14.Boric Acid Storage Tank Level D N.A.Note 4 15.Refueling Water Storage Tank Level N.A.N.A.16.Volume Control Tank Level N.A.N.A.17.Reactor Containment Pressure D M(1)1)Isolation Valve signal 18.Radiation Monitoring System D Area Monitors R1 to R9, System Monitor R17 19.Boric Acid Control N.A.N.A.20.Containment Drain Sump Level N.A.N.A.21.Valve Temperature Interlocks N.A.N.A.22.Pump-Valve Interlock N.A." N.A.23.Turbine Trip Set-Point N.A.R'(1)1)Block Trip 24.Accumulator Level and Pressure N.A.Amendment No.22 4.1-6 Proposed V E Cha Descr tion 39.Reactor Trip Breakers TABLE 4.Check N.A Continued)

Ca ibrate Test N.A.M Remarks Function test-Includes independent testing of both undervoltage and shunt trip attachment-of reactor trip breakers.Each of the two reactor trip breakers will be tested on alternate months.40.Manual Trip Reactor N.A.N.A.R Includes independent testing of both undervoltage and shunt trip circ-uits.The test shall also verify the operability of the bypass break-er.41a.Reactor Trip Bypass Breaker N.A.N.A.M Using test switches in the reactor protection rack manually trip the reactor trip bypass breaker using the shunt trip coil.41.b Reactor Trip Bypass Breaker N.A.N.A.R Automatically trip the undervoltage trip attachment.

NOTE 1 NOTE 2: Logic trains will be tested on alternate months corresponding to the reactor trip breaker testing.Monthly logic testing will verify the operability of all sets of reactor trip logic actuating contacts on that train (See Note 3).Refueling shutdown testing will verify the operability of all sets of reactor trip actuating contacts on both trains.In testing, operation of one set of contacts will result in a reactor trip breaker trip;the operation of all other sets of contacts will be verified by the use-of indication circuitry.

Testing shall be performed monthly, unless the reactor trip breakers are open or shall be performed prior to startup if testing has not bee performed within the last 30 days.NOTE 3 The source range trip logic may be excluded from monthly testing provided it is tested within 30 days prior to startup.NOTE 4: When BAST is required to be operable.Amendment No.4.1-7a Proposed

~Y" TABLE 4.1-2 MINIMUM FRE UENCIES FOR E UIPMENT AND SAMPLING TESTS Test FrecruFenc r 1.Reactor Coolant Chemistry Samples 2.Reactor Coolant Boron Chloride and Fluoride Oxygen Boron Concentration 3 times/week and at least every third day 5 times/week and at least every second day except when below 2504F Weekly 3.Refueling Water Storage Tank Water Sample Boron Concentration Weekly 4.Boric Acid Storage Boron Concentration Tank Twice/Week~+

5.Control Rods 0 6a.Full Length Control Rod 6b.Full Length Control Rod 7.Pressurizer Safety Valves 8.Main Steam Safety Valves 9.Containment Isolation Trip 10.Refueling System Interlocks Rod drop times of all full length rods Move any rod not fully inserted a sufficient number of steps in any one direction to cause a change of position as indicated by the rod position indication system Move each rod through its full length to verify that the rod position indication system transitions occur Set point Set point Functioning Functioning After vessel head removal and at least once per 18 months (1)Monthly , Each Refueling Shutdown Each Refueling Shutdown Each Refueling Shutdown Each Refueling Shutdown Prior to Refueling Operations Amendment No.22 4.1-8 Proposed

12.13.Service Water System Fire Protection Pump and Power Supply Spray Additive Tank Test Functioning Functioning NaOH Concent Freceeuenc Each Refueling Shutdown Monthly Monthly 14.Accumulator Boron Concentration Bi-Monthly 15.Primary System Leakage Evaluate Daily 16.Diesel Fuel Supply Fuel Inventory Daily 17.Spent Fuel Pit 18.Secondary Coolant Samples Boron Concentration Gross Activity Monthly 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2)(3)19.Circulating Hater Flood Protection EquipmentNotes: Calibrate Each Refueling Shutdown (2)Also required for specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods.Not required during a cold or refueling shutdown.An isotopic analysis for I-131 equivalent activity is required at least monthly whenever the gross activity determination indicates iodine concentration greater than 10'f the allowable limit but only once per 6 months whenever the gross activity determination indicates iodine concentration below 10%of the allowable limit.(4)When BAST is required to be operable.Amendment No.22 4.1-9 Proposed 0 ,1