ML18101B378

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Non-proprietary Fuel Upgrade & Margin Recovery Program:Loca Containment Integrity Analysis for Salem Nuclear Generating Station Units 1 & 2.
ML18101B378
Person / Time
Site: Salem  PSEG icon.png
Issue date: 08/31/1993
From: Kolano J, Spryshak J
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML18101B376 List:
References
WCAP-13839, NUDOCS 9605220065
Download: ML18101B378 (51)
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WESTINGHOUSE CLASS 3 WCAP-13839 SALEM NUCLEAR GENERATING STATION UNITS 1 & 2 FUEL UPGRADE AND MARGIN RECOVERY PROGRAM:

LOCA CONTAINMENT INTEGRITY ANALYSIS by J. J. Spryshak J. A. Kolano August 1993 J. A. Gresham, Manager Containment and Radiological Analyses Westinghouse Electric Corporation Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230.

~1993 Westinghouse Electric Corporation All Rights Reserved

~--,

9605220065 960325 PDR ADOCK 05000272 p PDR

TABLE OF CONTENTS SECTION LIST OF TABLES iii LIST OF FIGURES vi

1.0 INTRODUCTION

1-1 1.1* Purpose of analyses 1-1 1.2 Major analytical assumptions 1-2 2.0 LONG TERM LOCA MASS AND ENERGY RELEASE ANALYSIS 2-1 2.1 Introduction 2-1 2.2 LOCA mass and energy release phases 2-1 2.3 Break size and location 2-2 2.4 Application of single failure criteria 2-4 2.5 Mass and Energy release data 2-5 2.5.1 Slowdown mass and energy release data 2-5 2.5.2 Reflood mass and energy release data 2-5 2.5.3 Post-Reflood mass and energy release data 2-7 2.6 Sources of mass and energy 2-8 2.7 Significant modeling assumptions 2-10 3.0 LOCA.CONTAINMENT INTEGRITY (PEAK PRESSURE) ANALYSIS 3-1 3.1 Description of COCO model 3-1 3.2 Containment pressure calculation 3-2 3.3 Heat removal systems 3-3 3.4 Analysis results 3-9 3.5 Relevant acceptance criteria 3-10 00811: 1D/082793 i

TABLE OF CONTENTS (Cont'd)

SECTION PAGE

4.0 CONCLUSION

S 4-1

5.0 REFERENCES

5-1 00811: 1 D/082793 ii

,. LIST OF TABLES TABLE TITLE PAGE 2-1 Slowdown Mass and Energy Releases - DEPS 2-11 2-2 Slowdown Mass and Energy Releases - DEHL 2-12 2-3 Reflood Mass and Energy Releases - DEPS-MIN SI 2-13 2-4 Reflood Principal Parameters - DEPS-MIN SI 2-14 2-5 Post-Reflood Mass and Energy Releases - DEPS-MIN SI 2-15 2-6 Mass* Balance - DEPS-MIN SI 2-16 2-7 Energy Balance - DEPS-MIN SI 2-17 2-8 Salem Units 1 & 2 Specific Decay Heat Curve 2-18 2-9 System Parameters 2-19 3-1 Containment Structural Heat Sinks 3-15 3-2 Chronology of Events for LOCA - DEPS-MIN SI 3-17

  • 3-3 3-4 4-1 Chronology of Events for LOCA - DEHL Containment Analysis Parameters Containment Peak Pressure 3-18 3-19 4-2
  • 0081 1; 1D/082793 iii

LIST OF FIGURES FIGURE PAGE 3-1 Containment Pressure vs. Time - DEPS-MIN SI 3-11 3-2 Containment Pressure vs. Time - DEHL 3-12 3-3 Steam Temperature vs. Time - DEPS - MIN SI 3-13 3-4 Steam Temperatures vs. Time - DEHL 3-14 00811: 1D/082793 iv

1.0 INTRODUCTION

1.1 PURPOSE OF ANALYSES The purpose of this analysis was to calculate the Salem long term Loss of Coolant Accident (LOCA) mass and energy releases and resulting containment response for the double-ended pump suction (DEPS) and double-ended hot leg (DEHL) break cases using the proposed fuel upgrade conditions. This effort was performed concurrent with the Salem Fuel Upgrade and Margin Recovery Program and incorporated revised fan cooler heat removal rates, increased safeguard delays, and decreased safety injection flowrates.

This report provides the analytical basis with respect to the LOCA containment response for operation of the Salem Nuclear Power Plant Units 1 and 2 at the described conditions.

The analyses presented address the consequences of the mass and energy that

  • is released to containment as a result of a design basis LOCA. The mass and energy release data is subsequently used to verify, via calculations, that the containment design pressure is not exceeded in the event of a LOCA. In this manner, the analysis results demonstrate the acceptability of operation by showing that the containment peak pressure resulting from a design basis large break LOCA event will not exceed the design pressure. Section 2.0 presents the long term mass and energy release analysis. Section 3.0 presents the results of the containment integrity response calculations following a postulated LOCA.

Bounding initial temperatures and pressures for the containment integrity analyses were selected to envelop the limiting conditions for the proposed fuel upgrade. In this manner, the most limiting conditions for operation at full power (3411 MWt) were conservatively chosen.

1.2 MAJOR ANALYTICAL ASSUMPTIONS The evaluation model used for the long term LOCA mass and energy release calculations was the March 1979 model described in Reference 1. This D0811:1D/082793 1-1

evaluation model has been reviewed and approved by the NRC, and has been used in the analysis of other dry containment plants.

For the long term mass and energy release calculations, operating temperatures for the highest average coolant temperature case were selected as the bounding analysis conditions. The modeled power level of 3411 MWt (NSSS) adjusted for calorimetric error (+2 percent of power) was the basis in the analysis. The use of higher temperatures is conservative because the initial fluid energy is based on coolant temperatures which are at the maximum levels attained in steady state operation, Additionally, an allowance of +5.0 °F is reflected in the temperatures in order to account for instrument error and deadband. The initial RCS pressure in this analysis is based on a nominal value of 2250 psia. Also included is an allowance of

+50psi, which accounts for the measurement uncertainty on pressurizer pressure. The selection of 2250 psia as the limiting pressure is considered to affect the blowdown phase results only, since this represents the initial pressure of the RCS. The RCS rapidly depressurizes from this value until the

  • point at which it equilibrates with containment pressure.

The rate at which the RCS blows down is initially more severe at the higher RCS pressure (2250 psia). Additionally the RCS has a higher fluid density at 2250 psia (assuming a constant temperature) and subsequently has a higher RCS mass available for release. Thus, 2250 psia initial pressure was selected as the limiting case for the long term mass and energy release calculations.

These assumptions conservatively maximize the mass and energy in the RCS.

The selection of fuel allowance for the long term mass and energy calculation and subsequent LOCA containment integrity calculation is based on the need to conservatively maximize the core stored energy. The margin in core stored energy was chosen to be +15 percent. Thus, the analysis very conservatively accounts for the stored energy in the core. The fuel conditions were adjusted to provide a bounding analysis for the upgrade to Vantage Plus fuel with Performance Plus features.

Margin in RCS volume of 3% (which is composed of 1.6% allowance for thermal expansion and 1.4% for uncertainty) is modeled.

00811: 10/082793 1-2

  • Regarding safety injection flow, the mass and energy calculation considered only minimum safety injection flowrates. The case of minimum safety injection flow resulted in the limiting case for Salem Units 1 and 2 in Reference 11. In addition to a revised minimum safety injection assumption, the delay times for all of the safeguards equipment were increased when compared to the initial analysis in Ref. 11. Further details about these assumptions are contained in sections 2.4 and 3.2.

Thus, based on the above conditions and assumptions, a bounding analysis of Salem Units 1 and 2 is made for the release of mass and energy from the RCS in the event of a LOCA at the fuel upgrade conditions.

