ML20063P372

Proposed Tech Specs 3.4.4 & 4.4.4 Re RCS Pressurizer

ML20063P372
Person / Time
Site:	Calvert Cliffs
Issue date:	09/29/1982
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20063P367	List: ML20063P366 ML20063P372
References
NUDOCS 8210130342
Download: ML20063P372 (44)

Text

,

ATTACHMENT I

'. l PROPOSED TECHNICAL SPECIFICATIONS REACTOR COOLANT SYSTEM PRESSURIZER r

LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERA 8LE with a steam bubble and with at least 150 kw of pressurizer heater capacity capable of being supplied by 2:

P 5: d th'r _ 5 pre :t f-

--:i ::1 :. & c- '== ' : :?x g&QSQ emergency power.

" t; p. of. _-

APPLICABILITY: MODES 1 and 2.

ACTION:

With the pressurizer inoperable due to an inoperable emergency power a.

supply to the pressurizar heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in H0T SHUTDOWN within the following 12 a

hours.

b.

With the pressurizer otherwise inoperable, be in at least HOT STAND 8Y with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

SURVEILLANCE REQUIREMENTS 4.4.4 The pressurizer water level shall be determined to be within _ 5 ;r ewi; ef its p..g: :d ::!:: at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Ydt ab 4_% d l

CALVERT CLIFFS - UNIT 1 3/4 4-5 Amendment No. 53 CALVERT CLIFFS - UNIT 2 Amendment No. 36 i

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'8210130342 820929'

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PDR ADOCK 05000317 P.

PDR s

7.

o The pressurizer level shall be maintained within an operating band between 133 and 225 inches except when three charging pumps are operating and letdown flow is less than 25 GPM. If three charging pumps are operating and letdown flow is less than 25 GPM maximum pressurizer level shall be limited to I:-- th^n '".0 !;:P. :. h f 33 Q Z 10 d,

s, _

6 REACTOR COOLANT SYSTEM (dB5 cr BASES limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operat-ing at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint-(Pressurizer Pressure-High)'is reached. (i.e., no credit is taken' for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power o2erated relief valve or steam dump valves.. -

Demonstration of the safety valves' ' lift settings will occur only during -shutdown Land will be performed in accordance with the provisions of Section XI of the ASME Soiler and Pressure Vessel Code.

3!4.4.3 RELIEF VALVES T"e power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leak-8 age path.

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[Ci, 3/4.4.4 PRESSURIZER s

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A steam bubble in the pressurizer with level as programmed ensures that the RCS is not a hydraulically soli ys, tem and is capable of accommo-dating pressure surges during operation.

he lege=d lev b also protects the pressurizer code safety valves and power operated relief valve against water relief.

The power operated relief valves function to relieve RCS pressure during all design transients.

Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer--Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizar code safety valves.

The requirement.that 150 kw of pressurizar heaters and their associated l

contrcls be capable of being supplied electrical. power from an emergency bus provides assurance that these heaters can be energized during a loss of off-site cower condition to maintain natural circulation at HOT STANDBY.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure tnat the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.33, Revision 1.

Inservice inspection of steam generator tubing is essential in order to CALVERT CLIFFS - UNIT 1 3 3/4 4-2 Amendment No. 34, 53 E

CALVERT CLIFFS - UNIT 2 Amendment.o.

76, 26 J

4

@ The operating band for pressurizer level bounds the programmed ensures that RCS pressure remains within the bounds of an analyzed condition during the excessive charging event as well as during the limiting deptessurization event,fxcessbad. The operating band...

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ATTACMENT II The following study was perfomed to determine and justify Technical Specification limits for pressurizer level. The approach taken was to reanalyze the Loss of Load event and the Excessive Charging event to detemine the maximum pressurizer level and evaluate the Excess Load event to determine the minimum level. The Steam Generator Tube Rupture event was also analyzed to detemine site boundary doses, which increase-i due to higher primary to secondary leak.

