ML20147F562

Safety Evaluation in Support of Containment Temp Increase for North Anna Power Station

ML20147F562
Person / Time
Site:	North Anna
Issue date:	03/02/1988
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20147F540	List: ML20147F534 ML20147F550 ML20147F562
References
NUDOCS 8803070379
Download: ML20147F562 (79)

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[ SAFETY EVALUATION IN KDD0RT OF CONTAINMENT TEMPERATURE INCREASE s

FOR THE NORTH ANNA POWER STATION I

8803070379 880302 PDR ADOCK 05000338 P PDR

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TABLE OF CONTENTS

.Page-

. TAB LE O F C0N T ENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 LIST'0F FIGURES.............................................. 3 L I S T O F TAB L E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

1.0 INTRODUCTION

............................................. 5

' 2.0 PROPOSED TECHNICAL SPECIFICATION CHANGES................. 6 3.0 TRANSIENTS REQUIRING EVALUATION OR ANALYSIS.............. 9 4.0 LOCA ANALYSIS.... ~..................................... 10 4.1 Introduction..................................... 10 4.2 Peak Pressure Analysis........................... 12 4.3 Depressurization and Third Peak Analysis......... 15 4.4 Recirculation Spray Pump _NPSH Analysis........... 16 4.5 Low Head Safety Injection Pump NPSH Analysis.....

19 5.0 MSLB DISCUSSION AND EVALUATION.......................... 53 5.1 Introduction..................................... 53 5.2 Spectrum of Breaks Considered.................... 54 5.3 Mass and Energy Release Analysis................. 55 5.4 Containment Pressure and Temperature Analysis.... 58 5.5 Equipment Qualification.......................... 61

6.0 CONCLUSION

S................. ........................... 69 7.0 10 CFR 50.59 EVALUATION....-............................ 71 REFERFNCES.................................................. 74 l

L 2

l -

LIST OF FIGURES Page 4.2-1 Containment Pressure vs Time - Spectrum of Break Locations...................................... 31 4.2-2 Containment Pressure vs Time - Spectrum of Pump Suction Breaks.................................. 32 4.2-3 Steam Condensing Coefficient vs Time Pump Suction - Minimum ESF........................ 4 . 33 4.2-4 Containment Temperature vs Time Pump Suction DER - Minimum ESF....................... 34 4.2-5 Containment Pressure vs Time PSDER - Single Failure Analysis. . . . . . . . . . . . . . . . . . . . . . 35 4.3-1 Pressure Transients, Worst Case Depressurization, PSDER, Minimum ESF................................... 36 4.3-2 Condensing Coefficient, Worst Cise Depressurization PSDER, Minimun ESF................................... 37 4.3-3 Quench Spray Flow Rate, Worst Case Depressurization PSDER, Minimum ESF................................... 38 4.3-4 Temperature Transients, Worst Case Depressurization PSDER, Minimum ESF................................... 39 4.3-5 Recirculation Spray Cooler Duty, Worst Case RS Pump NPSH PSDER, Minimum ESF................................... 40 4.4-1 NPSHA Transients, Worst Case RS Pump NPSH - HLDER Normal ESF........................................... 41 4.4-2 Pressure Transients, Worst Case RS Pump NPSH - HLDER N o rm a l E S F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 4.4-3 Temperature Transients, Worst Case RS Pump NPSH -

H LD E R N o rma l E S F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 4.4 Recirculation Spray Cooler Duty, Worst Case RS Pump NPSH HLDER Normal ESF..................................... 44 4.4-5 Quench Spray Flow Rate, Worst Case RS Pump NPSH - HLDER N o rm a l E S F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 4.4-6 Condensing Coefficient, Worst Case RS Pump NPSH - HLDER N o rm a l E S F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 4.5-1 NPSHA Transients, Worst Case LHSI Pump NPSH - PSDER Minimum ESF.......................................... 47 4.5-2 Pressure Transients, Worst Case LHSI Pump NPSH - PSDER Minimum ESF.......................................... 48 4.5-3 Temperature Transients, Werst Case LHSI Pump NPSH - e PSDER, Minimum ESF................................... 49 4.5-4 Recirculation Spray Cooler Duty, Worst Case LHSI Pump NPSH - PSOER, Minimum ESF............................ 50

• 1; 3

LIST OF FIGURES (Continued):

Page 4.5-5 Quench Spray Flow Rate, Worst Case LHSI Pump NPSH -

PSDER, Minimum ESF................................... 51 4.5-6 Condensing Coefficient, Worst Case LHSI Pump NPSH -

PSDER, Minimum ESF................................... 52 5.4-1 Peak Containment Pressure, Main Steam Line Break...... 67 5.4-2 Peak Containment Temperature, Main Steam Line Break... 68 LIST OF TABLES Page 4.2-1 Input Data to LOCTIC Program North Anna Power Station. 22 4.2-2 Initial Conditions Assumed in the LOCA Analysis....... 23 4.2-3 Summary of Results of Containment Analysis - Spectrum of Break Sizes and Locations......................... 24 4.2-4 Comparison of Key Pressure Response Results........... 25 4.2-5 Chronology for PSOER L0CA............................. 26 t 4.2-6 Mass and Energy Releases To Containment............... 27 4.2-7 Energy Distribution................................. 4. 29 4.4-1 Recirculation Spray Pump NPSH Analysis Summary Results. 30 5.2-1 Main Steam Line Brea k Ca se Summary. . . . . . . . . . . . . . . . . . . . 62 5.3-1 Reactor Coolant System Parameters. . . . . . . . . . . . . . . . . . . . . 63 5.3-2 Seconda ry System Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . 64 5.4-1 MSLB Analysis Initial Conditions . . . . . . . . . . . . . . . . . . . . . 65

- 5.4-2 Summary of Main Steam Line Break Results. . . . . . . . . . . . . . 66 4

l:

1.0 INTRODUCTION

The North Anna Power Station (NAPS) Units 1 and 2 currently operate with an allowable containment air-temperature range of 86'F s T s105'F.

The upper temperature limit is approached during the summer months _ due to high ambient and service. water temperatures. This has required implementation of specialized manpower-intensive procedures in order to prevent a violation of the Technical Specification limit. Therefore, Virginia' Electric and Power Company is submitting this safety evaluation to justify increasing the upper containment temperature limit to a value

~

of 120'F.

The containment temperature limit is set by performiag the necessary transient analyses to ensure that the containment design criteria are met following a design basis accident. Temperature is a significant initial condition for these analyses. Another important analysis input is the volume of water from the Refueling Water Storage Tank (RWST) available for the Quench Spray system. This volume has been defined in such a way ,

as to permit the minimum Technical Specification volume to be reduced.