The COCO code has been used and found acceptable to calculate containment pressure transients for dry containment plants. This code has been successfully used for the other dry containment plants in their FSAR analyses *

  • As input to the COCO computer code, mass and energy release rates as described in Section 2 of this report will be used. Other major analysis assumptions will be that one diesel train will be assumed to fail, consistent with the requirements to analyze the worst single failure *
  • 00811: 1D/082793 1-3

2.0 LONG TERM LOCA MASS AND ENERGY RELEASE ANALYSIS

2.1 INTRODUCTION

This report section presents the long term LOCA mass and energy releases that were generated in support of the fuel upgrade and margin recovery program analysis effort for Salem Units 1 and 2. These mass and energy releases are then subsequently used in the COCO containment integrity analysis peak pressure calculation.

2.2 LOCA MASS .AND ENERGY RELEASE .PHASES The LOCA transient is typically divided into four phases:

1. Blowdown - which includes the period from accident initiation (when the reactor is at steady state operation) to the time that the RCS reaches initial equilibration with containment.
2. Refill - the period of'time when the lower plenum is being filled by accumulator and safety injection water. At the end of blowdown, a large amount of water remains in the cold legs, downcomer, and lower plenum. To conservatively consider the refill period for the purpose of containment mass and energy releases, this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum.

This allows an uninterrupted release of mass and energy to containment. Thus, the refill period is conservatively neglected in the mass and energy release calculation.

3. Reflood - begins when the water from the lower plenum enters the core and ends when the core is completely quenched.
4. Post-Reflood (Froth) - describes the period following the reflood transient. For the pump suction break, a two-phase mixture exits the core, passes through the hot legs, and is superheated in the 00811: 1D/082793 2-1

steam generators. After the broken loop steam generator cools, the break flow becomes two phase.

  • 2.3 BREAK SIZE AND LOCATION Generic studies have been performed with respect to the effect on the LOCA mass and energy releases relative to postulated break size. The double ended guillotine break has been found to be limiting due to larger mass flow rates during the blowdown phase of the transient. During the reflood and froth phases, the break size has little effect on the releases.

Three distinct locations in the reactor coolant system loop can be postulated for pipe rupture:

1. Hot leg (between vessel and steam generator)
2. Cold leg (between pump and vessel)
3. Pump suction (between steam generator and pump)

The break locations analyzed for this program are the double-ended pump suction guillotine break (10.48 ft2) and the double-end hot leg guillotine break (9.12 ft2). Pump suction break mass and energy releases have been calculated for the blowdown, reflood, and post-reflood phases of the LOCA and the hot leg break mass and energy releases have been calculated for only the blowdown phase. The following information provides a discussion on each break location.

The double ended hot leg guillotine has been shown in previous studies to result in the highest blowdown mass and energy release rates. Although the core flooding rate would be highest for this break location, the amount of energy released from the steam generator secondary is minimal because the majority of the fluid which exits the core bypasses the steam generators in venting to containment. As a result, the reflood mass and energy releases are reduced significantly as compared to either the pump suction or cold leg break locations where the core exit mixture must pass through the steam generators before venting through the break *

  • D0811: lD/082793 2-2

For the hot leg break, there is no reflood peak as determined by generic studies (i.e., from the end of the blowdown period the releases would continually decrease). Therefore the reflood (and subsequent post-reflood) releases are not calculated for a hot leg break. The mass and energy releases for the hot leg break have been included in the scope of this containment integrity analysis because the blowdown phase of a hot leg break has been shown to result in a limiting condition for some dry containment plants.

The cold leg break location has also been found in previous studies to be much less limiting in terms of the overall containment peak pressure. The cold leg blowdown is faster than that of the pump suction break, and more mass is released into the containment. However, the core heat transfer is greatly reduced, and this results in a considerably lower energy release into containment. Studies have determined that the blowdown transient is, in general, less limiting than the pump suction break. During reflood, the flooding rate is greatly reduced and the_energy release rate into the containment is reduced. Therefore, the cold leg break is not included in the scope of this analysis.

The pump suction break combines the effects of the relatively high core flooding rate, as in the hot leg break, and the addition of the stored energy in the steam generators. As a result, the pump suction break yields the highest energy flow rates during the post-blowdown period by including all of the available energy of the Reactor Coolant System in calculating the releases to containment. The analysis of this break location for Salem Units 1 and 2 as the limiting break for long term containment integrity is consistent with other dry containment plants.

In summary, the analysis of the limiting break locations for a dry containment have been performed and are shown in this report. The double-ended pump suction guillotine break has historically been considered to be the limiting break location for the post blowdown phase of the event, by virtue of its consideration of all energy sources present in the RCS. The analyses presented in this document are based on the results of Reference 11, which supports the conclusions of the DEPS as the limiting break case for the 00811: 1D/082793 2-3

post blowdown period, considering both the minimum and maximum safety injection cases. The double-ended pump suction break location provides a mechanism for the release of the available energy in the Reactor Coolant System, including both the broken and intact loop steam generators. See Section 3.4 for the limiting break case for this analysis.

2.4 APPLICATION OF SINGLE FAILURE CRITERIA An analysis of the effects of the single failure criteria has been performed on the mass and energy release rates for the double ended pump suction (DEPS) break. For the DEPS results presented in this report, an inherent assumption in the generation of the mass and energy releases is that offsite power is lost. This results in the actuation of the emergency diesel generators, required to power the safety injection system. This is not an issue for the blowdown period which is limited by the DEHL break.

Two cases have been analyzed previously for the effects of a single failure

[Reference 11]. The double ended pump suction case with both minimum and maximum safety injection for the conditions was analyzed. The limiting case for the Salem Units 1 and 2 is the minimum safeguards case. This was determined by prior generic and specific Salem analyses. In the case of minimum safeguards, the single failure postulated to occur is the loss of an emergency diesel generator. This results in the loss of one pumped safety injection train and the containment safeguards components on that diesel, thereby minimizing the safety injection flow. The analysis further considers the safety injection pump head curves to be degraded by at least 5%. This results in the greatest possible reduction for the Emergency Core Cooling System (ECCS) components.

2.5 MASS AND ENERGY RELEASE DATA 2.5.l Blowdown Mass and Energy Release Data The SATAN-VI code is used for computing the blowdown transient, and is the same as that used for the Emergency Core Cooling System (ECCS) calculation in 00811:10/082793 2-4

Reference 2. The methodology for the use of this model is described in Reference 1.

Table 2-1 presents the calculated mass and energy release for the blowdown phase of the DEPS break. The mass and energy release for the double-ended pump suction break, given in Table 2-1, terminate 27.0 seconds after the initiation of the postulated accident.

Table 2-2 presents the calculated mass and energy release for the blowdown phase of the DEHL break. The mass and energy releases for the DEHL break terminate 31.0 seconds after initiation of the postulated accident.

2.5.2 Reflood Mass and Energy Release Data The WREFLOOD code is used for computing the reflood transient, and is a modified version of that used in the ECCS calculation in Reference 2. The methodology for the use of this model is described in Reference 1.

An exception to the mass and energy evaluation model described in Reference 1 is taken, in that steam/water mixing in the broken loop has been included in this analysis. This assumption is justified and is supported by test data, and is summarized as follows:

The model assumes a complete mixing condition (i.e., thermal equilibrium) for the steam/water interaction. The complete mixing process, however, is made up of two distinct physical processes. The first is a two phase interaction with condensation of steam by cold injection water. The second is a single phase mixing of condensate and injection water. Since the steam release is the most important influence to the containment pressure transient, the steam condensation part of the mixing process is the only part that need be considered. (Any spillage directly heats only the sump.)

The most applicable steam/water mixing test data has been reviewed for validation of the containment integrity reflood steam/water mixing model.

This data is that generated in 1/3 scale tests (Reference 3), which are the largest scale data available and thus most closely simulates the flow regimes 00811: 1D/082793 2-5

and gravitational effects that would occur in a PWR. These tests were designed specifically to study the steam/water interaction for PWR reflood conditions.