The limiting pressurization event, Loss of Load (LOL),was analyzed such that the event initiated from a given pressurizer level does not cause the RCS pressure to exceed the design limit of 2750 psia. The Excessive Charging event initiated from a given pressurizer level was analyzed to assure that the operator has at least fifteen (15) minutes from initiation of a high pressurizer level alam to take corrective actions and terminate the event prior to filling the pressurizer solid. The maximum pressurizer level is then selected based on the results of the limiting event.

The limiting depressurization event, Excess Load, was analyzed to assure that the draining of the pressurizer and subsequent Reactor Vessel Upper Head (RVUH) voiding do not affect primary coolant circulation and do not result in exceeding fuel design limits. Hence, the Excess Load event was analyzed to determine the_ Technical Specification limit on the minimum pressurizer level.

1 Finally, a reanalysis of the Steam Generator Tube Ruputure evekt was performed to predict the increased site boundary doses, since the increased pressurizer level results in higher primary to secondary leah.

Detailed description and results of each analysis are attached. Based on the results of this study, it is concluded that:

1.

Based on the analysis of the Excess Load event, a minimum pressurizer level of 270 ft3 (this level corresponds to the level required to cover pressurizer heaters) is acceptable.'

l 2.

Based on the analysis of the Excess Charging event and the Loss of Load event, a range of maximum pressurizer level (915 ft3 to 975 ft )

3 is acceptable.

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1. - Loss of Load Event 3 in order to determine that A reanalysis of the Loss of Load Event was initiated at 975 ft the RCS pressure upset limit of 2750 psia is not exceeded. The transient DNBR was also evaluated to assess that the results are within the design limit of 1.23.

The assumptions used to maximize RCS pressure during the transient are:

a.

The event is assumed to result from the sudden closure of the turbine stop ' valves without a simultaneous reactor trip. This assumption causes the greatest reduction in the rate of heat removal from the reactor coolant system and thus results in the 1

l most rapid increase in primary pressure and the closest approach to the RCS pressure upset limit.

1 b.

The steam dump and bypass system, the pressurizer spray system, and the power operated pressurizer celief valves are assumed not to be operable.

This too maximizes the primary system pressure reached during the transient.

The Loss of Load Event was initiated at the. conditions shown in Table 1-1.

The combination of parameters shown maximizes the calculated peak RCS pressure. As can be inferred from the table, the key parameters for this event are the initial primary and l

secondary pressures and the moderator and fuel temperature coefficients of reactivity.

l The initial core average axial power distribution for this analysis was assumed to be a bottom peaked shape. This distributica is assumed because it minimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure The Moderator Temperature Coefficicat (MTC) of +0.5 x 10~gnd heat flux increases.

i Ap / F was assumed in this-analysis. This MTC, in conjunction with the increasing coolant temperatures, maximizes-t the rate of change of heat flux and the pressure at the time of reactor trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions was used in the analysis. This FTC causes the least amount of negative reactivity feedback to mitigate the transient increases in both the core heat flux and the pressure. The uncertainty on the FTC used in the analyses is shown in Table 1-1. The lower limit on initS1 RCS pressure is used to maximize the rate of change of pressu:e, and thus peak j

- pressure, following trip.

The Loss of Load Event, initiated from the conditions given in Table 1-1 results in a high pressurizer pressure trip signal at 6.2 seconds. At 10.1 seconds, the primary pressure reaches its maximum value of 2617 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 5.8 seconds, the secondary pressure reaches its maximum value of 1047 psia at 10.6 seconds after initiation of the event.

' Table 1-2 presents the sequence of events for this event. Figures 1-1 to 1-4 show the transient behavior of power, heat flux, RCS pressure, and RCS coolant temperatures.

The event was also teanalyzed with the initial conditions listed in Table 1-3 to determine that the acceptable DNBR Limit is not. exceeded.

The minimum transient DNBR calculated for the event is 1.34, compared to the design limit of 1.23.

The results of this analysis deponstrate that during a Loss of Load Event, initiated from l

l a pressurizer level of 975 ft, the peak RCS pressure and the minimum DNBR do not exceed their respective design limits.