This approach was chosen to permit the use of wide range level instrumentation for Technical Specification surveillance. This change to, the Technical Specifications is being implemented specifically to address concerns identified by the NRC resident inspector . 8 Section two discusses the individual Technical Specification changes.

The transients which need to be evaluated or analyzed in order to permit operation at the increased containment temperature are presented in 5

r section three. The individual transients are discussed in- subsequent-sections. Finally, a 10 CFR 50.59 evaluation is presented.  ;

'2.0 PROPOSED TECHNICAL SPECIFICATION CHANGES Several North Anna Technical Specifications need to be changed to incorporate the results of the analyses which' justify operation at the increased containment temperature and modified RWST level. Each change is discussed separately in the following paragraphs. A more detailed discussion of these changes can be found in Sections 4 and 5 of this safety evaluation. The Specifications that follow and the changes discussed are identical for Unit I and Unit 2.

3/4 3.5.5 Refueling Water Storage Tank LCO 3.5.5 establishes the minimum and maximum available water volume in the RWST. Additionally, this specification provides limits for the soluble boron concentraticn and solution temperature. The minimum volume is revised to be consistent with the value used in the containment transient analyses described herein and was chosen to permit the use of wide range level instrumentation for Technical Specification surveillance.

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3/4 6.1.2 Containment Leakage Rates The limiting condition for operation (LCO) and the surveillance.

requirement (SR) have been modified to account for the new maximum pressure following a LOCA. Tne peak pressure resultin;; from the LOCA analyses is 44.1 psig for a pump suction double end rupture. This result is incorporated in the LCO and SR.

3/4 6.1.3 Containment Air Lock The LCO and SR have been modified to account for the new maximum

', pressure following a LOCA. The peak pressure resulting from the LOCA analysis is 44.1 psig for a pump suction double end rupture. This result is incorporated in the LCO and SR.

3/4 6.1.4 Internal Air Partial Pressure The containment internal air partial pressure limit has been revised .

based on the analytical results. The minimum limit is set by the structural criteria for the containment mat. The revised diagonal limit line in Technical Specification figure 3.6-1 is set by the LOCA 1 depressurization analysis. The MSLB analysis sets the horizontal limit line at 11.1 psia. All of the limit lines have been conservatively changed to reflect the containment vacuum and leakage monitoring pressure instrumentation loop uncertainty which is used by the operator for periodic surveillance.

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c-3/4 6.1.5 Containment Air Temperature One goal of the analysis effort was to increase the maximum containment air temperature from 105'F to 120'F. Since all~of the analyses showed acceptable results, the temperature increase is proposed. The containment air temperature instrumentation uncertainty was also accommodated.

83/4.5.5 Refueling Water Storage Tank The basis for the RWST Technical Specification has been modified to state that instrument uncertainty was included in the safety analysis.

The operator does not have to account for level uncertainty during surveillance.

B3/4.6.1.4 and 83/4.6.1.5 Internal Pressure and Temperature The basis for the containment internal pressure and temperature Technical Specification has been modified to state that instrument uncertainty was considered in the safety analysis. The operator does not have to account for these uncertainties during surveillance.

83/4. 6.1.6 Containment Structural Integrity The basis for the structural integrity specification was modified -

because it contained reference to the numerical value of the peak LOCA temperature. Rather than list the revised value of 44.1 psig the -

8

numerical value has been removed. 1he magnitude of the value is not significant in the context of this specification since the discussion relates to'the way containment integrity is maintained.

3.0 TRANSIENTS REQUIRING EVALUATION OR ANALYSIS The containment bulk air temperature and the service water temperature limits represent a severe constraint to operation of the North Anna Power Station during the sunimer months. This situation exists because there is no practical way to reduce the temperature in a large enclosed volume or a large body of water in short periods of time. However, the ACTION f

statements in LCO 6.1.5 and LCO 7.5.1 provide eight and six hours respectively to restore the respective temperature to within it's limit.

Given the rigid ACTION statement for these two temperatures, it was decided to increase the Technical Specification limit in order to increase operational flexibility.

The plant design basis was reviewed to determine which transients are impacted by the increased service water temperature and containment bulk temperature. The containment design is based on two Condition IV transients; the Loss of Coolant Accident (LOCA) and the Main Steam Line Break (MSLB). As a result of the wrst LOCA or worst MSLB, containment integrity is assured if the following three conditions are satisfied:

= The peak calculated containment pressure is less than the design pressure of 45 psig. -

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= The containment is depressurized to subatmospheric within one hour of the accident.

= Once depressurized, the containment is maintained at a pressure less than atmospheric for the duration of the accident.

The Condition II and III transients either do not breach the reactor coolant pressure boundary or are less severe than the LOCA and MSLB transients analyzed as part of tho containment design. For the analysis reported in this licensing package several LOCA scenarios were analyzed.

The MSLB analysis includes a spectrum of break sizes and power levels t which was necessary because of a change in the methodology. The MSLB ,

analysis and the LOCA analysis are discussed in detail in the following sections.

4.0 LOCA ANALYSIS 4.1 Introduction The containment response to a LOCA was reported in the original FSAR.

Since that time the LOCA has been analyzed as a result of changes in tha operating conditions of the reactor coolant system. For example, the core average temperature was increased a total of 7.5'F and the core power was uprated to 2893 *t. In each instance the limiting LOCA break was analyzed to be certain that the containment design criteria were met with the revised operating conditions. The latest LOCA analysis was done to 10

support changes in the containment safety analysis initial conditions such as air temperature and service water temperature.

The maximum containment temperature was increased from 105'F to 120'F. A lower volume of RWST water was an assumed initial condition corresponding to a 7.5% reduction from the-Technical Specification limit of 487,000 gallons. Finally, the maximum service water temperature was assumed to be 97'F. Each of these initial conditions are reflected in the LOCA analysis presented in this licensing package. In addition, the 2% initial power uncertainty was included in the LOCA analysis.

Proposed Technical Specification changes for air temperature and RWST volume are presented in Section Two. The other changes to the Technical Specifications in Section Two are required by the analysis results. A change in the service water temperature Technical Specification limit is '

not requested at this time although the analysis supports a 97 'F limit.

The LOCA analysis consists of a peak pressure, a depressurization and d third peak analysis along with an NPSHA analysis for the containment spray pumps and the LHSI pumps. A spectrum of analyses were performed to determine the limiting break size and location. The depressurization analysis was performed for the limiting break which is a double ended -

rupture at the reactor coolant pump suction (PSDER). The limiting break size and location for the containment spray pump NPSHA analysis was also determined from a spectrum analysis. Therefore, only the double ended rupture of the hot leg was considered for the recirculation spray pumps.

