ML19312A218

Chapter 15 of S&W SWESSAR-P1, Accident Analysis.

ML19312A218
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Site:	05000495
Issue date:	11/29/1978
From:	NEW YORK STATE ELECTRIC & GAS CORP., STONE & WEBSTER, INC.
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NUDOCS 7909060013
Download: ML19312A218 (150)

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SWESSAR-P1 CHAPTER 15 ACCIDENT ANALYSIS LIST OF EFFECTIVE PAGES Page, Table (T), Amendment Page, cable (T) , Amendment or Figure (F) No. or Figtte (F) No.

15-a 39 F15.1.18-2 63 (W) 11 15-i/ii 19 F15.1.20-1Di) 8 15-iii 8 (2 sheets) 15 -i' r 21 F15.1.20-2 00 (3 sheets) 8 15-v 35 F15,1.2 0-3 (W) 0 15-vi,vii 21 F15.1.20-4 04) 8 15.1-1 thru 4 19 F15.1.14-1 thru 4(W-3S) 17 15.1-5/6 20 F15.1.16-162 (W-3S) 17 15.1-6A 21 F15.1.18-1 thru 3 (W-3S) 17 15.1-7/8 28 F ' 5.1.2 0 -1 [W-3S) 17 15.1-9 3 (2 sheets) 15.1-10,10A 7 F15.1.2 0-2 (W-3S) 17 15.1-11/12 11 (3 sheets) 15.1 *'/14 8 F15.1.20-3 (W-3S) 17 15.1-1b 33 F15.1.20-4 (W-3S) 17 15.1-16 35 F15.1.14-1 thru 4 (;BSJ) 5 15.1-16A 33 F15.1.16-1 (BSW) 5 15.1-17 thru 21 Orig F15.1.16-2 (B&W) 5 15.1-22 thru 24 8 F15.1.18-1 thru 4 (BSW) 5 T15.1-1 7 F15.1.20-1 (BSW) 5 T15.1-2 17 (2 sheets)

T15.1-3 7 F15.1.20-2 (BSW) 5 T15.1.13-162 19 (3 sheets)

T15.1.14-1 19 F15.1.20-3 (BSW) 5 T1 f .1.14-2 (4 sheets) 19 (2 sheets)

T15.1.16-1 Pi) 17 F15.1.20-4 (BSW) 5 T15.1.16-1 (W-3S) 17 F15.1.14-1 thru 4 (C-E) 11 T15.1.16-1 (B&W) 1 F15.1.16-1 (C-E) 5 T15.1.16-1 (C-E) 3 F15.1.16-2 (C-E) 5 T15.1.18-1 19 F15.1.18-1 (C-E) 11 T15.1.18-2 (3 sheets) 19 (2 sheets)

T15.1.20-1 24 F15.1.18-2 6 3 (C-E) 11 T15.1.20-2 (sheet 1) 24 F15.1.2 0-1 (C-E) 24 T15.1.20-2 (sheets 2S 3) 19 (2 sheets)

T15.1.23-1 35 F15.1.20-2 (C-E) '

24 T15.1.2 3-2 (2 sheets) 35 (3 shaets)

T15.1.23-3 35 F15.1.20-36 4 (C-E) 24 F15.1.13-1 7 F15.1.2 3-1 (W) 9 F15.1.13-2 5 F15.1.23-2 PO 9 F15. l .13-3 (3 sheets) 5 F15.1.23-1(W-3S) 17 F15.1.13-4 thru 7 5 F15.1.23-2 [W-3S) 17 F15.1.13-869 23 F15.1.2 3-1 (C-E) 9 F15.1.14-1 thru 4 pi) 5 F15.1.2 3-2 (C-E) 9 F15.1.16-162 pi) 17 15.2-1 21 F15.1.18-1 (W) 5 T15.2-1 Di) (4 sheets) 21 15-a _g n, - Amendment 39 U/U U[3 7/14/78