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4 TABLE.1-1 KEY PARAMETERS ASSUMED IN Tr!E LOSS OF LOAD A!!ALYSIS TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Reference

• This Parameter Units Cycle Analysis Initial Core Power Level MWt 2754 2754 O

Initial Core Inlet Coolant F

550 550 Temperature 6

Core Coolant' Flow x10 1bm/hr 133 9 133 9 Initial Reactor Coolant psia 2200+

2155++

System Pressure Initic Steam Generator psia 864 864 Pressure Moderator Temperature x10-4ap / F

+0.5

+0.5 0

Coefficient Doppler Coefficient

-0.85 0.85 Multiplier CEA Worth at Trip b

-4.7

-4.7 Time to 90% Insertion sec 31 31 of Scram Rods Reacter Regulating System Operating Mode Manual Manual Steam Dtnp and Bypass Operating Mode Inoperative Inoperative l

• Unit 1, Cycle 6

+ Corresponds to Technical Specificatien minimum indicated pressure of 2225 psia. The value includes an uncertainty of 25 psia.

++ Corresponds to Technical. Specification minimum indicated pressure of 2200 psia. The value includes an uncertainty of 45 psia.

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TABLE 1-2 KEY PARAMETERS ASSUMED IN THE LOSS CF LOAD ANALYSIS TO CALCULATE TRANSIENT MINIMUM DNBR Reference This Parameter Units Cycle Analysis Initial Core Power Level MWt 2700**

2700**

O

. Initial Core Inlet Coolant F

548**

548**

Temperature 6

Core Coolant Flow x101bm/hr 138.5**

138.5**

Initial Reactor Coolant psia 2225**

2200**

System Pressure Initial Steam Generator psia 864 864 Pressure Integrated Radial Peaking 1 75**'+

1 75**'+

Factors, F A (Bank 5 Inserted 25%)

Moderator Temperature x10-Mao / F

+0.5

+0.5 0

Coefficient Doppler Coefficient 0.85 0.85 Multiplier CEA Worth at Trip

% ap

-4.7

-4.7 I

Time to 90% Insertion sec 31 31 of Scram Rods l

Reactor Regulating Operating Mode Manual Manual l

System i

Steam Dump and Bypass Operating Mode Inoperative Inoperative l

System

• Unit 1, Cycle 6

• • Effects of uncertainties on these parameters were accounted for statistically.

+7he values assumed are conservative with respect to the Technical Specification limits.

TABLE ~ 1-3 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Time (sec)

Event Setpoint or "value

-0.0 Loss of Secondary Load 5.8 Steam Generator Safety Valves 1000 psia Open

~

6.2 High Pressurizer Pressure Trip 2422 psia Signal Generated 7 6. -

CEAs Begin to Drop Into Core 79 Pressurizer Safety Valves open 2500 psia

~ 617 paia 2

10.1 Maximum RCS Pressure 10.6 Maximum Steam Generator Pressure 1047 psia 12 3 Pressurizer Safety Valves are 2500 psia Fully Closed 4.

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LOSS OF LOAD EVENT Figure

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REACTOR COOLANT SYSTEM PRESSURE VS TIME 1-3

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REACTOR COOLANT SYSTEM TEMPERATURES VS TIME 1-4

4 2.

Excessive Charging Event The Excessive Charging event is assumed to occur'by inadvertent initiation of charging flow. The event initiated from maximum pressurizer level was analyzed to assure that the operator has at least fifteen (15) minutes

~ ~ 'from initiation of a high pressurizer level alarm to take corrective I

action and teminate the event prior to filling the pressurizer solid.

The time required to fill the pressurizer solid was calculated using

- Equation 2-1.

T=YS~YSL - YT (Eq. 2-1)

(FCH-FLD) h where:

V

• steam volume in the pressurizer 3

V3g = equivalent saturated liquid volume of pressurizer steam volume V

= volume above the spray nozzles T

FCH = charging flow rate FLD = letdown flow rate vj = specific volume of liquid at charging and letdown conditions 4

v2 = specific volume of liquid at pressurizer conditions The analysis was perfomed for three combinations of charging and letdown flows. Table 2-1 presents the initial conditions assumed in the analysis and the results of the analysis. As seen from the table, all three combina-tions of charging and letdown flows analyzed provide at least fifteen l

minutes after initiation of high level alam for the operator to take l

corrective actions and terminate the event prior to filling the pressurizer solid.