Similarly, the PSDER with minimum engineered safety features (ESF) gives ,

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y 4

the limiting NPSHA for the low head safety injection pumps. Each of these scenarios were analyzed with the LOCTIC computer code and the results are reported herein.

The LOCTIC computer code 8 was used to perform the LOCA analysis. This program calculates the temperature and pressure of the containment atmosphere as a function of time following a LOCA. The accident starts with the break which is assumed to be the zero reference time. The program considers the various heat sources and sinks as a function of time in a given containment configuration to calculate the temperature and pressure transients of the containment atmosphere. The calculational process is a digital integration of the changes taking place. The program assumes that no significant changes take place between each time step and that the calculations during each time step are on a steady-state basis.

At the end of each time step, the heat inflows and outflows are summed and new containment and primary and secondary system conditions established.

4.2 Peak Pressure Analysis A peak pressure analysis was performed to insure that the maximum contaI$Inent design pressure of 45 psig was not exceeded as a result of the revised initial conditions. The maximum temperature is also compared against the design containment temperature in this analysis. A summary of the LOCTIC input used in the LOCA analysis is presented in Table 4.2-1.

A summary of the initial conditions is presented in Table 4.2-2.

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The analysis incorporates two significant input differences relative to previous analyses. The core power level includes the two percent uncertainty on the initial core power level. The decay heat curve is

~ based on the ANSI /ANS 5.1 - 1979 decay heat standard 8 The two percent power uncertainty provides added conservatism in the results to account for possible calorimetric errors. The 1979 versten of the ANSI /ANS decay heat standard is less conservative than earlier versions and has been shown to be a more realistic model of decay heat production following reactor trip. .

The analysis retains the conservative assumptions of loss of offsite power and a single failure. The loss of offsite power is assumed to occur simultaneously with the start of the accident. The single failures analyzed include a diesel generator failure (i.e. minimum ESF) and a quench spray pump failure.

The analyses include a sensitivity study of break location, break size and single failure assumptions. Figure 4.2-1 shows three pressure plots for a spectrum of break locations including: a hot leg double ended rupture (HLDER), a PSDER and a pump discharge uouble ended rupture l (PDOER). Table 4.2-3 lists peak temperature and pressure results for a spectrum of break sizes and locations. Table 4.2-4 presents a sumary of the key pressure results for the PSDER. As shown in these tables, the PSDER yields the peak pressure, the peak temperature and the peak subatmospheric pressure; therefore, it is the limiting transient. The PSDER is limiting because a'1 the fluid leaving the top of the core passes l

l through the steam generators and may become superheated before exiting '

l l

, 13 l

l

the break. The PDDER and the HLDER represent paths to the containment which can be reached without first passing through the steam generators.

The PSDER also has a relatively high flow rate through the core and out the break because there are less restrictions between the core and the break than in the case of a PDDER where the pump itself acts to impede flow from the core.

A spectrum of break sizes has been analyzed for this transient as shown in Figure 4.2-2. The three different pump suction ruptures analyzed show that the double ended rupture is the most limiting.

The chronology of the limiting pressure transient is presented in Table 4.2-5. This table illustrates the significant aspects of the pressure response to a LOCA using the LOCTIC code. There are three pressure peaks.

The first peak occurs as a result of the blowdown from the accident. The second peak occurs due to the heating and subsequent release of the safety i injection water after reflood. Finally, the third peak r'esults from the transition between the infection mode and the recirculation mode of emergency core cooling after the RWST empties.

The mass and energy data from the PSDER are presented in Table 4.2-6.

The energy distribution in the containment for this accident is presented in Table 4.2-7. The steam condensing coefficient is presented in Figure 4.2-3. The containment atmosphere and sump temperature transients are shown in Figure 4.2-4. -

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q. . .- . .

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A single failure analysis was performed for the PSDER LOCA. The i'

diesel generator failure results in minimum engineered safeguards which consist of the following equipment:

= All three safety injection accumulators. ,

= One out of three charg'ng pumps

= One out of two low head safety injection pumps

= One out of two trains of the quench spray (QS) subsystem

= One out of two trains of the recirculation spray (RS) subsystem I

One inside recirculation spray pump (IRS)

One outside recirculation spray pump (ORS)

One casing cooling pump This failure is shown to produce the highest containment pressure of 44.1 psig. Containment pressures for the failures analyzed are shown in Figure 4.2-5.

4.3 Depressurization and Third Peak Analysis The one hour time limit to return the containment below atmospheric pressure is analyzed in the same manner as the peak pressure analysis. '

The limiting transient is the PSOER and the limiting single failure is the loss of one emergency diesel generator which results in the failure ,

of one train of ESF to actuate (i.e. mini.num ESF). However, the initial -

conditions must be specified differently for the depressurization analysis to conservatively determine the peak subatmospheric pressure (i.e. third peak). Sensitivity studies have shown that the depressurization time and peak subatmospheric pressure are insensitive -

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to initial dry bulb and dewpoint temperature for an air partial pressure of 9.25 psia. The initial conditions which lead to conservative depressurization results are listed in Table 4.2-2.

The results of the worst case depressurization analysis are shown in Figures 4.3-1 through 4.3-5 for the following parameters:

= Containment total pressure

= Steam condensing coefficient

=

QS flow rate

= Temperature transient

= Recirculation spray cooler duty 4.4 Recirculation Spray Pump NPSH Analysis The LOCTIC computer program was used to calculate the net positive suction head available (NpSHA) for the inside and outside recirculation spray pumps. Since the most Ilmiting assumptions for an NPSH analysis are different from those for a containment depressurization analysis both analyses must be performed. The NPSH analysis is performed to make certain that the NPSHA exceeds that required (NPSHR) for the flow rate assumed throughout the analysis. The depressurization analysis is intended to show that the containment can conservatively be returned to subatmospheric within one hour and remain below subatmospheric ,

thereafter.

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The assumptions made for the depressurization analysis maximize the energy release to the containment atmosphere (minimize energy release to the sump) in order to overestimate the containment pressure. The assumptions made for NPSHA analyses of the recirculation spray pumps minimize the energy release to the containment atmosphere and maximize the energy release to the containment floor. Thus, the containment pressure is underestimated and the containment floor water vapor is overestimated. Since containment pressure is a positive term in the NPSHA equation and the floor water vapor pressure is a negative term, a

~

conservative calculat1on of NPSHA results.

NPSHA sensitivity studies were performed for the original FSAR analysis'. These studies were performed to determine the limiting case for NPSHA and included various break locations, service water temperatures, RWST temperatures, break sizes and initial conditions.