SWESSAR-P1 CHAPTER 15 ACCIDENT ANALYSIS TABLE OF CONTENTS Section Pace 15.1 GENERAL 15.1-1 15.1.1 Uncontrolled Rod Assembly Withdrawal from 15.1-4 a Subcritical Condition 15.1.2 Uncontrolled Control Rod Assembly With- 15.1-4 19 drawal at Critical Power 15.1.3 Control Rod Misoperation or Sequence of 15.1-4 Misoperations 15.1.4 Chemical and Volume Control System Mal- 15.1-4 function 15.1.5 Partial and Total Loss of Reactor Coolant 15.1-4 Flow Force Including Trip of Pumps and Pump Shaft Seizures 15.1.6 Startup of an Inactive Reactor Coolant 15.1-4 Ix>op or Recirculating Loop at Incorrect Temperature 15.1.7 Loss of External Electrical Load and/or 15.1-4 Turbine Stop Valve Closure 15.1.8 Loss of Normal and/or Emergency Feedwater 15.1-4 Flow 15.1.9 Loss of All A-c Power to the Station 15.1-4 Auxiliaries 15.1.10 Heat Removal Greater than Heat Generation 15.1-5 Due to (1) Feedwater System Malfunctions, (2) a Pressure Regulator Failure, or Inad-vert ent Opening of a Relief Valve or Safety Valve, and (3) a Regulating Instrument Failure 15.1.11 Failure of Regulating Instrumentation 15.1-5 39 15.1.12 Internal and External Events (Fires, Storms, 15.1-5 Floods, and Earthquakes) 15.1.12.1 Identification of Causes 15.1-5 15-i n4 - Amendment 19

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SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section Page 15.1.12.2 Analysis of Ef f ects and Consequen'_:es 15.1-5 15.1.13 Loss of Coolant Accident -

15.1-5 15.1.13.1 Identification of Causes 15.1-5 15.1.13.2 Analysis of Effects and Consequences 15.1-6 15.1.14 Steam and Feedwater Breaks 15.1-7 15 .1.1 .1 Identification of Causes 15.1-7 15.1.14.2 Analysis of Ef f ects and Consequent :s 15.1-7 15.1.14.3 NSSS Modification (RESAR-41 only) 15.1-7 15.1.15 Inadvertent Loading and Operation of a 15 1-9 Fuel Assembly into an Improper Position 15.1.16 Process Gas Charcoal Bed Adsorber 15.1-9 Line Rupture 15.1.16.1 Identification of Causes 15.1-9 15.1.16.2 Analysis of Effects and Consequences 15.1-10 19 15.1.17 Failure of Air Ejector Lines 15.1-10A 15.1.18 Steam Generator Tube Rupture 15.1-10a 15.1.18.1 Identification of Causes 15.1-11 15.1.*8.2 Analysis of Effects and Consequences 15.1-11 15.1.19 Failure of Charcoal or Cryogenic 15.1-11 System (BWR) 15.1.20 Rod Ejection Accident 15.1-11 15.1.20.1 Identification of Causes 15.1-11 15.1.20.2 Analysis of Effects and Consequences 15.1-13 15.1.21 The Spectrum of Rod Drop Accidents 15.1-15 15.1.22 Break in Instrument Line or Other Lines 15.1-15 from Reactor Coolant P'; essure Boundary

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SWESSAF-P1 TABLE OF CONTENTS (CONT)

Section Pace that Penetrate Containment 15.1.23 Fuel Handling Accident 15.1-15 15.1.23.1 Identification of Causes 15.1-15 15.1.23.2 Analysis of Eff ects and Consequences 15.1-16 15.1.24 la Small Spills or Leaks of Radioactive 15.1-16 Eaterial Outside Containment 15.1.24.1 Identification of Causes 15 1-16 15.1.24.2 Andlysis of Eff ects and Consecmences 15.1-16 15.1.25 Fuel Cladding Failure Combined with Steam 15.1-17 Generator Leak 15.1.26 Control Room Uninhabitability 15.1-17 15.1.26.1 Identification of Causes 15.1-17 15.1.26.2 Analysis of Effects and Consequences 15.1-17 15.1.27 Failure or Overpressurization af Low 15.1-17 Pressure Residual Heat Pemoval System 15.1.28 Ioss of Condenser vacuum 15.1-18 15.1.28.1 Identification of Causes 15.1-18 15.1.28.2 Analysis of Eff ects and Consequences 15.1-18 15.1.29 Turbine Trip with Coincident Failure of 15.1-18 Turbine Bypass Valves to Open 15.1.29.1 Identification of Causes 15.1-18 15.1.29.2 Analysis of Eff ects and Consequences 15.1-19 15.1.30 Ioss of Reactor Plant Service Water 15.1-19 System 15.1.30.1 Identification of Causes 15.1-19 15.1.30.2 Analysis of Eff ects and Consequences 15.1-19 15.1.31 Loss of One D-c System 15.1-19 15-15i

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SWESSAR-P1 TABLE OF CONTENTS (CONT)