From the entire series of 1/3 scale tests, a group corresponds almost directly to containment integrity reflood conditions. The injection flowrates for this group cover all phases and mixing conditions calculated during the reflood transient. The data from these tests were reviewed and discussed in detail in Reference 1. For all of these tests, the data clearly indicates the occurrence of very effective mixing with rapid steam condensation. *The mixing model used in the containment integrity reflood calculation is therefore wholly supported by the 1/3 scale steam/water mixing data.

Additionally, the following justification is also noted. The limiting break for the containment integrity peak pressure ~nalysis is the double ended pump suction break. For this break, there are two flowpaths available in the RCS by which mass and energy may be released to containment. One is through the outlet of the steam generator, the other via reverse flow through the reactor coolant pump. Steam which is not condensed by ECCS injection in the intact RCS loops passes around the downcomer and through the broken loop cold leg and pump in venting to containment. This steam also encounters ECCS injection water as it passes through the broken loop cold leg, complete mixing occurs and a portion of it is condensed. It is this portion of steam which is condensed that is taken credit for in this analysis. This assumption is justified based upon the postulated break location, and the actual physical presence of the ECCS injection nozzle. A description of the test and test results is contained in References 1 and 3.

The methodology previously discussed and described in Reference 1 has been utilized and approved on the Dockets for Catawba Units 1 & 2, McGuire Units 1

& 2, Sequoyah Units 1 & 2, Watts Bar Units 1 & 2, Millstone Unit 3 and Beaver Valley Unit 2.

Table 2-3 presents the calculated mass and energy release for the reflood phase of the double-ended pump suction break with minimum safety injection.

00811:10/082793 2-6

A significant discharge occurs during the period when the accumulators are injecting (from 34.7 to 63.1 seconds for the minimum safety injection case as illustrated in Table 2-3).

The transient of the principal parameters during reflood* are given in Tables 2-4 for the minimum safety injection double-ended pump suction break case.

2.5.3 Post-Reflood Mass and Energy Release Data The FROTH code* [Reference 4] is used for computing the post-reflood transient. The methodology for the use of this model is described in Reference 1. The mass and energy release rates calculated by FROTH are used in the containment analysis to the time of containment depressurization.

After depressurization, the mass and energy release from decay heat is based on the 1979 ANSI/ANS Standard, shown in Reference 5, and the following input:

  • 1. Decay heat sources considered are fission product decay and heavy element decay of U-239 and Np-239.
2. Decay heat power from fissioning isotopes other than U-235 is assumed to be identical to that of U-235.
3. Fission rate is constant over the operating history of maximum power level.
4. The factor accounting for neutron capture in fission products has been taken from Table 10 of ANS (1979).
5. Operation time before shutdown is 3 years.
6. The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.

00811: 1D/082793 2-7

  • 7. Two sigma uncertainty (2 times the standard deviation) has been applied to the fission product decay.

Table 2-5 presents the two phase (froth) mass and energy release data for the double-ended pump suction break with minimum safety injection.

2.6 SOURCES OF MASS AND ENERGY The sources of mass considered in the DEPS LOCA mass and energy release analysis are given in Table 2-6. These sources are the reactor coolant system, accumul~tors, and pumped safety injection.

The energy inventories considered in the DEPS LOCA mass and energy release analysis are given in Table 2-7. The energy sources include:

1. Reactor Coolant System Water
  • 2.

3.

Accumulator Water Pumped Injection Water

4. Decay Heat
5. Core Stored Energy
6. Reactor Coolant System Metal
7. Steam Generator Metal
8. Steam Generator Secondary Energy
9. Secondary Transfer of Energy {feedwater into and steam out of the steam generator secondary)

In the mass and energy release data presented, no Zirc-water reaction heat 00811: 1D/082793 2-8

was considered because the clad temperature did not rise high enough for the rate of the Zirc-water reaction heat to be of any significance.

System parameters needed to perform confirmatory analyses are provided in Table 2-9.

The consideration of the various energy sources in the mass and energy release analysis provides assurance that all available sources of energy have been included in this analysis. Thus the review guidelines presented in Standard Review Plan section 6.2.1.3 have been satisfied.

The mass and energy inventories are presented at the following times, as appropriate:

1. Time zero (initial conditions)
2. End of blowdown time
3. End of refill time
4. End of reflood time
5. Time of full depressurization
6. End of analysis The methods and assumptions used to release the various energy sources are given in Reference 1, except as noted in section 2.5.2 of this document, which has been approved as a valid evaluation model by the Nuclear Regulatory Commission *
  • 00811:10/082793 2-9

2.7 SIGNIFICANT MODELING ASSUMPTIONS The following items ensure that the mass and energy releases are conservatively calculated for maximum containment pressure:

1. A maximum expected operating temperature of the reactor coolant system;
2. An allowance in temperature for instrument error and dead band

{+5°F);

3. A margin in volume of +3% (composed of 1.6% allowance for thermal expansion, and 1.4% for uncertainty);
4. A power level of 3411 MWt;
5. An allowance for calorimetric error (+2 percent of the license power);
6. Conservatively modified coefficients of heat transfer;
7. An allowance in core stored energy for effect of fuel densification;
8. A margin in core stored energy (typically +15 percent included to account for manufacturing tolerances);
9. An allowance for RCS pressure uncertainty (+50 psi); and
10. A maximum containment backpressure based upon design pressure *
  • 00811: 1D/082793 2-10
  • TABLE 2-1 SLOWDOWN MASS AND ENERGY RELEASES DEPS 3411 MWt TIME BREAK PATH N0.1 FLOW BREAK PATH N0.2 FLOW THOUSAND THOUSAND SECONDS LBM/SEC BTU/SEC LBM/SEC BTU/SEC 0.0000 o.o o.o o.o o.o 0.0503 40094.7 21598.6 23752.8 12729.7 0.100 40103.8 21641.6 21395.3 11508.9 0.251 43487.8 23691.6 23141. 1 12462.8 0.401 44962.0 24827.8 22622.8 12205.9 0.901 42863.5 24927.0 19361.5 10466.5 1.40 38265.6 23113.4 18357.7 9927.4 2.20 31932.9 20490.0 17960.4 9711.8 2.80 25686.9 17237 .5 16900.1 9141.8 3.00 20753.8 14067.3 16534.1 8946. 1 3.20 20355.0 13945.2 16125. 1 8727 .1 4.00 16222.6 11179.0 14728.2 7980.7 4.60 14040.8 9691.6 13946.8 7564.5 5.25 12679.9 8739.9 13271.2 7205. 7 6.25 11660.9 7950.3 12374.5 6729.8 6.75 11484.4 7762.3 13260.4 7220.7 8.25 12180.2 7948.8 12294.6 6717.3 8.50 11772.6 7720.5 12246.3 6694. 1 9.00 9339.9 6868.5 12017. 7 6570.9 9.75 8564.3 6483.0 11628.9 6361.5 12.0 8372.3 5893.4 10406.2 5691.5 17.3 5016.6 4262.0 7753.2 4259.3 18.5 4327.5 3923.2 6806.7 3662.8 19.0 4026.3 3748.1 7431.6 3815.3 19.3 3888.6 3668.2 6410.6 3205.4 19.5 3759.8 3610.9 9730.7 5008.7 19.8 3639.6 3599.8 5566.4 2834.2 20.3 3321.9 3540.2 6109.6 2922.0 20.8 2780.5 3264.7 5038.1 2301.9 22.0 1810.8 2246.7 4031.5 1609 .4 22.5 1517.0 1891. 1 5135.5 2006.1 23.3 1005.8 1261 .6 2238.0 873.3 23.5 839.9 1055.9 1664.6 653.3 25.3 296.4 375.0 917.6 308.2 27.0 0.0 0.0 o.o o.o P5""~:Z."I
  • 00811: 1D/082493 2-11
  • TABLE 2-2 SLOWDOWN MASS AND ENERGY RELEASES DEHL 3411 MWt TIME BREAK PATH N0.1 FLOW BREAK PATH N0.2 FLOW THOUSAND THOUSAND SECONDS LBM/SEC BTU/SEC LBM/SEC BTU/SEC 0.0000 o.o o.o o.o 0.0 0.0502 53889.4 34408.8 26825.7 16997. 1 0.100 45536.9 28941.8 26956.9 17081.0 0.200 34665.8 22293.5 22967.7 14484.3 0.351 32957.2 21151.2 20046.6 12400.5