These studies were not redone for the present analysis since the initial condition changes are relatively small variations and do not affect the sensitivities only the magnitude of the results. The sensitivity studies result in the following conclusions:

= HLDERS are more limiting than PSOERs for the RS pumps

= The IRS pumps are insensitive to the single failure assumption

= Normal ESF and failure of one ORS pump are the most Ilmiting single failures for both the IRS and ORS pumps

= For the RS pumps, cold service water is more limiting because containment pressure is lower which is a positive term in the NPSH equation -

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l l

=

Warm RWST temperature is more limiting for the IRS pumps since it results in less effective cooling of the water at the pump suction

=

Break size has no effect on the NPSHA for the RS pumps

= NPSHA is insensitive to dewpoint

=

A higher containment temperature is more Itmiting All NPSH values are referenced to the centerline of the first stage of the pump impeller. The NPSHA analyses for the ORS show that the NPSHA for a flow of 3640 gpm is always above the NPSHR. Therefore, NPSHA is adequate for the entire transient. The NPSHA analysis for the IRS pumps shows that NPSHA for a flow of 3300 gpm is always above the NPSHR. The results for current licensing analysis are presented in Table 4.4-1.

A HLDER for the lower extreme of service water temperature (35'F) is the limiting case for both the ORS and the IRS spray pumps. Figures 4.4-1 through 4.4-6 show time histories for the following.

= NPSHA i Sump water level

= Containment tot al pressure

= IRS and ORS suction vapor pressures

= Containment temperature

= Sump water temperature

= RS cooler duty

=

QS flow

= Condensing coefficient 18

4.5 Low Head Safety Injection Pump NPSH Analysis An analysis of the NPSHA for the low head safety injection pumps has been perforreed to ensure that the pumps can deliver up to 4030 gpm following the worst case LOCA. The injection mode and the recirculation mode have been evaluated previously and it was found that the recirculation mode is limiting. The calculation of NPSHA in the recirculation mode considers the static head and suction line pressure drop, the vapor pressure of the liquid in the sump and the containment pressure. This calculation ensures that the NPSHA meets the pump requirements.

The calculation of NPSHA is as follows:

NPSHA = h con press - h y ,p+ h stat -h !oss where h con press = containment pressure, ft hy ,p = vapor pressure of the fluid in the sump, ft h stat = static head of the fluid in the sump, f t hioss = suction line losses, ft The original NPSHA sensitivity studies have not been analyzed at the new containment conditions because the new conditions are only slightly different than the existing conditions and therefore do not invalidate the sensitivities. These sensitivities show that the PSDER with minimum .

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ESF is the worst case'. Maximizing the sump water temperature minimizes the NPSH available. This is done by mir,imizing containment heat removed and maximizing the sump and RWST water temperatures. This yields a minimum NPSHA of 13.5 ft; whereas'the NPSHR is 13.4 ft.

The NPSHR and NPSHA are within about 1% of each other at the time of switchover from the injection to the recirculation mode of emergency core cooling. Thus, the NPSHA is always greater than the NPSHR. Furthermore, the NPSHA margin is le,ss than 10% for only about a minute; thereafter, the margin increases rapidly as shown in Figure 4.5-1. Since the analysis assumptions lead to conservatively low values of NPSHA it is reasonable to conclude that cavitation of the !HSI pumps does not occur for the worst case LOCA. Even if all of the conservatisms happened to occur simultaneously the NPSHA is always greater than the NPSHR and within about a miaule the NPSHA is substantially increased. '

Time histories are shown in Figures 4.5-1 through 4.5-6 for the following key parameters:

= NPSHA

= Sump water level

= Containment total pressure a Sump vapor pressure

= Containment temperature a Sump water temperature l

= Recirculation Spray Cooler Duty 4 = Quench Pump Spray Flow Rate , i l

20 l

l

= Steam condensing coefficient The NPSHA for the centrifugal charging pump was evaluated for both the injection and recirci ation modes of operation following a LOCA. The NPSHA was determined from atmospheric pressure, the elevation head, the vapor pressure of the water in the RWST, which is at atmospheric' pressure, and the pressure drop in the suction piping from the tank to the pumps.

The only parameter in the above list which changed as a result of this analysis is the initial elevation head. However, the minimum NPSHA occurs ,

at the end of the injection mode so the original analyses are bounding.

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Table 4.2-1 Input Data To The LOCTIC Program North Anna Power' Station Reactor and Reactor Coolant Systems Parameter Maximum rated core power (MWT) including 2958 '

2% uncertainty-Internal energy of reactor coolar.t water 245.1 (includespressurizer)(MBtu)

Total water in system (1b) 4.092 x 105 Temperature (mass ave, rage excluding 586.8 pressurizer) (*F) ,

System pressure (psia) 2280.

Reactor coolant system volume 8394.

.(excluding pressurizer) (ft'), .

Pressurizer volume total (ft') 1479.9

1. Water Volume (ft') 932.9 '

2. Steam Volume (ft') 547.

1

• The fluid volume contained in the primary system, calculated at room temperature, reflects the system volume. This volume i is calculated from component dimensions, plus a factor of 1.016 for i thermal expansion and 1.014 to account for uncertainties. This gives a total factor of 1.03.

22  ;

3 y

e Table 4.2-2 Initial Conditions Assumed In The LOCA Analysis Initial Peak NPSHA Depressurization Condition- Pressure Analysis Analysis Value Value Value Containment Operating Air 12.0 8.9 (a)

Partial Pressure, psia Containment Relative Kumidity 100.0 100.0 100.0 Containment Operating Bulk 120.0 120.0 85.0 Temperature, 'F RWST Temperature, 'F 50.0 50.0 50.0 Service Water Temperature, 'F 97.0(b) 35-97 35-97 Minimum Volume of Water 435,361 435,361 435,361 Available in RWST, gal Valume of Water Exhausted at 315,252 315,252 371,075 End of LHSI Switchover Based on above, gal s

i (a)This value is dependent on the service water temperature. It varies from 9.15 psia to 12.0 psia for service water temperature of 97 ,

'F to 35'F, respectively. l (b)The initial tempercture was increased by 2 'F to 99 'F,in order to cortservatively account for service water heatup during the transient.

l 23 .

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Table 4.2-3 SLDMARY OF RESULTS OF CONTAIf@ TENT ANALYSIS SPECTRUM OF BREAK SIZES AN LOCATIONS Initial Contairument Peak contairement Time of Peak Peak containment-Broek Break Area Pressure / Temperature Pressure Pressure Temperature Location Eft 2 y gp,g,j gj e gp,gg, g,,,, g 8 7, Ptap suction 10.48 '13.69/120 43.0 192.0 270.1 Ptap suction 6.28 13.69/120 43.0 192.0 270.3 Ptap suction 3.00 13.69/120 42.7 198.0 269.8 Pump Discharge 8.25 ,

13.69/120 41.3 11.1' 268.5 Hot Lee 9.17 , 13.69/120 41.2 11.9 268.5 Notes: 1. All cases asse f ailure of one gaanch spray ptsp.