Section '- Page 15.1.31.1 Identification of Causes 15.1-19 15.1.31.2 Analysis of Ef f ects and Consequences 15.1-20 15.1.32 Inadvertent Operation of ECCS During Power 15.1-21 Operation 5.1.33 Turbine Trip with Failure of Generator 15.1-21 Breaker 15.1.33.1 Identification of Causes 15.1-21 15.1.33.2 Analysis of Ef fects and Consequences 15.1-21 15.1.34 Loss of Instrument Air System 15.1-22 15.1.34.1 Identific-tion of Causes 15.1-22 15.1.34.2 Analysis of Effects and Consequences 15.1-22 15.1.35A Malfunction of Turbine Gland Se- ling 15.1-23 System for General Electric Turbine 15.1.35A.1 Identification of Causes 15.1-23 15.1.35A.2 Analysis of Effects and Consequences 15.1-23 15.1.35B Malfunction of Turbine Gland Sealing 15.1-24 System for Westinghouse Turbine 15.1.35B.1 Identification of Causes 15.1-24 15.1.35B.2 Analysis of Effects and Consequences 15.1-24 p 15.2 BALANCE OF PLANT ASSUMPTIONS ITPILIZED 15.2-1 IN THE NSSS VENDOR ACCIDENT ANALYSES O

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SWESSAR-P1 LIST OF TABLES Table 15.1

• Not Used 15.1-2 Iodine Concentration in Steam Generator Liquid with 1,000 GPD Primary-to-Secona -v Teakage and 1 Percent Failed Fuel 15.1-3 Not Used 15.1.13-1 Reactor Containment Nuclide Inventory Available for Leakage Immediately after LOCA 15.1.13-2 Parameters Used for the Loss of Coolant Accident Analysis 15.1.14-1 Steam Pipe Break Accident Releases 15.1.14-2 Parameters Used for the Steam Pipe Break Accident Analysis 15.1 16-1 Maximum Radioisotope Releases from the Process Gas Charcoal Bed Adsorber and Associated Piping 15.1.18-1 Steam Generator Tube Rupture Releases 15.1.18-2 Parameters Used for the Steam Generator Tube Rupture Accident Analysis 15 .1.20 -1 Rod Ejection Accident Releases 15.1.20-2 Parameters Used for the Rod Ejection Accident Analysis 15.1.23-1 Fuel Handling Accident Releases to Atmosphere 15.1.23.2 Parameters Used for the Fuel handling Accident Analysis 15.1.23-3 Deleted 35 15.2-1 Balance of Plant Assumptions Used in Westinghouse Accident Analysis 15-v Amendment 35

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SWESSAR-P1 LIST OF FIGURES Figure 15.1.13-1 Source Strength in Containment as a Function of Time at Various Energy Groups 15.1.13-2 Loss of Coolant Accident, Dose vs CHI /Q, 2 Hour Dose at Exclusion Eoundary 15.1.13-3 Loss of Coolant Accident, Dose vs CHI /Q, 30 Day Dose at Low Population Zone, Sheet 1 15.1.13-3 Loss of Coolant Accident, Dose vs CHI /Q, 30 Day Dose at Low Population Zone, Sheet 2 15.1.13-3 Loss of Coolant Accident, Dose vs CHI /Q, 33 Day Dose at Low Population Zone, Sheet 3 15.1.13-4 Loss of Coolant Accident, Dose vs Distance, 2 Hour Dose at Exclusion Boundat}

15.1.13-5 Loss of Coolant Accident, Dose vs Distance, 30 Day Dose at Low Population Zone 15.1.13-6 Direct Dose Rate from Containment Structure at Dif f erent Distances and Times af ter a LOCA (

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15.1.13-7 Cumulative External (Whole Body) Dose trom Direct Radiation from Containment Structure after a LOCA 15.1.13-8 Increase in Iodine Release Due to Exfiltration After a LOCA 21 15.1.13-9 Ef f ect ot Wind Speed on the 0-2 Hour Thyroid Dose 15.1.14-1 Steam Pipe break Accident, Dose vs CHI /O 15.1.14-2 Steam Pipe Break Accident, Dose vs CHI /Q 15.1.14 -3 Steam Pipe Ereak Accident, Dose vs Distance 15.1.14-4 Steam Pipe Break Accident, Dose vs Distance 15.1.16-1 Process Gas Charcoal Decay Bed Adsorber Line Rupture, Dose vs CHI /Q, 2 Hour Dose at Exclusion Boundary 15.1.16-2 Process Gas Charcoal Decay Bed Adsorber Line Rupture, Dose vs Distance, 2 Hour Dose at Exclusion boundary n^;