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TABLE 2-1 naximum Initial III Pressurizer Level High Level Alarm Time to Fill Volume Control Assumptions Assumed in Analysis Analysis Setpoint Pressurizer Charging Flow Letdown Flow Liquid Volume LevefN Liquid Volume Level (3)

(minutes) 3 3

(GPM)

(GPM)

(ft )

(in)_

(ft )

(in)

-l 1.

132 0

915 227 920 228 15 2.

132 29 975(2) 242 1040 257 15 3.

88 0

975(2) 242 1100 272 15 i

IIIFrom time of initiation of high pressurizer level alarm (2)tiaximum limit based on Loss of Load event (3) Referenced to the 1" level nozzle at.the bottom of the pressurizer

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3.

Excess Load Event 3

The Excess Load event initiated at a pressurizer level of 270 ft was analyzed to evaluate the impact of reactor ressel upper head (RVUH) voiding on fuel design limits and on reactor coolant circulation. The analysis included the automatic initiation of auxiliary feedwater three minutes after initiation of reactor trip signal and a manual trip of the reactor coolant pumps (RCPs) following a safety injection actuation signal (SIAS) due to low pressurizer pressure. The RCP coastdown results in a proportionately reduced RVUH flow until natural circulation is established; at that time all flow to the RVUH is assumed to terminate.

The Excess Load event is initiated by the instantaneous opening of steam dump and bypass valves which have a combined capacity of approximately 45% of the nominal full power steam flow. This Excess Load event persists until the steam generators are isolated on steam generator isolation signal (SGIS) due to low sm:ondary pressure. The full power event maxi-mizes primary cooldown, shrinkage and consequently RVUH voiding.

The magnitude of RVUH voiding at full power is significantly greater than at zero power because of the higher RVUH coolant temperatures and the larger primary coolant shrinkage that occurs at the higher system tem-peratures. Therefore, only the full power excess load transient results are described herein.

The key parameters assumed in the analysis are given in Table 3-1.

The analysis conservatively assumed a Moderator Temperature Coefficient of

+0.5x10-%o/*F. This MTC,in combination with decreasing coolant temperatures; inserts negative reactivity and causes the core power to decrease. The decreasing core power does not allow either the High Power trip or TM/LP trip to be initiated,and thus the time of reactor trip is delayed until a low pressurizer pressure trip (i.e., floor of the TM/LP j

trip) is generated. The longer time required to initiate reactor trip causes the pressurizer to drain and thus maximizes RVUH voiding.

The analysis conservatively assumed that all three charging pumps were inoperable and that one High Pressure Safety. Injection (HPSI) pump fails to start on SIAS due to low pressurizer pressure.

The effect of auxiliary feedwater was explicitly evaluated by analyzing the event both with and without auxiliary feedwater initiated three minutes after reactor trip signal is generated. An auxiliary feedwater flow of 172 lbm/see to each steam generator is conservatively assumed (i.e.,10.5% of full power main feedwater flow per generator). The maximum auxiliary feedwater flow causes the fastest primary cooldown and thus enhances the bubble fonnation in.the upper head.

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This analysis shows that some RVUH voiding will occur as a result of the RCS depressurization caused by an Excess Load transient. Voiding the RVUH starts when RCS pressure control is lost due to the primary coolant shrinkage which drains the pressurizer. During the period of RVUH voiding, RCS pressure is controlled by the saturation pressure within the upper head. This reduces the RCS depressurization in the latter part of the transient as compared to analyses which do not explir.itly model RVUH voids.