2. All breaks are full dm&le-ended ruptures IDERs ) except the 6.28-fs t and 3.00-ft I ptmp suction breaks, Alch are limited displacement rsytures.

3. Peak contairement pressure and peak contairument temperature occur concurrently.

4 Results are based on power uprate parametars listed in Tables 4.2-1 t

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~ Table 4.2-4 Sunwary of Key Pressure Results Containment- Core Uprate Current Parameter. Analysis Analysis Containment Peak- Pressure, psig - -

40.94 44.10 Containment Depressurization 3400.00 3310.00 Time, Sec Subatmospheric Peak Pressure, psig -0.01 -0.02 t

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Table 4.2-5 Chror. ology for PSOER LOCA Time Event (sec).

0.0 Accident occurs 2.4(a) Containment depressurization actuation signal 15.1 First containment peak pressure occurs 24.4 End of blowdown; core reflooding begins 30.0 Safety-injection pumps.become effective 37.0 Accumulators empty 62.5 Quench spray subsystem and casing cooling effective 204.0 Core reflooding ends; post-reflood frothing begins 302.0 Second peak containment pressure occurs; recirculation spray system becomes effective 1277.0 Post-reflood frothing ends 2830.0 Containment pressure becomes subatmospheric 3460.0 Safety injection pumps switch to recirculation mode 5350.0 Quench spray flow stops, RWST is empty 5710.0 Subatmospheric third peak containment pressure occurs (a)For a large break LOCA, the time of quench spray pump start is independent of the time the CDA signal occurs.

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TABLE 402-6 1

MASS AND ENERGY RELEASE TO CONTAll# TENT

PUMP SUCTION DER (MINIMuri ESF) i -Continued-

--GATE DATA-.----.---- --

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l 5.0 MSLB DISCUSSION AM) EVALUATION 5.1 Introduction The other design basis event for containment design is the MSLB inside

. containment. While less mass is present in a steam generator than the reactor coolant system, the fluid enthalpy is much higher so there is no intuitive way to determ'ine which transient, the LOCA or the MSLB, presents more of a challenge to the containment structure and containment safety systems.

The traditional MSLB analysis for containment response considers only the clouble ended rupture of the main steam line upstream of the flow restrictor (i.e. 4.6 ft ) at 2

hot standby conditions. The hot zero power condition is more limiting because the temperature of the fluid in the steam generator is nearly the same as the full power value and the mass is highest at the no load condition.

The analysis reported here differs from the traditional approach in one significant way which makes it necessary to consider a spectrum of MSLB scenarios. The mass and energy release data were calculated using the LOFTRAN code' which has an approved entrainment model for steam line break analysis'. The use of an entrainment model requires a spectrum analysis because only larger break sizes contain water entrained with the steam. The amount of entrained water decreases as the break size decreases. The mass and energy data from LOFTRAN were input to the the a 53

1 1

I LOCTIC computer- code' to determine the containment pressure and  !

temperature response. The LOFTRAN and LOCTIC models are discussed in more

' detail in the following sections.

5.2 Spectrum of Breaks Considered The spectrum of breaks necessary to bound the_ effects of break size and power level on the mass and energy released from a ruptured steam line have been defined based'on extensive analysis'. The postulated break area can have competing effects on blowdown results. Larger breaks are more.

likely to have water entrainment however -these breaks also result in earlier protection signal generation. So, for power levels of 102%, 70%,

30% and 0% of nominal full power, five critical break sizes have been.

defined and their characteristics quantified. The break areas analyzed are defined as follows:

= A full double-ended rupture at the outlet of one steam generator nozzle

= A full double-ended rupture downstream of the flow restrictor in one steam line

= A small double-ended rupture at the steam generator nozzle having an area just larger than that at which water entrainment occurs a A small double-ended rupture at the steam generator nozzle having an area just smail enough to preclude entrainment

= A small split rupture that will neither generate a steam line isolation signal from the Westinghouse Solid State 54

l Protection System nor result in water entrainmerit in the break effluent.

The double-ended break sizes were assumed to occur at the outlet of one steam generator and downstream of the flow restrictor. Flow restrictors in the steam line inmit the effective area of a full DER to a maximum of 1.4 ft2 per steam generator 'r the break occurs downstream of the restrictors. Upstream, the outlet nozzles of the steam generator limit the effective break area to 4.6 ft 2- Table 5.2-1 presents a list of the individual case's that were analyzed.

5.3 Mass and Energy Release Analysis The key reactor coolant system variables are initial power level, RCS pressure, RCS temperature, and RCS loop flow. For this analysis, the standard 2% uncertainty on power level was used. The thermal design flow was assumed along with the nominal RCS pressure of 2250 psia. The RCS average temperature was assumed to be 4 'F above the nominal value to account for measurement and control system uncertainties. Table 5.3-1 summarizes the RCS initial conditions.

The core kinetic parameters were chosen to simulate end-of-cycle conditions with the most reactive rod stuck out of the core. These assumptions maximize th'e positive reactivity insertion due to moderator feedback during cooldown. Additionally, minimum safety injection was assumed to restrict the flow of borated water to a rate corresponding to the operation of one charging pump. The safety injection lines downstream -

55

1 l

l 1

l of the. boron injection tank are assumed to have a zero boron concentration. 'These assumptions minimize the magnitude of the negative reactivity inserted.

The assumptions regarding the secondary system are intended to produce conservative results. The main feedwater system is designed to maintain feedwater flow equal to steam flow. Therefore, the feedwater flow rate increases following a steam line break resulting in several effects.

First, the steam pressure is lower due to the presence of subcooled water in the generator. Second, the heat transfer from primary to secondary is increased. Finally, for large break cases the increased feedwater flow increases the amount of entrained water in the steam exiting the break.

Since these are competing conditions it is not possible to define the worst feedwater transient for all plant conditions. Therefora in order to insure conservative results each of the above parameters were defined at its least positive or most negative extreme.

The feedwater flow rate was conservatively modeled by assuming an increase in response to the steam line break. For split breaks and small double-ended ruptures feedwater flow was increased proportionally to the steam line flow increase. For the large double ended rupture cases the feedwater flow was instantaneously ramped to a maximum of 220% of nominal full feedwater flow in response to the decreasing steam generator pressure.

There are inany other less significant secondary parameters which were modeled conservatively including: auxiliary feedwater flow rate, initial ,

56

steam generator fluid mass, critical flow model loss coefficient and steam line blowdown volume. These parameters are specified in Table 5.3-2.