. 0.650 32411.5 20843.6 17926.5 10587.3 1.20 30349.2 19938.5 16433.1 9228.7 1.80 27980.2 18835.8 16425.9 8962.0 2.50 25020.6 17080.2 16946.3 9088.4 2.90 23578.3 16142.2 17057.4 9113. 1 3.60 21582.5 14723.5 16873.6 9001.7 4.50 19997.8 13413.4 16280.6 8708.0 5.50 19306.6 12602.0 15020.5 8089.4 6.75 19862.2 12611.1 12887.6 7029.5 7.00 14624.6 10225 .1 12433.2 6798.8 7.25 15414.1 10581. 7 11969.9 6561.8 8.50 15772.6 10483.3 10032.2 5559.0 9.25 15726.1 10346.5 9112 .1 5079.3 9.50 15414.9 10169. 1 8842.2 4938.3

~0.5 15188.5 9877.7 7854.0 4423.3 11.8 14237. 7 9158.1 6753.8 3858.7 12.8 B061.:. 8396.8 5944.4 3454.6 14.8 10539.1 6946.0 4611.6 2816.7 16.5 7989.3 5732.7 3577. 7 2342.2 18.0 4921.2 4605.6 2481.0 1820.9 18.5 3581.1 3783.1 2320.3 1713.8 19.0 2747.5 3112.2 2197.2 1622.8 20.0 1842.7 2182.7 1770.2 1473 .1 21.3 1130.3 1397 .1 743.0 917 .1 23.5 452.8 581.2 256.8 322.4 26.3 158.3 206.4 82.7 105.7 28.0 466.8 582.S 163.S 207.4 28.8 349.3 446.7 97.5 124.4 30.S 373.3 469.1 108.7 139.0 31.0 o.o o.o o.o 0.0

\.\LE.1111 00811: 1D/082493 2-12

TABLE 2-3 REFLOOO MASS AND ENERGY RELEASES DEPS - MIN SI TIME BREAK PATH N0.1 FLOW BREAK PATH N0.2 FLOW THOUSAND THOUSAND SECONDS LBM/SEC BTU/SEC LBM/SEC BTU/SEC 27.0 o.o 0.0 0.0 0.0 28.0 1.2 1.4 o.o 0.0 28.8 41.3 48.7 o.o o.o

31. 1 101.1 119.2 0.0 0.0 34.1 148.5 175. 1 0.0 o.o 34.7 162.4 191.5 433.2 60.0 35.1 294.5 348.1 2738.9 395.3 36.1 419.5 497.0 4122.6 629.2 37.1 449.5 533.0 4475.7 658.6 38.1 449.4 532.7 4440.4 659.9 42.1 421.5 499.4 4178.5 628.4 44.1 408.9 484.4 4056.8 613.6 46.1 397.2 470.4 3942.1 599.6 48.1 386.3 457.4 3834.2 586.4
52. 1 366.8 434.1 3636.5 562.3 53.1 362.3 428.8 3590.5 556.7 55.1 353.7 418.6 3502.2 545.8
59. 1 338.2 400.0 3338.6 525.8 60.1 334.5 395.7 3300.2 521. 1 62.1 327.6 387.4 3225.9 512.0 63.1 274.0 323.7 2596.4 442.3
64. 1 423.0 501.2 319.0 226.2 67.1 404.8 479.5 310.8 215.3
71. 1 378.5 448. 1 299.2 199.8 74.1 361.4 427.6 291.6 189.8
81. 1 326.0 385.5 276.1 169.5
88. 1 296.3 350.2 263.4 152.8 92.1 281.8 333.0 257.1 144.8 106. 1 242.4 286.3 240.6 123.6 126. 1 210.3 248.2 227.S 106.9 138. 1 200.4 236.S 223.4 101 .8 162. 1 192.3 226.9 220.0 97.4 182. 1 191 .9 226.4 219.6 96.9 190. 1 195. 7 231 .o 226.3 99.2 198. 1 200.S 236.6 240.3 102.7

~s,.~; ;.:*\

  • 00689: 10/121291 2-13

TABLE 2-4 REFLOOD PRINCIPAL PARAMETERS DEPS - MIN SI TIME FLOODING CARRYOVER CORE DOWNCOJlllER FLOW INJECTION TEMP RATE FRACTION HEIGHT HEIGHT FRACTION TOTAL ACCUMULATOR SPILL ENTHALPY SECONDS DEGREE F IN/SEC FT FT (POUNDS JlllASS PER SECOND) BTU/LBJlll 27.0 249.0 0.000 o.ooo 0.00 0.00 0.250 o.o o.o o.o 0.00 27.7 245.8 21.024 0.000 0.50 1. 16 o.ooo 6843.3 6843.3 o.o 89.59 28.0 242.6 24.777 o.ooo 1. 10 1.23 o.ooo 6775.2 6775.2 o.o 89.59 28.3 241. 7 2.702 o. 104 1 .31 1.70 o. 199 6689.9 6689.9 0.0 o.o 89.59 28.4 241.6 2.821 0. 113 1.33 1.92 o. 194 6679. 1 6679.1 o.o 89.59 29.4 241.3 2.360 0.312 1.50 4.28 0.333 6443.2 6443.2 89.59

30. 1 241. 1 2.291 0.393 1.58 5.72 0.347 6319.5 6319.5 o.o 89.59 34.7 240.2 2.615 0.639 2.00 15.39 0.382 5580.6 5580.6 o.o 89.59 37.1 238.8 4.313 0.696 2.26 16.07 0.595 5269. 1 4737.3 o.o 87 .41
38. 1 238.3 4.178 0.710 2.36 16.07 0.592 5209.4 4672.7 o.o 87.37 39.6 237.6 4.024 0.723 2.51 16.07 0.589 5072.6 4533.3 o.o 87.29 45.5 235.8 3.638 0.745 3.00 16.07 0.578 4620.0 4071.7 0.0 87.03 52.4 235.0 3.356 0.753 3.50 16.07 0.566 4204.8 3648.4 o.o 86.73 60.0 235.2 3. 131 0.756 4.00 16.07 0.553 3837.6 3274.6 0.0 86.42
63. 1 235.5 2.764 0.755 4.20 16.07 0.514 3042.6 2467.9 o.o 85.51
64. 1 235.6 3.644 0.759 4.27 15.95 0.600 541.8 o.o o.o 68.00 68.1 236.3 3.441 0.759 4.55 15.38 0.596 547.2 o.o o.o 68.00 75.0 238.5 3. 120 0.759 5.00 14.60 0.589 556.3 o.o o.o 68.00
84. 1 242.8 2.783 0.759 5.54 13.88 0.579 565.0 o.o o.o 68.00 92.7 247.6 2.532 0.759 6.00 13.46 0.569 570.7 o.o o.o 68.00 104.1 254.4 2.282 0.760 6.55 13.20 0.558 575.7 o.o o.o 68.00 114.4 259.7 2.121 0.761 7.00 13.18 0.548 578.6 0.0 o.o 68.00 128. 1 265.6 1 .981 0.764 7.56 13.37 0.539 580.9 o.o o.o 68.00 139.8 269.9 1.907 0.767 8.00 13.65 0.534 582.1 'l.O o.o 68.00 154. 1 274.3 1.855 0.771 8.52 14.08 0.530 582.8 J.0 o.o 68.00 168.0 278. 1 1.828 0.776 9.00 14.56 0.529 583. 1 o.o o.o 68.00 182. 1 281.4 1 .816 0.781 9.47 15.06 0.529 583.1 o.o o.o 68.00 184. 1 281 .9 1.819 0.782 9.54 15.14 0.530 583.0 o.o o.o 68.00 198. 1 284.7 1.846 0.788 10.00 15.57 0.540 582.0 o.o o.o 68.00 f'51'1\4?l""I
  • 00689: 10/121291 2-14