O 15-vi [jU U'I Amendment 21 2/20/76

SWESSAR-P1 LIST OF FIGURES (COh"I)

_ Figure 15.1.18-1 Steam Generator Tube Rupture Dose vs CHI /Q at Exclusion Boundary 15.1.18-2 Steam Generator Tube Rupture, Dose vs Distance, 2 Hour Dose at Exclusion Boundary 15.1.18-3 Steam Generator Tu.e Rupture, Dose vs Distance, 30 Day Dose at Iow Population Zone 15.1.18-3 Steam Generator Tube Rupture, Dase vs Distance, 2 Hour Dose at Exclusion Boundary (BSW only) 15 .1.18-4 Steam Generator Tube Rupture, Dose vs Distance, Doses at Exclusion Boundary (B&W only) 15.1.20-1 Rod Ejection Accident, Dose vs CHI /0, 2 Hour 90se at Exclusion Boundary, Sheet 1 15.1.20-1 Rod Ejection Accident, Dose vs CHI /Q, 2 Hour Doce at Exclusion Boundary, Sheet 2 15.1.20-2 Rod Ejection Accident, Dose vs CHI /Q, 30 Day Dose at Low Population Zone, Sheet 1 15.1.20-2 Rod Ejection Accident, Dose vs CHI /Q, 30 Day Dose at Lo5i Population Zone, Sheet 2 15.1.20-2 Rod Ejection Accident, Dose vs CHI /Q, 30 Day Dose at Low Population Zone, Sheet 3 15.1.20-3 Rod Ejection Accident, 2 Hour Dost at Exclusion Boundary, Sheet 1 15.1.20-3 Rod Ejection Accident, 2 Hour Dose at E::clusion Boundary, Sheet 2, BSW Only 15.1.20-4 Rod Ejection Accident, 30 Day Dose at Iow Population Zone 15.1.23-1 Fuel Handling Accident, Dose vs CHI /Q 15.1.23-2 Fuel Handling Accident, Dose vs Distance 15-vii n ,3 g Amendment 21 6]O U t. L 2/20/76

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SWESSAR-P1 CHAPTER 15 3 ACCIDENT ANALYSES Site-independent dose analysis for postulated accidents for both the 2 h our exclusion boundary (EB) dose and the 30 day low population zone (LPZ) dose are reported graphically as dose vs CHI /Q. This is to permit evaluation of the design on the basis of meteorological conditions experienced at specific sites as described in the Utility-Applicant's SAR. Upon determination of the short term meteorology for a specific site, the doses can be determined directly from these curves. Accident meteorological data collected for a number ot existing sites are presented in Chapter 2.

In order to provide an indication of the EB and LPZ distances under consideration, the analysis of doses associated with postulated acci dents is reported on a dose vs distance basis.

The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses are plotted vs EB distance using Regulatory Guide 1.4 assumptions and meteorological dispersion. Similarly the 30 day doses are plotted vs distance to the outer boundary of the LPZ. The distance to the EB and LPZ for potential sites can be read of f these curves to abtain the calculated doses.

A review of 53 sites , comprising 81 nuclear units for which the USNRC has issued site-specific environmental impact statements, indicates a range of exclusion boundary distances from 910 to 7,000 feet with an average value of 2,600 ft. Tl range of distances to the outer boundary of the low populativa zones for these sites is 1.5 to 10 miles with an average value of 3.9 miles.

15.1 GENERAL Normal and abnormal operations of the various systems and com-ponents and the susceptibility of individual compo'ents to mal-function or f ailure are discussed in applicable sections throughoet this report. This chapter summarizes and further explores the consequences of these abnormal situa tions or failures. All accidents discussed in this chapter have potential offsite radiological consequences and therefore have been evaluated at a core power level of 4,100 MWt ( (C-E , W-41) , 3,876 g MWt (BSW) or 3,636 MWt (W-3 S) . These power levels are consistent with the regu' rements of Regulatory Guide 1.49, " Power levels of Water-Cooled duclear Power Plants," Revision 1, issued December 1973. The transient conditions resulting from all accidents are shown by analysis to oe:

1. Inherently terminated.

2. Terminated by the Operation of the reactor protection systam which maintains the integrity of the fuel and/or the reactor coolant sys. em.

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SWESSAR-P1

3. Terminat ed by o ther conditions vhich result in the operation of engineered safety features, which maintain ['

the integrity of the core and/or the containment and L.

reduce the potential offsite doses to the public when one or more of the protective barriers are not effective.