The interface between the primary and the secondary systern was also modeled conservatively. Reverse heat transfer from the intact steam generators to the primary loop was modeled which resulted in the release of more energy to the containment.

Various system component failures were evaluated to determine which failure results in the. largest increase in releases to the containment.

The failure of one safeguards train to operate was assumed in the analysis along with the failure of the non-return valve in the steam line with the faulted steam generator. The safeguards train failure reduces boron delivery to the core while the non-return valve failure allows the steam generators to blowdown until the main steam isolation valves on the intact loops are isolated. Since the main steam trip valves at the North Anna power Station do not prevent reverse flow, the nonisolatable volume in the main steam system continues to blowdown even after steam line isolation occurs.

Mass and energy release rates were generated for each of the cases listed in Table 5.2-1. Generally, the transients are characterized by rapid increases in mass flow rate and energy flow rate lasting a few seconds and begin to exponentially decrease. The mass flow rate is largest for the 4.6 ft DER 2

breaks and the hot zero power cases. The energy release rate is largest for the 4.6 ft8 break and for the 102 %

power cases. The actual data aie presented in WCAP-11431'.

57 L

5.4 Containment Pressure and Temperature Analysis Containment response calculations were performed using the LOCTIC computer code to determine pressure and temperature response to a main steam'line break inside co;itainment. The LTTIC input is chosen based on conservative assumptions that include the following:

= Eight percent partial revaporization of condensate containment thermodynamic model

=

Failure'of the nonreturn valve on the broken steam line to close (the most limiting single failure).

= Minimum quench spray failure initiation 60 seconds after containment depressurization actuation (CDA).

= Initial containment conditions that yield a maximum pressure and temperature including: air partial pressure, bulk temperature and dewpoint.

= Auxiliary feedwater initiation inrnediately (0 sec) after the break with a rate of 900 gpm to the affected steam generator.

l Table 5.4-1 presents a summary of peak containment pressure and temperature analysis initial conditions. The initial conditions shown in the table are conservative for the desired result. For example, when containment peak pressure is calculated, the maximum operating l containment pressure is the conservative initial pressure condition; l whereas, when containment peak temperature is calculated, the minimum operating containment pressure is the conservative initial pressure -

l i

58 L

condition. The initial air temperature is conservative for the peak temperature analysis because the transient starts at the highest possible temperature. For the peak pressure analysis the containment heat sinks play a dominant role in the magnitude of the peak pressure and therefore are initialized at the highest operating temperature. A conservative, maximum service water temperature was used for the MSLB analyses although this parameter has a negligible affect on the peak temperature and pressure because of the relatively long period of time required for the recirculation spray sys,tems to become operable.

Single failure considerations were applied to the containment response calculations in a manner consistent with the mass and energy release calculations. In particular, the failure of one emergency bus was assumed. For the containment analysis this assumption results in the failure of one inside recirculation spray pump, one outside recirculation spray pump and one quench spray pump. The single failure assumptions made in the LOFTRAN analyses are inherent in the mass and energy data. In effect, two single failures have been included in tha main steam line break results: the failure of the non-return valve and the failure of one emergency bus. This approach is certainly conservative but not overly conservative because the two effects occur at different times. The non-return valve failure affects the early portion of the transient while the emergency bus failure affects primarily the later stages of the transient.

The loss of offsite power was also treated in a conservative fashion.

The mass and energy release calculations assume the reactor coolant pumps .

59

remain functional (i.e. no loss of offsite power) to provide the maximum heat transfer from the primary to the secondary. The containment response calculations provide for tre loss of offsite power by assuming containment spray delays representative of a loss of offsite power condition. This approach represents another inherent conservatism in the analysis because the offsite power must affect all of the plant systems consistently.

The results are summarized in Table 5.4-2 which show the peak containment temperature and peak pressure for each of the cases that were analyzed. As shown in'the table, the limiting steam line break for both containment temperature and pressure is the DER of the 30 inch ID (4.6 ft )2 line upstream of the flow restrictor at hot zero power. The next

nost limiting break is the size just small enough to preclude water entrainment.

The peak containment temperature of 357.4 'F is achieved with initial containment conditions that minimize the initial containment air and

! water vapor masses. The peak temperature exceeds the containment design temperature of 280 'F for a short period of time. Based on the large heat capacity of the structures and the fact that condensation occurs on the l surface, the temperature of the structure will not exceed the saturation l

temperature.

l A summary of the results obtained for each of the cases is shown in Table 5.4-2. The largest break size is limiting at each of the power levels involved. The hot zero power analysis provides the limiting pressure and temperature. Figure 5.4-1 presents the pressure response .

60

for the limfLing case. The limiting temperature response is also presented in figure 5.4-2.

5.5 Equipment Qualification The effects of the proposed changes in allowable containment temperature and pressure on the environmental qualification of electrical equipment have been evaluated as required by 10 CFR 50.49. The evaluation addressed the following environmental qualification parameters: equipment operation time, accident environmental conditions and effects on equipment qualification, and equipment aging effects including service '

Itfe and maintenance schedule. Based on this evaluation it has been concluded that the environmentally qualified electrical equipment which is located inside containment is qualified for the new containment temperature and pressure conditions. Further, it has been determined that the previously established equipment lifetimes and maintenance schedules will remain valid provided the annual operation time as a function of containment temperature does not exceed a limit derived using the Arrhenius methodology. Appropriate periodic test procedures will be written to monitor containment temperature for coneparison against the annual operation time versus temperature limit and equipment service lifetimes as well as maintenance schedules will be adjusted as required.

61

4 k

-Table 5.2-1 Main Steam Line Break Case Sumary Case Power DER Size Split Size

1. -(%FP). ft2 ft8 4

1 102 4.6 -

'2 102 1.4 -

3 102 0.7 -

4 102 0.6 -

5 102 , -

0.645 6 70 ' 4.6 -

7 70 1.4 -

8 -70 0.6 -

9 70 0.5 -

t 10 70 -

0.681

. 11 30 4.6 -

12 30 1.4 -

13 30 0.5 -

l 14 30 0.4 -

. 15 30 -

0.707 i 16 0 4.6 -

17 0 1.4 -

18 0 0.2 -

19 0 0.1 -

20 0 -

0.300 1

l 62 ,

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Table 5.3 Reactor Coolant System Paramete'rs Parameter Value NSSS Power, MWt 2910  ;

n Reactor Power, MWt 2898 ,

Reactor Coolant Pump Heat, MWt 12 Vessel Flow, MWt 278,400 Pressurizer Pressure, psia 2250.