TABLE 2-5

  • POST-REFLOOD MASS AND ENERGY RELEASES DEPS - MIN SI TIME BREAK PATH NO. 1 FLOW BREAK PATH NO. 2 FLOW THOUSAND THOUSAND SECONDS LBM/SEC BTU/SEC LBM/SEC BTU/SEC 198.2 215.0 266 .9 378.3 128.1 200*.2 214.3 266.0 379.0 127.B 258.2 208.8 259.2 384 .5 127.1 278.2 206.6 256 .5 386.7 126.B 283 .2 206.6 256.5 386.7 126.5 303.2 203.8 253.0 389.5 126.4 308.2 203.7 252.B 389.6 126.2 363.2 196.9 244.4 396.4 125.5 383.2 194.1 240.9 399.2 125.3 388.2 193.8 240.6 399.5 125.1 403 .2 191.5 237.7 401.B 125.0 408.2 191. 3 237.4 402.1 124.9 438.2 187.1 212 .2 406.2 124.6 443.2 186.7 231.8 406.6 124.5 463.2 183.9 228.3 409.4 124.3 678.2 183.9 228.3 409 .4 124.3 678.3 92.4 113.9 500.9 145.5 718.2 91.4 112.6 501.9 143 .5 798.2 89.3 110.0 504 .1 142 .5 823.2 88.7 109.3 504.6 141.0 908.2 86.9 107.1 506.4 141.9 912.2 86.B 107.0 506.5 141.5 1093.2 83.6 102.9 509.B 140.0 1098.2 83.5 102.B 509.B 139.5 1183.2 82.2 101.2 511.l 138.8 1188.2 82.1 101.1 511.2 138.0 1258.2 81.0 99.8 512.3 137.1 1293.2 80.5 99.1 512. B 137.4 1443.3 80.5 99.1 512.B 137.4 1443.4 77.5 89.2 515.B 44.5 1704.53 77.5000 89.20000 515.800 44.50000 1704. 63 82.1061 94.60475 322.504 63.65456 3599.00 82.1061 94.60475 322.504 63.65456 3600.00 61.6000 70.90000 343.000 67.70000 3600.10 51.1000 58.10000 353.600 53.00000 4473 .87 127.7600 54.05800 80.772 11.48053 10000.00 36.8300 42.37218 90.930 12.92434 100000.00 19.6900 22.65295 108.070 15.36053 1000000.00 8.4400 9.71005 119.320 16.95955 10000000.00 2.4940 2.86930 125.266 10.28684 00689: 10/121291 2-15

TABLE 2-6 MASS BALANCE - DEPS - MIN SI MASS BALANCE TIME (SECONDS> 0.00 27 .00 27.00 198. 13 6!!3.20 1443.28 3600.00 MASS <THOUSAND LBM)

INITIAL IN RCS AND ACC 753.93 753.93 753. 93 753.93 753.93 753.93 753.93 ADDED MASS PUMPED INJECTION 0.00 0.00 0.00 92.19 379.94 803.53 1676.16 TOTAL ADDED 0.00 0.00 0.00 92.19 379.94 803.53 1676.16

      • TOTAL AVAILABLE *** 753.93 753.93 753. 93 846.12 1133.87 1557.46 2430.09 DISTRIBUTION REACTOR COOLANT 532.30 52.26 67.46 136.22 136. 22 136.22 136.22 ACCUMULATOR 221.63 174.28 159.08 0.00 0.00 0.00 0.00 TOTAL CONTENTS 753.93 226.54 226.54 136.22 136.22 136.22 136. 22 EFFLUENT BREAK FLOW o.oo 527.38 527.38 709.89 997.65 1421.24 2293.87
  • ECCS SPILL TOTAL EFFLUENT
      • TOTAL ACCOUNTABLE ***

0.00 0.00 753.93 0.00 527.38 753.92 0.00 527.38 753.92 0.00 709.89 846.11 0.00 997.65 1133.87 0.00 1421.24 1557.45 2293.87 2430.09 0.00 PS.n..;~:..ai 00689: 10/121291 2-16

TABLE 2-7 ENERGY BALANCE - DEPS - MIN SI ENERGY BALANCE Til'IE (SECONDS) 0.00 27.00 27.00 198.13 683.20 1443.28 3600.00 ENERGY (MILLION BTU)

INITIAL ENERGY IN RCS,ACC,S GEN 885.15 885. 15 885.15 885 .15 885.15 885.15 885.15 ADDED ENERGY PUMPED INJECTION 0.00 0.00 0.00 6.27 25.84 59.45 190.35 DECAY HEAT 0.00 8.82 8.82 29.01 72.34 127 .57 251.24 HEAT FROl'I SECONDAR 0.00 -1 .49 -1.49 -1.49 2.97 8.63 8.63 TOTAL ADDED 0.00 7.33 7.33 33.79 101.15 195 .65 450.22

    • "' TOTAL AVAILABLE *** 885.15 892.49 892.49 918.95 986.30 1080.80 1335.37 DISTRIBUTION REACTOR COOLANT 311.62 14.74 16.10 37.01 37.01 37.01 37.01 ACCUMULATOR 19 .86 15 .61 14.25 0.00 0.00 0.00 0.00 CORE STORED 25.02 12.57 12 .57 4.85 4.64 4.30 3.33 PRIMARY l'IETAL 155.79 146.54 146.54 119 .59 85.70 67.30 52.33 SECONDARY l'IETAL 98.26 98.42 98.42 89.50 70.46 50.95 39.69 STEAi'! GENERATOR 274.61 281 .21 281.21 252.10 197.23 145.41 1, 4. 72 TOTAL CONTENTS 885.15 569.10 569.10 503.05 395. 10 304.96 247.08 EFFLUENT BREAK FLOW 0.00 323.40 323.40 407.95 583.25 767.89 1080.35 ECCS SPILL 0.00 0.00 0.00 0.00 o.oo 0.00 o.oo TOTAL EFFLUENT 0.00 323.40 323.40 407.95 583.25 767.89 1080.35
      • TOTAL ACCOUNTABLE *** 885.15 892.50 892.50 911.00 978.36 1072.85 1327.43
'SN'lo..i~'t..°'
  • 00689: 101121291 2-17

TABLE 2-8 3411 MWt SPECIFIC DECAY HEAT CURVE Time Energy Rate (sec) (BtuZsec) 1.00E+Ol 1. 741E+05 l.SOE+Ol 1.628E+05 2.00E+Ol 1. 551E+05 4.00E+Ol l.370E+05 6.00E+Ol l.268E+05 8.00E+Ol 1.198E+05 l.OOE+02 1.146E+05 l.50E+02 l.057E+05 2.00E+02 9.996E+04 4.00E+02 8.749E+04 6.00E+02 8.055E+04 8.00E+02 7.557E+04 l.OOE+03 7.158E+04 l.50E+03 6.436E+04 5.917E+04 2.00E+03 4.00E+03 4.775E+04 6.00E+03 4.213E+04 8.00E+03 3.877E+04 l.OOE+04 3.638E+04 1.50E+04 3.262E+04 4.00E+04 2.513E+04 1.00E+05 l.945E+04 4.00E+05 l.218E+04 6.00E+05 l.034E+04 8.00E+OS 9.156E+03 l.OOE+06 8.336E+03 l.50E+06 8.174E+03 2.00E+06 6.167E+03 4.00E+06 4.378E+03 6.00E+06 3.525E+03 8.00E+06 2.995E+03 l.OOE+07 2.610E+03 2.00E+07 l.541E+03 4.00E+07 8.691E+02 6.00E+07 6.074E+02 8.00E+07 4.523E+02 l.OOE+08 3.586E+02 D0811: 1D/082793 2-18

TABLE 2-9 SYSTEM PARAMETERS Parameter Core Inlet Temperature (+5.0°F) 548.2 °F Initial Steam Generator Steam Pressure 828.0 psig Assumed Maximum Containment Back Pressure 61. 7 psi a

  • 00811: 1D/082793 2-19
  • 3.0 CONTAINMENT INTEGRITY ANALYSIS

3.1 DESCRIPTION

OF COCO MODEL Calculation of containment pressure and temperature transients is accomplished by use of the digital computer code, COCO (Reference [6]). The COCO code has been used and found acceptable to calculate containment pressure transients for many dry containment plants. Transient phenomena within the reactor coolant system affect containment conditions by means of convective mass and energy transport through the pipe break.