The events examined are those listed in Table 15-1 of Regulatory Guide 1.70, "Stan dard Format and Content of Safety Analysis heports of Nucl ear Power Plants," (Revision 1) , issued October 19 1972. This listing agrees closely with the accidents recommended by ANSI N18.2-1973, Nuclear Safety Criten A for the Design of Stationary Pressurized Water Reactor Plants. Although all the accidents recommended by ANSI N18.2 are not required by Regulatory Guide 1.70, they are incorporated as part of other accidents which are required by the Guide. Only those accidents pertinent to the plant design are included in this SAR. Other accidents are analyzed in the SAR of the NSSS Vendor or Utility-Applicant.

In each accident analysis a description of cause and eff ect and order of occurrence for the postulated event are provided. The amcunt of depen dence on the rea ctor protection system (RPS) operation and the conservative values of important design parameters such as reactivity feedback coefficients are indicated in each analysis. Each accident analysis states the RPS or engineered safety features function used to terminate the g tr ansi ent . Reactor system variables which are monitored to provide core protection are summarized in Chapter 7 of the NSSS (

Ver. dor 's SAR .

All systems or functions utilized in these accident analyses are designed in a manner such that a single failure of an active component does not prevent them from performing the intended safety function. No operator action is required for reactor protection. Operator action for maintaining hot shutdown conditions or cooldown to cold shutdown conditions is required only where adequate time and suf ficient system monitoring are available to the operator.

The evaluation of each accident is based upon conservative engineering assumptions to provide a margin for uncertainties in calculational me thods . RPS trip values used in the accident analyses are based on the maximum setpoint value plus the maximum itncertainties in measurement. The uncertainties associated with equipment and instrumentation performance are discussed in Chapter 7 of the NSSS Vendor's SAR. Uncertainties in calculated values of parameters are considered by a sensitivity analysis for those parameters.

The design and interdependence of the various engineered safety feature (ESF) systems are discussed in Chapter 6. Eac' ESF system has been designed with sufficient capacity, structural integrity, and redundancy so that a single malf unction or f ailure  ;

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of an active component within any one or even within each system n-15.1-2 f.7P '

Amendment 19 12/12/75

SWESSAR-P1 does not compromise the intended operation of tne other ESF g systems which are directly or indirectly involved in controlling 5 the fission product release or limiting the leakag from the containment stru cture . Whenever ESF systems are used in the accident analysis, systematic malfunctions or failures are taken into account by assuming the minimum cerformance level.

An evaluation of environmental ef fe cts is presented for each accident that results in offsite radiation exposures in excess of normal operating releases. In general, the dose to the thyroid and gamma and beta doses as a function of CHI /Q and distance are given for these accidents. Additional dose values (as described in Reguldtory Guide 1.70) are presented in the unalysis when they are required in the overall evaluation of the consequences of a particular accident. Dose values are also presanted for analyses made in accordance with enrrent Regulatory Guidas governing certain accidents. The atmosy ric dispersion factors used in calculating the doses vs distance are based on Regulatory Guide 1.4 (Section 3A.1-1.4) and are given in Section 2.3.

A description of the physical and mathematical models employed in calculating radiation source terms is given in Chapter 11.

Radiation source terms used in the dose calculations include individual isotopic activities of fission products in tuel, fuel rod gap, reactor coolant system, and steam and power conversion system.

Table 11.1.1-1 lists the iodine and noble gas inventory in the core based on 14,700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> of operation at 4,100 MWt ( C-E ,

W-41 ) , 3,876 MWt (B&W) or 3,6 36 MWt (W-35).

Table 15.1-2 lists the iodine concentration in the stean; generator liquid based on 1 percent failed fuel and primary-to-secondary leakage of 1,000 gpd.

Table 11.1.1-1 lists the iodine and noble gas activities in fuel gaps used in evaluating the accidents described in Regulatory Guides 1.26 and1.77 (Sections 3A.1-1.26 and 3A.1-1.77) .

In evaluating the following accidents for the B-SAR 205 design, the curies released and the resulting doses are generally taken from previous analyses which were based on values of parameters slightly different than those given for the accidents listed below. Since the results are consistent with those already presented for W and C-E KSSS, revisions in the calculations are unnecessary for this application for Prelimina:;y Design Approval.

a. Steam line break, Section 15.1.14 19

b. Process gas charcoal bed adsorber line rupture, Section 15.1.16

c. Steam generator tube rupture, Section 15.1.18

d. Fod ejection accident, Se ction 15.1.20 670 026 15.1-3 Amendment 19 12/12/75

SWESSAR-P1 19 e. Fuel handling accident, section 15.1.23 -

15 .1.1 Uncontrolled Control Rod Assembly Withdrawal f rom a Subcritical Condition This condition is within the NSSS Vendor 3s scope and SAR.