Reactor Coolant Temperatures, 'F Core Outlet 624.0 Vessel Outlet 621.2 Core Average 590.4 Vessel Average 586.8 Vessel / Core Inlet 552.3 Steam Generator Steam Temperature, 'F 525.2 Steam Pressure, psia 850 Zero Load Temperature, 'F 547.0 63

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i Table 5.3-2 Secondary System Parameters-Parameter Value

- Auxiliary Feedwater Flow Rate, gpm Intact Loops 350-Faulted Lodp 900 Steam Generator Fluid Mass, % of prog ivl +5 Unisolated volume per feedwater line, ft' 404 Auxiiiary feedwater isolation time, minutes 30 Auxiliary feedwater purge volume, ft' 40 Auxiliary feedwater isolation delay, seconds 60 Moody contraction coefficient, fL/D 0 Unisolated steamline volume, ft' 7,283 64

L Table 5.4-1 MSLB Analysis Initial Conditio'ns-Initial Condition Peak Pressure - Peak Temperature Analysis Value Analysis Value Containment Operating Air 11.2* 8.9 Partial Pressure, psia Containment Relative Humidity, % 100.0 50.0 Containment Operating Bulk 120.0 120.0 Air Temperature, 'F RWST Temperature, 'F 50.0 50.0 Servica Water Temperature, 'F 97.0 97.0

• This value is from the limiting break size analysis, other limiting values were established at different power levels (e.g. 30 % power - 12.2 psia; 70 % power - 12.8 psia; and 102 % power - 13.5 psta).

r i

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65

[ .

Table 6.4-2 Sumary of Main Steam Line Break Reselts Case Peak Temperature Analysis Peak Pressure Analysis Temperature and Time Pressure and Time 1 356.0 9 '28.8 40.44 9 79.5 2' 322.6 9 18.5 35.76 9 170.0 3 340.8 9 61.0 33.86 9 324.0 4 352.2 9 43.8 32.82 9 392.0 5 348.6 9 118.0 33.29 9 298.0 i

6 355.2 9 27.9 42.52 9 92.0 7 329.1 9 65.0 39.82 9 220.0 8 224.8 9 76.0 33.01 9 495.0 9 345.7 9 39.4 30.71 9 600.0 10 346.9 9 98.0 34.04 9 460.0 L

11 356.5 9 22.0 43.27 9 124.0  :

12 326.3 9 39.6 41.04 0 302.0 L 13 313.0 9 75.0 32.18 9 595.0 14 348.1 9 51.0 30.91 9 920.0 15 344.4 9 84.0 35.91 9 435.0 16 357.4 9 18.0 44.90 9 94.0 17 329.2 9 15.5 42.35 9 214.0 l

l 18 302.6 9 97.0 15.75 9 358.0 19 299.5 9 190.0 13.47 9 190.0 I

t 20 309.6 9 138.0 21.40 91030.0 9 .

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6.0 CONCLUSION

S The design basis transients have been analyzed to support a change in the envelope of acceptable air partial pressure 'and service water temperature conditions. In addition, the containment bulk operating air temperature upper limit was increased to 120 'F and the maximum service water temperature analyzed was 97 'F. Each of these parameters are key inputs to the containment design basis transients.

Additionally, the minimum volume of RWST water available was decreased to support a lower minimum technical specification limit on this parameter.

The LOCA analysis considered a spectrum of break sizes and locations.

Several different single failure scenarios were also considered. The LOCA analysis actually consists of a peak pressure analysis, a depressurization and a third peak analysis as well as NPSHA analyses for the rec.f rculation spray pumps and the LHSI pumps. The initial conditions were determined for each of these analyses to provide the most conservative result. The peak containment pressure was found to be 44.1 psig. The length of time to subatmospheric conditions was found to be 3310 seconds. The maximum subatmospheric peak was -0.02 psig. The NPSHA analyses showed 5.8 ft and 2.5 ft margin for the outside and inside recirculation spray pumps respectively and 0.1 ft margin for the LHSI pumps. ,

j 69

The MSLB scenarios involved several combinations of break size, break type and power level. In all twenty cases were run for the peak temperature cases and then repeated for the peak pressure cases. The large number of cases were required by the use of an entrainment model which makes it impossible to intuitively determine the limiting break size and power level. The results indicate that the hot zero power, 4.6 ft*

DER of the main steam line is limiting for both temperature and pressure.

The peak pressure was fenind to be 44.9 psig and the peak temperature was found to be 357.4 'F. ,

The above results indicate that the containment design criteria are not violated at the initial conditions of 120 'F containment temperature, 97 'F service water temperature and the lower volume of RWST water.

Operability of the electrical and mechanical equipment within the ,

containment at temperatures up to 120 'F has also been considered and found to be acceptable provided procedural changes are implemented.

Therefore, the appropriate technical specif';ations can be changed to permit the' higher containment air temperature and the lower minimum RWST volume. The maximum service water temperature while analyzed at 97 'F is not going to be changed at this time because the service water system modifications recently put in place can maintain the reservoir temperature below the currently allowable 95 'F.

Therefore, the containment design criteria of

= Peak pressure < 45 psig

= Depressurize to subatmospheric < 3600 sec -

70

1

= Maint!.in pressure subatmospheric for duration of the accident are- not violated and a technical specification change governing containment air temperature and minimum RWST volume - is technically acceptable.

7.0 10 CFR 50,59 EVALUATION The proposed changes have been reviewed against the criteria of 10 CFR 50.59 resulting in the conclusion that an unreviewed safety question does not exist. This determination was reached based on the following specific considerations:

1. The probability or consequences of any UFSAR event do not increase.

Accident probability is independent of the initial conditions maintained by T.S. 3/4.3.5.5, 3/4.6.1.4, and 3/4.6.1.5. These ,

specifications deal with RWST volume, air partial pressure and air i teniperature. A change to any of these parameters does not affect accident probability because they are unrelated to pipe ruptures which are the design basis events for the containment, Accident consequences are not increased by the proposed Technical Specification changes. As shown in the safety evaluation, the containment design criteria are met for both the LOCA and MSLB transients which form the containment design basis. Therefore, the consequences are not increased by these changes. In fact, the transient peak containment temperature is substantially lower as a 71

. result of the entrainment assumption used in the MSLB analysis.

Finally, T.S. 83/4.5.5, 83/4.6.1.4 and 3/4.6.1.5 have been changed to state that the appropriate instrument uncertainties have been considered in the safety analyses which provides further evidence that accident consequences are not increased by these changes.

The changes proposed for T.S. 3/4.6.1.2 and 3/4.6.1.3 are submitted to reflect the peak pressure resulting frorn the revised LOCA analysis.