For analytical rigor and convenience, the containment air-steam-water mixture is separated into systems. The first system consists of the air-steam phase; the second consists of the water phase. Sufficient relationships to describe the transient are provided by the equations of conservation of mass and energy as applied to each system, together with appropriate boundary conditions. As thermodynamic equations of state and conditions may vary

  • during the transient, the equations have been derived for all possible cases of superheated or saturated steam and subcooled or saturated water.

Switching between states is handled automatically by the code. The following are the major assumptions made in the analysis:

(a) Discharge mass and energy flow rates through the reactor coolant system breaks are established from the analysis in Section 2.

(b) For the blowdown portion of the LOCA analysis, the discharge flow separates into steam and water phases at the break point. The saturated water phase is at the total containment pressure, while the steam phase is at the partial pressure of the steam in the containment. For the post-blowdown portion of the LOCA analysis, steam and water releases are input separately.

(c) Homogeneous mixing is assumed. The steam-air mixture and the water phase each have uniform properties. More specifically, thermal equilibrium between the air and steam is assumed. This does not 00811:10/082793 3-1

imply thermal equilibrium between the steam-air mixture and water phase.

(d) Air is taken as an ideal gas, while compressed water and steam tables are employed for water and steam thermodynamic properties.

(e) The saturation temperature at the partial pressure of the steam is used for heat transfer to the heat sinks and the fan coolers.

3.2 CONTAINMENT PRESSURE CALCULATION The following are the major input assumptions used in the COCO analysis for the pump suction pipe rupture case with the steam generators considered as an active heat source for the Salem Units 1 and 2 Nuclear Plant Containment:

1. Minimum safeguards are employed in all calculations, e.g., one of two spray pumps and three of five fan coolers. Additionally used
  • are one of two RHR pumps, one of two recirculation pumps and one of two RHR heat exchangers providing flow to the core. [Due to the duration of the hot leg transient, no safeguards equipment is modeled.]
2. The blowdown, reflood, and post reflood mass and energy releases described in Section 2 are used.
3. A service water temperature of 95°F is used on the component cooling heat exchanger and the fan cooler units.
4. The service water flowrate for the fan cooler was 2500 gpm.
5. The initial conditions in the containment are a temperature of 120°F and a pressure of 15.0 psia.
6. Containment structural heat sinks are assumed with conservatively low heat transfer rates. (See Tables 3-1,3-2) 00811: 1D/082793 3-2
7. The operation of one RHR heat exchanger (UA = 1.447 x 106 Btu/hr-°F) for core cooling. The component cooling heat exchanger was modeled at UA = 3.72 x 106 Btu/hr-°F until the time that spray switchover would occur. After that time a UA = 1.0 x 106 Btu/hr-°F was used because of reduced sump water flow rates.
8. The service water flow to the component cooling heat exchanger was modeled as 8000 gpm.
9. The delay times for the operation of the fan cooler and containment spray.pumps were increased to 55 seconds and 85 seconds respectfully, incorporating an increase in the diesel start time to 20 seconds (Reference 12).
10. A reduction in fan cooler heat removal rates over previous analyses have been used, as listed in Table 3-4 *
  • 11. Additional reductions in SI pump flows have been used to incorporate anticipated plant conditions.

3.3 HEAT REMOVAL SYSTEMS The significant heat removal source during the early portion of the transient is the structural heat sinks. Provision is made in the containment pressure transient analysis for heat transfer through, and heat storage in, both interior and exterior walls. Every wall is divided into a large number of nodes. For each node, a conservation of energy equation expressed in finite-difference form accounts for transient conduction into and out of the node and temperature rise of the node. Tables 3-1 and 3-2 are summaries of the containment structural heat sinks used in the analysis.

The heat transfer coefficient to the containment structure is calculated by the code based primarily on the work of Tagami (Reference [7]). From this work, it was determined that the value of the heat transfer coefficient increases parabolically to a peak value at the end of blowdown for LOCA. The 00811: 1D/082793 3-3

value then decreases exponentially to a stagnant heat transfer coefficient which is a function of steam-to-air-weight ratio.

Tagami presents a plot of the maximum value of h as a function of "coolant energy transfer speed," defined as follows:

total coolant energy transferred into containment (containment volume) (time interval to peak pressure)

From this, the maximum h of steel is calculated:

--~--}***

h max = 75 ( t V p

(3.3-1) where:

hmax = maximum value of h (Btu/hr ft2 °F}.

  • tp v

=

=

time from start of accident to end of blowdown for LOCA and steam line isolation for secondary breaks (sec).

containment volume (ft3).

E = coolant energy discharge (Btu).

The parabolic increase to the peak value is given by:

h =h (.!_)o.s , O :::; t :::; t (3.3-2) s max t P p

where:

hs = heat transfer coefficient for steel (Btu/hr ft2 °F).

t = time from start of accident (sec) *

  • 00811: 10/082793 3-4

For concrete, the heat transfer coefficient is taken as 40 percent of the value calculated for steel (Reference [8]).

The exponential decrease of the heat transfer coefficient is given by:

-0.05(t-t ) (3.3-3) h =h +(h -h )e P t>tp s stag max stag where:

hstag = 2 + 50X, 0 ~ X ~ 1.4.

hstag = h for stagnant conditions (Btu/hr ft2 °F).

x = steam-to-air weight ratio in containment.

For a large break, the engineered safety features are quickly brought into operation. Because of the brief period of time required to depressurize the

  • reactor coolant system, the containment safeguards are not a major influence on the blowdown peak pressure; however, they reduce the containment pressure after the blowdown and maintain a low long-term pressure. Also, although the containment structure is not a very effective heat sink during the initial reactor coolant system blowdown, it still contributes significantly as a form of heat removal throughout the rest of the transient.

During the injection phase of post-accident operation, the emergency core cooling system pumps water from the refueling water storage tank into the reactor vessel. Since this water enters the vessel at refueling water storage tank temperature, which is less than the temperature of the water in the vessel, it can absorb heat from the core until saturation temperature is reached. During the recirculation phase of operation, water is taken from the containment sump and cooled in the residual heat removal heat exchanger.

The cooled water is then pumped back to the reactor vessel to absorb more decay heat. The heat is removed from the residual heat exchanger by component cooling water *

  • 00811: 1D/082793 3-5

Another containment heat removal system is the containment spray

  • Containment spray is used for rapid pressure reduction and for containment iodine removal. During the injection phase of operation, the containment spray pumps draw water from the RWST and sprays it into the containment through nozzles mounted high above the operating deck. As the spray droplets fall, they absorb heat from the containment atmosphere. Since the water comes from the RWST, the entire heat capacity of the spray from the RWST temperature to the temperature of the containment atmosphere is available for energy absorption. During the recirculation phase of post-accident operation, water can be drawn from the residual heat removal heat exchanger outlet and sprayed into the containment atmosphere via the recirculation spray system. However, recirculation spray is not modeled *in the COCO code in the analyses reported herein.