15.1.2 Uncontrolled Control Rod Assembly Withdrawal at Critical Power This condition is within the NSSS Vendt -'s scope and SAR.

15.1.3 Control Rod Misoperation or Sequence of Misoperations This condition is within the NSSS Vendor's scope and SAR.

• 5.1.4 Chemical and Volume Control System Malfunction This condition is within the NSSS Vendor's scope and SAR. 15.1.5 Partial and Total Loss of Reactor Coolant Flow Force Including Trip of Pumos and Pump Shaft Seizures This conditi(n is within the NSSS Vendor' scope and SAR.

15.1.6 Startup of an Inactive Reactor Coolant Loop or Recirculating Loop at Incorrect femperature This condition is within tne NSSS Vendor's scope and SA; .

15.1.7 Loss of External Electrical Load and/or Turbine Stoo Valve Closure This condition and its consequences are discussed in Chapter 15 of the NSSS Vendor's SAR.

The PWR Reference Plant design can accommodate up to 50 percent load rejection without reactor or turbine trip, and without steam release to atmosphere. A load rejection greater than 50 percent will initiate both a reactor and a turbine trip, end may release steam to atmosphere depending on condenser availability. The radiological consequences resulting from this steam release have as an upper bound the radiological consequences evaluated for a steam pipe rupture (Section 15.1.14).

15.1.8 Loss of Normal and/or Emergency Feedwater Flow This condition is within the NSSS Vendor's scope and SAR.

15.1.9 Loss of All A-c Power to the Station Auxiliaries (Station blackout)

The radiological consequences of this accident have as an upper bound the radiological consequences evaluated for a steam pipe 1) rupture (Section 15.1.14).

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15.1.10 Heat Removal Greater Than Heat Generation Due to (1)

Feedwater System Malfunctions, (2) a Pressure Regulator Failure, or Inadvertent Opening of a Relief Valve or Safety Valve, and (3) a Regulating Instrument _ Failure This condition is within the NSSS Vendor's scope and SAR.

15,1.11 Failure of Regulating Instrumentation This condition is within the NSSS Vendor's scope and SAR.

15.1.12 Internal and External Events (Fires, St)rms, Floods, or Earthquakes 15.1.12.1 _Id.entification of Causes The severe natural phenomena considered to have a potential for an abnormal effect on the safety related plant structures and equipment are tornadoes, severe earthquakes, and floods.

20 The probability of occurrence and descriptions of these evrnts are given in the Utility-Applicant's SAR. Missile generation from both internal and external events is discussed in Section 3.5.

The possibility of a fire exists in areas where smoking is allowed, in areas where chemicals e.re stored, fuel storage tank areas, and areas containing electrical equipment as discussed in Section 9.5.1.

15.1.12.2 Analysis of Effects and Consequences Safety related (QA Categroy I) stru ctures , systems, and components listed in Table 3.2.5-1 are designed to withstand the severe natural phenomena described above without impairment of their safety functions.

For tornado protection, safety related equipment is housed in and protected by structures designed to withs tand the effects of tornado forces. Ite tornado criteria are oiven in Section 3.3.2 and the design of structuret to withstand tornadoes is discussed in Section 3.8.

Safety related equipment, along with the structures which support that equipment, are designed to withstand the safe shutdown earthquake (SSE) and the operating basis earthquake (OBE).

Section 3.7 describes the seismic analysis an'd design methods for Seismic Category I equipment piping. The seismic analysis of Seismic Categ ry I structures is described in Section 3.7.2 and the design of such structures is discussed in Section 3.8.

Missile protection for the various missiles considered is discussed in Sections 3.5 and 3.8.

15.1-5 <Tr, Amendment 20 D O f'4 v

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SWESSAR-P1 The fire protection system is designed to detect, extinguish, and mitigate the consequences of a fire by utilizing the water fire protection system, the CO2 fire protection system, portable fire er.tinguishers, and fire hoses as discussed in Section 9.5.1.

The plant design accommodates the various internal and external events summarized above. For a detailed discussion of specific occurrences, consult the referenced sections.