Similarly, B3/4.6.1.6 is submitted with the numerical value of the peak pressure rembved rather than list the 44.1 psig value. This approach maintains the intent of the specification and enables future changes to the peak pressure without changing the BASES section.

2. No new or different accident type is generated as a result of the revised containment initial conditions specified in T.S. 3/4.3.5.5, 3/4.6.1.4 and 3/4.6.1.5. The changes proposed to these specifications do not involve hardware modifications but changes to the initial conditions assumed in the safety analysis. Therefore, the proposed changes do not involve alterations to the physical plant which would introduce any new or unique operational modes or accident precursors.

3. The margin of safety is not reduced. The proposed changes to T.S.

and 3/4.3.5.5, 3/4.6.1.4 and 3/4/6.1.5 have been incorporated into the safety analysis for the containment. The results of the analysis show that none of the containment design bases are violated. That is, the following inequalities remain valid: ,

72

= Peak pressure: 44.1 psig(LOCA), 44.9 psig (MSLB) < 45 psig

= Depressurization to subatmospheric: 3310 seconds < 3600 seconds

= Maintain subatmospheric pressure: -0.02 psig < 0.0 psig Operability of the electrical equipment within the containment at temperatures up to 120 'F has also been considered and found to be acceptable provided procedural chenges are implemented. Procedural changes are required to ensure that maintenance schedules are adjusted to maintain established equipment lifetimes.

Additionally, the Net Positive Suction Head Allowable (NPSHA) exceeds the Net Positive Suction Head Required (NPSHR) for the recirculation spray pumps and the LHSI pumps. Finally, the peak containment temperature frm the steam line break is substantially lower than previously reported, even with the revised initial conditions, because of the entrainment assumptions.

73

References

1. Letter from Mr. V. L. Brownlee (NRC) to Mr. W. L. Stewart (VP)

Subject:

Report Nos. 50-338/84-43 and 50-339/84-43, Decen6er 20, 1984.

2. Hynek, S. J. "Loss-of-Coolant Transient Inside Containment (LO; TIC),"

User Manual Reissued February 1982 by E. W. Chui, Stone and Webster Engineering Corporation, (NU-100).

3. ANSI /ANS 5.1-1979,'"American National Standard for Decay Heat Power In Light Water Reactors," Approved August 29, 1979.

4. North Anna Power Station Units 1 and 2, Updated Final Safety Analysis Report, Revision 4 June, 1986, pages 6.2-256 through 6.2-279.

5. Burnett, T.W.T. , et al. , "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary),WCAP-7907-A(Non-Proprietary), April,1984.

6. Land, R. E., "Mass and Energy Releases Following A Steam Line Rupture," WCAP-8822 (Proprietary), WCAP-8860 (Non-Proprietary),

September, 1976.

7. Butler, J.C. "Mass and Energy Releases Following A Steam Line Rupture For North Anna Units 1 and 2," February, 1987, WCAP-11431 (Proprietary), WCAP-11432 (Non-Proprietary).

74

Attachment 4 10 CFR 50.92 Evaluation 4

l i

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10 CFR 50.92 EVALUATION The proposed changes do not involve a significant hazards consideration for North Anna Units 1 and 2'for the following reasons:

1. The probability or consequences of any UFSAR event do not increase.

Accident probability is independent of the initial conditions maintained by T.S. 3/4.3.5.5, 3/4.6.1.4, and 3/4.6.1.5. These specifications deal with RWST volume, air partial pressure and air temperature. A dhange to any of these parameters does not af fect accident probability because they are unrelated to pipe ruptures which are the design basis events for the containment.

Accident consequences are not increased by the proposed Technical Specification changes. The containment design criteria are met for both the LOCA and MSLB transients which form the containment design basis. Therefore, the consequences are not increased by these changes. In fact, the transient peak containment temperature is substantially lower as a result of the entrainment assumption used in the MSLB analysis. Finally, T.S. B3/4.5.5, 83/4.6.1.4 and 3/4.6.1.5 have been changed to state that the appropriate instrunent uncertainties have been considered in the safety analyses which provides further evidence that accident consequences are not increased by these changes.

The changes proposed for T.S. 3/4.6.1.2 and 3/4.6.1.3 are submitted to reflect the peak pressure resulting from the revised LOCA analysis. -

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Similarly, B3/4.6.1.6 is submitted with the numerical value of the peak pressure removed rather than list the 44.1 psig value. This approach maintains the intent of the specification and enables future changes to the peak pressure without changing the BASES seution.

2. No new or different accident type .is generated as a result of the revised containment initial conditions maintained in T.S. 3/4.3.5.5, 3/4.6.1.4 and 3/4.6.1.5. The changes proposed to these specifications do not involve hardware modifications but changes to the initial conditions assumed in the safety analysis. Therefore, the proposed changes do not involve alterations to the physical plant which would introduce any new or unique operational modes or accident precursors.

3. The margin of safety is not reduced. The proposed changes to T.S.

3/4.3.5.5, 3/4.6.1.4 and 3/4/6.1.5 have been incorporated into the safety analysis for the containment. The results of the analysis show that none of the containment design bases are violated. Margin remains between each of the design basis limits and the calculated value of the limit. Net positive suction head available exceeds that required for the recirculation spray pumps and the LHSI pumps.

Furthermore, the peak containment temperature from the steam line break is substantially lower than previously reported, even with the revised initial conditions, because of the entrainment assumptions.

Operability of the electrical equipment within the containment at temperatures up to 120 'F has also been considered and found to be 1

acceptable provided procedural changes are implemented. Procedural

l changes are required to ensure that maintenance schedules are adjusted to mintain established equipment lifetimes.

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ATTACBGMT 5

-MEASURED VERSUS PROJECTED LEAK RATE TEST RESULTS Projected- Proposed Nuclear -Test Test Measured Leak Rate Tech. Spec.

Unit Type Pressure Leak Rate At 44.1 PSIG Limit 1 Integrated 43.7 psig .0326%/ day .0329%/ day .1%/ day 1 Local 44 psig 171.6 SCFM* 171.8 SCFM 182.5 SCFM**

2 Integrated - 44.2 psig .0692%/ day *** .1%/ day 2 Local 44 psig 171.6 SCFM* 171.8 SCFM 182.5 SCFM**

• It was_ assumed for purposes of this calculation that the measured local leak rate (combined) was at the current Tech. Spec.. limit of 0.6 L a

(171.6 SCFM at 40.6 psig)

• • The proposed Tech. Spec. local leak rate limit is 0.6 L at 44.1 psig a

which corresponds to 182.5 SCFM.

• • • The last integrated leak rate test for Unit 2 was performed at a pressure greater than that required by the proposed Tech. Spec.

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