RVUH coolant temperatures initially follow the core outlet temperatae and therefore decrease innedtately following a reactor trip. The decrease in RVUH coolant flow reduces the convective heat transfer and inhibits upper head cooldown. This leads to elevated RVUH coolant temperatures which raise the upper head saturation pressure and therefore increase RCS pressure during the period of voiding. The increased RCS pressure does not adversely impact the approach to any SAFDL; however, safety injection flow is decreased. The reduced safety injection flow does not result in a return to criticality; however, the decreased flow diminishes the mitigating effect of safety injection on coolant shrinkage and there-fore enhances voiding. Subsequent RYUH cooldown is accomplished through an exchange of coolant between the RVUH and the core outlet plenum.

This exchange of coolant is driven by the expansion and contraction of the steam '..cbble.

Additional RVUH cooling is accomplished through heat conduction across the upper guide structure.

At the time of maximum RVUH voiding approximately 63% of the head is occupied by steam. Since this steam bubble does not expand beyond the upper head, primary coolant circulation is unaffected. Tables 3-2 and 3-3 present the sequence of events for the event initiated without and with auxiliary feedwater flow. Figures 3-1 through 3-14 present the transient behavior of the system variables during the event. The analysis demonstrates that the addition of auxiliary feedwater prolongs the duration of RVUH voiding and delays repressurization of the RCS. However, since auxiliary feedwater is delivered after the time of maximum voiding the j

peak void fraction is unchanged.

In conclusion, the potential RVUH voiding associated with an Excess Load transient initiated from a pressurizer level of 270 ft3 does not extend beyond the upper head and therefore will not affect primary coolant circulation.. In addition, approach to SAFDLs are not impacted by RVUH yoiding. The results of the analysis also shows that the NSSS achieves stable conditions and that shutdown cooling procedures can be initiated if deemed necessary.

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TABLE 3-1 KEY PARAMFTERS ASSUMED IN THE EXCESS LOAD EVENT ANALYSIS Parameter Units Value Initial Core Power Level MWt 2754 Core Inlet Temperature

'F 550 Reactor Coolant System psia 2300 Pressure 6

Core Mass Flow Rate x101bm/hr 133.6 Moderator Temperature x10-4

+.5 e

Coefficient CEA Worth Available at Trip

-4.7 Auxiliary Feedwater Flew Rate Ibm /sec 172.0/S.G.

Low Pressurizer Pressure psia 1728 Analysis Trip Setpoint SIAS Analysis Setpoint psia 1556 SGIS Analysis Setpoint psia 548 9

r v

TABLE 3-2 l

SEQUENCE OF EVENTS FOR THE _ EXCESS LOAD EVENT ilITHOUT AUXILIARY FEEDWATER

\\

Time (sec)

Event Setpoint or Value l

0.0 Steam Dump and Bypass Valves Fully Open 24.2 Pressurizer Empties 28.9 Pressurizer Pressure Trip Setpoirit 1728 psia Reached 29.8 Trip Breakers Open 30.3 CEAs Begin to Drop Into Core Feedwater Starts Rampdown 32.6 SIAS is Initiated

'. 1556 psia Reactor Coolant Pumps Manually Tripped 67.1 SGIS is Generated 548 psia

~

68.0 Main Steam Isolation Valves Begin to Close 80.0 Main Steam Isolation Valves are Closed

'90.3 Feedwater Rampdown to 5% is Completed c

107.6 Maximum RVUH Void is Reached 63%

i j1 210.3*

Main Feedwater Isolated 521.6 Miniaun RCS Pressure 728.50 717.8 Upper Head Void is Zero Pressurizer Starts to Refill

• Main feedwater would have been isolated 80 seconds after SGIS is initiated (i.e.,

at 147.1 seconds). The analysis conservatively assumed that main feedwater is isolated at 210.3 seconds to prolong the duration of RVUH voiding.

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TABLE 3-3 SEQUENCE OF EVENTS FOR THE EXCESS LOAD EVENT WITH AUXILIARY FEEDWATER Time-(sec)

Event Setpoint or Value 0.0 Steam Dump and Bypass Valves Fully Open 24.2 Pressurizer Empties 28.9 Low Pressurizer Pressure Trip Setpoint 1728 psia is Reached 29.8 Trip Breakers Open 30.3 CEAs Begin to Drop Into Core Feedwater Starts Rampdown Turbine Valves Begin to Close 32.6 SIAS is Initiated 1556 psia i

Reactor Coolant Pumps Manually Tripped 67.1 SGIS is Generated 548 psia 68.0 Main Steam Isolation Valves Begin to Close 80.0 Main Steam Isolation Valves are Closed 90.3 Feedwater Rampdown to 5% is Completed 107.6 Maximum RVUH Void is Reached 63%

210.3

• Main Feedwater Isolated 172 lbm/S.G.