15.1.13 Loss of Coolant Accident The loss of coolant accident is described in Chapter 15 of the NSSS Vendor's SAR.

15.1.13.1 Identification of Causes Causes of the above accident are delineated in Chapter 15 of the NSSS Vendor 5s SAR.

15.1.13.2 Analysis of Ef fects and Consequences Nonra diological consequences of the above accident are described in Chapter 15 of the NSSS Vendor's SAR and in.Cha: 1er 6.

This section describes the method of radiological analyses for this incident. These analyses are divided into two parts: (1) -

thyroid dose from inhaling the iodines in the containment leakage plume, and (2) garma and beta exposure as a result of immersion in the leakage plume and additional gamma dose from the activity in the containment structure.

The fission product inventory available for leakage from the reactor containment structure is shown in Table 15.1.13-1.

Methods and assumptions similar to those given in TID-14844 and Regulatory Guide 1.4 are used. These assumptions are listed in Table 15.1.13-2. Additional assumptions are added for Options A and B in Tables A15.1.13-2 and B15.1.13-2, respect ~ vely. The iodine and noble gas inventory in the core is listed in Table 11.1-1. Source strength inside the containment structure as a function of time at various energy groups is shown in Fig. 15.1.13-1.

The site-independent 2-hour thyroid, beta, and gamma doses from immersion in the leakage plume as a function of CHI /Q, are shown in Fig. 15.1.13-2. Thirty-day thyroid, beta, and gamma doses from immersion in the leakage plume as a function of CHI /O are shown in Fig. 15.1.13-3, Sheets 1 through 3 respectively. Fig.

15.1.13-2 through 15.1.13-5 are revised for the effects of the enclosure building under both Options A and B in Fig. A15.1.13-2 through A 15 .1.13- 5 and B15.1.13-2 through B15.1.13-5, respectively.

Two-hour thyroid, gamma, and beta doses at various distances from the containment structure are shown in Fig. 15.1.13-4. Thirty-15.1-6 Amendment 20 6 /l;.

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SWESSAR-P1 day thyroid, gamma, and beta doses at various distances from the i containment structure are shown in Fig. 15.1.13-5. The gamma dose is based on the summation of the integrated external doses from immersion in the leakage plume and direct dose from the activity contained within the containment structure. External dose rates and cumulative dose from the activity contained within the contaiment structure as functions of time and distances are given in Fig. 15.1.13-6 and 15.1.13-7 respectively.

As can be seen from the radionuclide inventory in Table 15.1.13-1, the analysis at 4,100 MWt is limiting, though by a relatively small amount. Thus, Fig. 15.1.13-1 through 15.1.13-7 reflect the 4,100 MWt analysis only and represent an upper bound for the 3,876 and 3,636 MWt analysis.

The increase in activity released due to exfiltration from areas served by the SLCRS as a function of windspeed is presented in Fig. 15.1.13-8 based on the exfiltration rates presented in Section 6A.3. The effect of exfiltration on offsite doses is l21 site-related and will be addressed in the Utility-Applicar t's SAR. In most cases, however, the concurrent improvement in the atmospheric dispersion factor, CHI /Q, will be sufficient to of f set the increased iodine release. An example of this is given in Fig. 15.1.13-9 where the change in thyroid dose from both exfiltration and CHI /Q as a function of wind speed is compared to 21 the thyroid dose based on the 1 m/sec wind speed used in Regulatory Guide 1.4 (Section 3A.1 -1.4) .

This accident will be addressed in the Utility - Applicant's SAR, relative to the limiting value of CHI /Q.

- fi'" , tj b! c 15.1-6A Amendment 21 2/20/76

SWESSAR-P1 15.1.14 Steam and Feedwater Breaks Rupture of a main steam pipe inside or outside of containment and rupture of a feedwater pipe are described in Chapter 15 of the NSSS Vendor's SAR.

15 .1.14.1 Identification of Causes Causes of the above accidents are delineated in Chapter 15 of the NSSS Vendor's SAR.

15.1.14.2 Analysis of Effects and Consecuences Nonradiological consequences of the above accidents are described in Chapter 15 of the NSSS Vendor's SAR. The effect on containment pressure is covered in Section 6.2.1.

The source of radiation discharged. to the environment from a steam pipe rupture presumes a primary-to-secondary systtan leak.

We absence of such a leak precludes the discharges of any radioactive material f rom the steam and power conversion system.

A conservative analysis of radiological consequences resulting from a main steam pipe break outside the containment structure is presented with steam generator leakage and assuming loss of offsite power. The assumptions used in the analysis of the main steam pipe break are given in Table 15.1.14-2.

Total iodines and noble gases released during the accident are given in Table 15.1.14-1.