When a spray drop enters the hot, saturated, steam-air containment environment following a loss-of-coolant accident, the vapor pressure of the water at its surface is much less than the partial pressure of the steam in the atmosphere. Hence, there will be diffusion of steam to the drop surface and condensation on the drop. This mass flow will carry energy to the drop.

Simultaneously, the temperature difference between the atmosphere and the drop will cause the drop temperature and vapor pressure to rise. The vapor pressure of the drop will eventually become equal to the partial pressure of the steam, and the condensation will cease. The temperature of the drop will essentially equal the temperature of the steam-air mixture.

The equations describing the temperature rise of a falling drop are as follows:

d (3.3-4)

-(Mu)= mh +q dt g d (3.3-5)

-(M) = m dt

  • where:

00811: 1D/082793 3-6

q = hcA (Ts - T).

m = KgA (Ps - Pv).

The coefficients of heat transfer (he) and mass transfer (kg) are calculated from the Nusselt number for heat transfer, Nu, and the Nusselt number for mass transfer, Nu'.

Both Nu and Nu' may be calculated from the equations of Ranz and Marshall

{Reference [9]).

Nu= 2 + 0.6(Re) 112 (Pr) 113 (3.3-6)

Nu'= 2 + 0.6 (Re) 112 (Sc) 113 {3.3-7)

Thus, Equations 3.3-4 and 3.3-5 can be integrated numerically to find the internal energy and mass of the drop as a function of time as it falls through the atmosphere. Analysis shows that the temperature of the (mass) mean drop produced by the spray nozzles rises to a value within 99 percent of the bulk containment temperature in less than 2 seconds.

Drops of this size will reach temperature equilibrium with the steam-air containment atmosphere after falling through less than half the available spray fall height.

Detailed calculations of the heatup of spray drops in post-accident.

containment atmospheres by Parsly (Reference [10]) show that drops of all sizes encountered in the containment spray reach equilibrium in a fraction of their residence time in a typical pressurized water reactor containment.

These results confirm the assumption that the containment spray will be 100 percent effective in removing heat from the atmosphere. Nomenclature used in this section is as follows:

  • 00811: 1D/082793 3-7

Nomenclature A = area.

he = coefficient of heat transfer.

kg = coefficient of mass transfer.

hg = steam enthalpy.

M = droplet *mass.

m = diffusion rate.

Nu = Nusselt number for heat transfer.

Nu' = Nusselt number for mass transfer,

  • Ps Pv

= steam partial pressure.

= droplet vapor pressure.

Pr = Prandtl number.

q = heat flow rate.

Re = Reynolds number.

Sc = Schmidt number.

Ts = droplet temperature.

T = steam temperature *

  • 00811: 10/082793 3-8

t = time.

u = internal energy.

The reactor containment fan coolers are a final means of heat removal. The main aspect of a fan cooler from the heat removal standpoint are the fan and the banks of cooling coils. The fans draw the dense atmosphere through banks of finned cooling coils and mix the cooled steam/air mixture with the rest of the containment atmosphere. The coils are kept at a low temperature by a constant flow of cooling water. Since this system does not use water from the RWST, the mode of operation remains the same both before and after the spray system and emergency core cooling system change to the recirculation mode.

With these assumptions, the heat removal capability of the containment is sufficient to absorb the energy releases and still keep the maximum calculated pressure below the design pressure for the LOCA transient.

3.4 ANALYSIS RESULTS The results of the analysis shows that the maximum calculated containment pressure for the double-ended pump suction minimum safeguards break case is 41.2 psig and for the double-ended hot leg break case is 39.4 psig. The pressure peaks occur at approximately 712.8 seconds and 23.4 seconds, respectively.

The following plots show the containment integrity transient, as calculated by the COCO code.

Figure 3-1, .Containment Pressure Transient - DEPS Minimum SI Figure 3-2, Containment Pressure Transient - DEHL Figure 3-3, Containment Temperature Transient - DEPS Minimum SI Figure 3-4, Containment Temperature Transient - DEHL Table 3-1 shows the containment structural heat sink and material properties data used in the analysis, respectively.

00811:10/082793 3-9

The accident chronology for the double-ended pump suction loss-of-coolant accident (minimum SI case) and the double-ended hot leg case are shown in Tables 3-2 and 3-3, respectively.

Table 3-4 shows the assumptions and initial conditions for the containment analysis.

3.5 RELEVANT ACCEPTANCE CRITERIA The LOCA mass and energy analysis has been performed in accordance with the criteria shown in the Standard Review Plan (SRP) section 6.2.1.3. In this analysis, the relevant requirements of General Design Criteria (GDC) 50 and 10 CFR Part 50 Appendix K have been met since the calculated pressure is less than the design pressure, and because all available sources of energy have been included. These sources include: reactor power, decay heat, core stored energy, energy stored in the reactor vessel and internals, metal-water reaction energy, and stored energy in the secondary system.

The containment integrity peak pressure analysis has been performed in accordance with the criteria shown in the SRP section 6.2.1.1.A, for dry PWR containments. Conformance to GDC's 16, 38, and 50 is demonstrated by showing that the containment design pressure is not exceeded at any time in the transient. This analysis also demonstrates that the containment heat removal systems function to rapidly reduce the containment pressure and temperature in the event of a LOCA.

00811:10/082793 3-10

FIGURE 3-1 CONTAINMENT PRESSURE vs. TIME DEPS - MIN SI 45.

40. A
35. ~ L/ \

- 50. II \

\

t!J en Cl.

1...1

~ 25.

en en 1...1 C£ Cl.

z 20. J I

1...1

c z

a:

i!§ IS.

I\

I \

u 10.

s.

I ~

L/ \------

102 10 4 TIME !SECONOSl 07/28/93 00689: 10/121291 3-11

FIGURE 3-2 CONTAINMENT PRESSURE vs. TIME DEHL 40.

v- r'---

35. I I'-

30.

I I

~

UI

~ 25. I/

lJ ll

°'rn::::> V' rn

~ 20.

a... (

z lJ

c a:.....z I

0 u

z 15.

/

v 10.

5 i,....-

.... _..I"

/

Ii!

IB TI 11£ I SE:CONOS I 07/23/93

  • 00689: 10/12, 291 3-12

FIGURE 3-3 STEAM TEMPERATURE vs. TIME DEPS - MIN SI r ./

\

I \ \

I...

.20E*3 I ~

°'cr

i t-

...°'~

... .18E*3 I \

t-I ~

.14E*3

/ \

  • IC!E *3 l/ ~ \

10C!

Tl11E lSECONOSl 07/28/93 00689: 10/121291 3-13

  • FIGURE 3-4 STEAM TEMPERATURE vs. TIME DEHL 270.

260. ~

250. / r---.

,,, .. /

240.

v 250. I

- 220.

I/

Iii 210. I I

J a:

~ 200. I I/

Cl.

c
c 1q0.

a:

U'l 190. I 170. I 160.

I l!