Auxiliary Feedwater is Initiated 900.0 Operator Action To Isolate Auxiliary Feedwater to-Steam Generators

• Main Feedwater would have been isolated 80 seconds after SGIS is initiated (i.e.,

at 147.1 seconds). The analysis conservatively assumed that main feedwater is isolated at 210.3 seconds to prolong the duration of RVUH voiding.

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EXCESS LOAD EVENT WITH AUXILIARY FEEDWATER FIGURE calvert cliffs REACTOR CO0l>JIT SYSTEM TEMPERATURES VS TIME 3-11 Nuclear Power P.lant

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EXCESS LOAD EVENT WITH AUXILIARY FEEDWATER FIGURE calvert cliffs STEAM GENERATOR PRESSURE VS TIME 3-12 Nuclear Power Plant

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EXCESS LOAD EVENT WITH AUXILIARY FEEDWATER FIGURE calvert cliffs CLOSURE HEAD VOID FRACTION VS TIME

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Steam Generator Tube Rupture Event j

The Steam Generator Tube Rupture (SGTR) event was analyzed te verify that the site boundary doses will not exceed the guidelines of 10CFR100 for the 3

event initiatad from a pressurizer level of 975 ft.

The analysis included the effects of manually tripping the Reactor Coolant Pumps on SIAS due to low pressurtzer pressure.

The design basis ! IR is a double ended break of one steam generatur U-tube.

Table 4-1 lists the key transient related parameters used in this analysis.

In the analysis, it is assumed that the initial RCS pressure is as high as 2300 psia. This initial RCS pressure maximizes the amount of primary coolant transported to the secondary steam system since the leak rate is directly proportional to the difference between the primary and secondary pressure.

In addition, the higher pressure delays the lower pressurizer pressure trip which prolongs the transient and therefore maximizes the total primary to secondary mast and activities transported.

For this event, the acceptable DNBR limit is not exceeded due to tne action of the Thermal Margin / Low Pressure (TM/LP) trip which provides a reactor i

trip to maintain the DNBR above 1.23.

Since the SGTR event does not signifi-cantly affect the core power distribution, the PLHGR SAFDL is not approached.

The Thermal Margin / Low Pressure trip, with conservative coefficients which account for the limiting radial and axial peaks, maximum inlet temperature, RCS pressure, core power, and conservative CEA scram characteristics, would be the primary RPS trip intervening during the course of the transient.

However, to maximize the coolant transported from the primary to the secondary and thus the radioactive steam releases to the atmosphere, the analysis was perfonned assuning the reactor does not trip until the minimum setpoint (floor) of the Thennal Margin / Low Pressure trip is reached. This prolongs the steam releases to the atmosphere and thus maximizes the site boundary doses.

Tne Steam Generator Tube Rupture was analyzed assuming a manual trip of reactor coolant pumps on Safety Injection Actuation Signal (SIAS).

The Stear Generator Tube Rupture (SGTR) with RCS trip on SIAS results in higher site boundary doses because: (1) RCP coastdown increases pressure difference between the primary and the secendary, which increases the leak rate, and (2) RCP coastdown decreases the rate of decay heat removal,which increases the steam flow through the atmospheric dump valves.

l The Sequence of Events for the SGTR event with manual trip of RPC on SIAS l

1s presented in Table 4-2.

Figures 4-1 through 4-5 present the transient behavior of core power, heat flux, RCS pressure, RCS temperatures', and steam generator pressure.

I-131 activity release is based on the primary to secondary leak and on the steam flow required to reach cold shutdown conditions. This release is calculated as the product of steam flow, the time dependent steam activity and the decontamination factors applicable to each release pathway.