I 150. '/

/

140 v 10 TI P1E: I SECONDS I DE t\O?:i' 07 /23/93

  • 00689: 1D/121291 3-14

TABLE 3-1

  • Wall Area CONTAINMENT STRUCTURAL HEAT SINKS Thickness Thermal Conductivity Volumetric Heat Capacity No. ill2l Laver Composition (ft) (Btu/hr-°F-ft2l (Btu/ft3-°F) 1 45169 1 Paint 0.000625 0.083 39.6 2 Steel 0.03125 27.0 58.8 3 Concrete 0.5 0.92 22.6 4 Concrete 4.0 0.92 22.6 2 14206 .1 Insulation 0.2083 0.024 3.94 2 Steel 0.03125 27.0 58.8 3 Concrete 0.5 0.92 22.6 4 Concrete 4.0 0.92 22.6 3 29249 1 Paint 0.000625 0.083 39.6 2 Steel 0.04167 27.0 58.8 3 Concrete 0.5 0.92 22.6 4 Concrete 3.0 0.92 22.6 41 11611 1 Paint 0.0015 0.083 39.6 2 Concrete 0.5 0.92 22.6 3 Concrete 3.0 0.92 22.6 5 6806 1 Paint 0.0015 0.083 39.6 2 Concrete 0.5 0.92 22.6 3 Concrete 0.5 0.92 22.6 4 Concrete 0.5 0.92 22.6 6 9424 1 Paint 0.0015 0.083 39.6 2 Concrete 0.5 0.92 22.6 3 Concrete 1.21 0.92 22.6 7 31660 1 Paint 0.00117 0.083 39.6 2 Concrete 0.5 0.92 22.6 3 Concrete 1.0 0.92 22.6 8 13279 1 Stainless 0.01773- 8.0 53.6 Steel 2 Concrete 0.5 0.92 22.6 3 Concrete 1.4 0.92 22.6 9 47590 1 Paint 0.000625 0.083 39.6 2 Steel 0.011 27.0 58.8 NOTE:
  • 1 In contact with sump.

D0811: 1D/082793 3-15

TABLE 3-1 (Cont'd)

CONTAINMENT STRUCTURAL HEAT SINKS Volumetric Thermal Heat Wall Area Thickness Conductivity Capacity

~ ill2l Layer Composition (ft) (Btu/hr-°F-ft21 (Btu/ft3~

10 76741 1 Paint 0.000625 0.083 39.6 2 Steel 0.02102 27.0 58.8 11 19348 1 Paint 0.000625 0.083 39.6 2 Steel 0.0437 27.0 58.8 12 9330 1 Paint 0.000625 0.083 39.6 2 Steel 0.0611 27.0 58.8 13 7452 1 Paint 0.000625 0.083 39.6 2 Steel 0.086 27.0 58.8 14 3218 1 Paint 0.000625 0.083 39.6 2 Steel 0.11124 27.0 58.8 15 1553 1 Paint 0.000625 0.083 39.6 2 Steel 0.217 27.0 58.8 16 43740 1 Paint 0.000625 0.083 39.6 2 Steel 0.0052 27.0 58.8 17 4272 1 Stainless 0.0329 8.0 53.6 Steel 18 53745 1 Paint 0.000625 0.083 39.6 2 Steel 0.0211 27.0 58.8 19 11244 1 Paint 0.000625 0.083 39.6 2 Steel 0.0379 27.0 58.8 20 2989 1 Paint 0.000625 0.083 39.6 2 Steel 0.15806 27.0 58.8

    • 00811: 1D/082793 3-16

TABLE 3-2 CHRONOLOGY OF EVENTS FOR LOCA - OEPS MIN SI Time (Seconds) Event a.a Start of accident 27.0 End of blowdown phase 37.1 Pumped Safety Injection starts 56.28 Containment fan coolers start 89.70 Containment sprays start

  • 198.1 712.8 End of reflood phase Peak Pressure Reached 1704.0 Sump recirculation starts
    • 00811: 1 D/082793 3-17
    • TABLE 3-3 CHRONOLOGY OF EVENTS FOR LOCA - DEHL Time (Seconds) Event 0.0 Start of accident 27.0 End of blowdown phase
  • 00811: 1D/082793 3-18

TABLE 3-4

  • CONTAINMENT ANALYSIS PARAMETERS Service water temperature (°F) 95 Refueling water temperature (°F) 100 RWST water deliverable volume (gal) 3.13 x 105 Initi a1 containment temperature (°F) 120 Initial containment pressure (psia) 15.0 Initial relative humidity (%) 20 Net free volume (ft3) 2.62 x 106 Fan Coolers
  • Total Operating minimum Setpoint (psig) 5 3

6.0 Delay time (sec) 55 Spray Pumps Minimum safeguards spray flow per pump (gpm) 2600 Total 2 Operating maximum 2 Operating Minimum 1 Setpoint (psig) 17.0 Delay time (sec) 85

  • D0811: 1D/082793 3-19

TABLE 3-4 (Cont'd)

  • .CO'NT,AINMENT ANALYSIS PARAMETERS Fan Cooler Heat Removal Btu/sec per fan Containment Temperature (°F) Heat Removal

. 120 1600

  • 140 3355.

160 5389.

180 7667.

200 9897.

240 14511.

260 16858.

  • 271 280 18047.

19053.

(based on a service water flow of 2500 gpm per fan cooler)

  • DOB 11: 1D/082793 3-20

4.0 CONCLUSION

S Based upon the information provided in this report, it is demonstrated that upgrading the fuel to Vantage Plus fuel with Performance Plus features, reducing fan cooler heat removal, increasing the safeguards delays, and reducing SI flow will yield an acceptable pressure margin in the containment.

This analysis resulted in an overall calculated peak containment pressure of 41.2 psig and a overall peak containment temperature of 261.5°F for the DEPS break. The DEHL case was less limiting, and resulted in a peak pressure and temperature of 39.4 psig and 260.6°F.

The calculated pressure margin is 5.8 psi available to the containment design pressure of 47.0 psig for the DEPS minimum safeguards case. Thus the most limiting LOCA cases have been considered and have been demonstrated to yield acceptable pressure results. PSE&G has scope to revise the baseline equipment qualification temperature profile to address the temperatures resulting from this analysis *

  • 0081,: 10/082793 4-1

TABLE 4-1 CONTAINMENT PEAK PRESSURE Peak Pressure BREAK (psig)

Double-ended pump suction - Min SI 41.2 3411 MWt Fuel Upgrade Conditions Reduced Fan Cooler Performance Reduced RHR, Ch/SI & !HSI flows Increased Delays Reduced Containment Spray Setpoint

  • Double-ended hot leg 3411 MWt Fuel Upgrade Conditions 39.4
  • 00811: 1D/082793 4-2

5.0 REFERENCES

1. "Westinghouse LOCA Mass and Energy Release Model for Containment Design -

March 1979 Version", WCAP-10325-P-A, May 1983 (Proprietary), WCAP-10326-A (Non-Proprietary).

2. "Westinghouse ECCS Evaluation Model - 1981 Version", WCAP-9220-P-A, Rev. 1, February 1982 (Proprietary), WCAP-9221-A, Rev. 1 (Non-Proprietary).
3. EPRI 294-2~ Mixing of Emergency Core Cooling Water with Steam: 1/3 Scale Test and Summary, (WCAP-8423), Final Report June 1975. *
4. "Westinghouse Mass and Energy Release Data For Containment Design",

WCAP-8264-P-A, Rev. 1, August 1975 (Proprietary), WCAP-8312-A (Non-Proprietary).

ANSI/ANS-5.1-1979, "American National Standard for Decay Heat Power in Light Water Reactors", August 1979.

6. "Containment Pressure Analysis Code (COCO)", WCAP-8326, July 1974 (Non-Proprietary) WCAP-8327 (Proprietary).
7. Takashi Tagami, "Interim Report on Safety Assessments and Facilities Establishment Project in Japan for Period Ending June 1965 11 , No. 1.
8. "Salem Generating Station Updated Final Safety Analysis Report",

Section 15.4.8.1.3, Heat Sinks.

9. E. W. Ranz and W. R. Marshall, Jr., "Evaporation for Drops", Chemical Engineering Progress, 48, pp. 141-146, March 1952.
10. L. F. Parsly, "Spray Tests at the Nuclear Safety Pilot Plant", Nuclear Safety Program Annual Progress Report for Period Ending December 31, 1970, ORNL-4647, 1971, p. 82
  • 00811: 1D/082793 5-1
11. "Salem Nuclear Generating Station Units 1 and 2, Nuclear Steam Supply System and Turbine Generator Rerating Feasibility Report", WCAP-12491, January 1990 (Proprietary).
12. PSE-93-647, "Fuel Upgrade/Margin Recovery Program Consolidated Input Assumptions Document for the Accident Analyses," June 29, 1993 *
  • 00811:10/082793 5-2
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