The O to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> I-131 site boundary does, is calculated from:

I 31 I 31*BRxgxCF DOSE (REM)

A

=

c

I where:

A -131, g,.31 activity released I

BR

= breathing rate x/Q

= dispersion coefficient CF -131 = I-131 dose conversion factor I

In detemining the whole body dose, the major assumption made is that all noble gases leaked through the ruptured tube will be released to the atmo-sphere. Therefore, the whole body dose is propsrtional to the total primary to secondary leak and is calculated using the following equation.

Whole Body Dose = [.25 (T + $ E,)]

• k

• L*A RCS where:

f, ' = average energy release _ by gama decay E

= average energy release by beta decay g

L

= total primary to secondary mass transport ARCS = noble gas activity of primary coolant x/Q = dispersion coefficient The results of the analysis are that 85616 lbs. of primary cool nt are transported to the steam generator secondary side. Based on this mass transport and values in Table 4-3, the site boundary doses calculated are:

Thyroid (DEQ I-131)

= 0.34 REM I

Whole Body (DEQ Xe-133) = 0.18 REM The reactor protective system (TM/LP) is adequate to protect the core from exceeding the DNBR limit. The doses resulting from the activity released as a consequence of a double-ended impture of one steam generator tube, assuming the maximum allowable Tech Spec activity for the primary concen-tration at a core power of 2754 MWt, are significantly below the guidelines of 10CFR100.

C

. ],.' s TABLE 4-1 KEY PARAMETERS ASSUMED IN THE STEAM GENERATOR TUBE RUPTURE EVENT KEY TRANSIENT RELATED PARAMETERS:

Parameter Units Value Power MWt 2754 MTC x10~4ao/*F

-2.5 Doppler Coefficient Multiplier 1.15 Scram Worth

%ao

-4.7 T

'F 550 in RCS Pressure psia 2300 0

Initial Core Mass Flow Rate x10 lb/hr 133.9 Initial Secondary Pressure psia 810 Tube ID inches

.654 Flow Constant 1.17 ASI(forscram)

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TABLE 4-2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TUBE RUPTURE EVENT WITH RCP COASTDOWN ON SIAS Time (sec)

Eventi, Setpoint or Value 0.0 Tube Rupture Occurs 786.5 Low Pressurizer Pressure Trip 1728 psia Signal Generated 787.6 Dump Valves Open 787.9 CEAs Begin to Drop Into Core 791.3 Pressurizer Empties 792.6 Safety Injection Actuation Signal 1556 psia Generated, RCP's Manually Tripped 795.5 Bypass Valves Open 797.1 Maximum Steam Generator Pressure 906 psia 854.0 Minimum RCS Presst 1064 psia

, 1800.0 Operator Isolates Damaged Steam Generator and Begins Cooldown to 300*F 12797.0 Operator Initiates Shutdown Cooling (TAV=300*F)

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TABLE 4-3 ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR THE STEAM GENERATOR TUBE RUPTURE Parameter _

Units Value Reactor Coolant System Maximus uC1/gm 1.0 Allowable Concentration (DEQ I-131)j Steam Generator Maximum A]lowable uC1/gm

.1 Concentration (DEQ I-131)

Reactor Coolant System Maximum Allowable pC1/gm

'100/E ConcentrationofNobleGases(DEQXe-133)j

.1 Steam Generator Partition Factor

.0005 Air Ejector Partition Factor Atmospheric Dispersion Coefficient sec/M 1.80x10-4 2

3 Breathing Rate M /sec 3.47x10-4 3

Dose Conversion Factor (I-131)

REM /C1 1.48x10-0 O

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STEAM GENEPATOR TUBE FAILURE EVENT FIGURE calvert cliffs CORE POWER VS TIME 4-1 Nuclear Power Plant s

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STEAM GENEPATOR TUBE FAILURE EVENT FIGURE

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STEAM GENERATOR TUBE FAILURE EVENT FIGURE nuc$Erlo0Niant REACTOR COOLANT SYSTEM PRESSURE VS TINE

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