The site-independent thyroid doses as a function of CHI /Q are shown in Fig. 15.1.14-1. The corresponding gamma and beta deses are presented in Fig. 15.1.14-2.

The thyroid, gamma, and beta doses for the main steam pipe break at various distances from point of release, based on Regulatory Guide 1.4 (Section 3A.1-1.4) meteorology, are shown in Fig. 15.1.14-3 and 15.1.14-4.

As can be seen from the radionuclide releases in Table 15.1.14-1, the analysis for 4,100 MWt is limiting . Mus , Fig. 15.1.14-1 through 15.1.14-4 are not provided for B&W.

15.1.14.3 NSSS Modification (RL- 41 & 3S Only)

Table 15.4-21 of RESAR-41 and Table 15.4-7 of RESAR-3S include 28 equipment required for main steam line, feedwater line, and steam generator blowdown line breaks as follows:

1. Emergency boration system, including the pumps, valves, boron injection tank, and piping.

2. Main feedwater control valves (trip closed features) .

15.1-7 , ,3 . , Amendment 28 U/V U, > 31 8/6/76

SWESSAR-P1

3. Bypass f eedwater control valves (trip closed f eatures) .

equipment required to trip the main

4. Circuits and/or feedwater pumps.

5. Main feedwater isolation valves (trip closed f eature) .

6. Main steam line stop valves (trip closed f eature) .

7. Main steam line stop valve bypass valves (trip closed feature).

In the SWESSAR design, an additional feedwater isolation valve is by added, in series with the feedwater isolation Refer valve to specified Fig. 10.4.7-2A.

Westinghouse, in each feedwater line.

This arrangement has the following advantages:

a. Feedwater isolation is achieved by isolation valves without the necessity to use control valves in a dual f unction .

b. For each f eedwater line only two valves need to receive isolation signals instead of three.

c. The feedwater flow control valves and their bypass valves and associated piping can be moved to the turbine building, which is a more logical location. (~

d. The piping installation inside the annulus building is simplified.

e. The safety related wiring and circuitry for feedwater isolation are confined to the annulus building.

feedwater isolation valves are installed, the Since redundant requirement for tripping control a) the main feedwater control valves, b) valves, and c) the main feedwater the bypass feedwater pumps on a feedwater isolation signal is eliminated.

With the exception of the above three items, all the items listed in Table 15.4-21 of RESAR-41 and Table 15.4-7 of RESAR-3S 28 ,

meet the requirements of IEEE 279-1971 and IEEE 308-1971.

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15.1-8 Amendment 28 8/6/76

SWESSAR-P1 15.1.15 Inadvertent Loading and Operation of a Fuel Asegmbly i

Into an Improper Position This condition is within the NSSS Vendor's scope and SAR.

15.1.16 Process Gas Charcoal Bed Adsorber Line Rupture 15.1.16.1 Identification of Causes The radioactive gaseous waste system (Section 11.3) removes fission product gases from the reactor coolant on a continuous basis. The system consists of a degasifier to strip the reactor coolant letdown of noble gases, process gas refrigerant dryers and a process gas water separator to dehumidify the gas stream, and process gas charcoal bed adsorbers to selectively hold up radioactive noble gases to provide time for decay of the radioactive constituents. The process gas is at approximately 2 psig pressure leaving the degasifier. In addition, the process gas is dehumiditied bef ore entering the charcoal bed adsorbers.

Upon leaving the adsorber s , the gas is either compressed to 75 psig for recycle to the volume control tank or released to the ventilat: ;n vent.

15.1-9 Amendment 3 10/15/74

,c C l 'I,

()f(,i kIJ J

SNESSAR-P1 The maximum amount of stored waste gases is present in the charcoal bed adsorbers just prior to retueling.

(

The accident is defined as an unexpected and uncontrolled release of part of the radioactive xenon and krypt.on fission product gases stored in the process gas portion of the radioactive gaseous waste system, as a consequence of a failure of the first process gas charcoal bed adsorber or its associated piping. The released gases include all the gases in the associated pipinc and, of the gases stored in the charcoal bed adsorber tank, only the fractions of isotopes indicatec in Table 15.1.16-1.

15.1.16.2 Analysis of Ef fects and Consecuences All noncondensable gases are removed from the reactor coolant letdown stream in the degasifier. All the noble gases in the piping and the equipment between the degasifier and , but not including, the first process gas charcoal bed adsorber are a ssumed to be released without decay. In addi tion, part of the gas adsorbed en the charcoal is released as a consequence of the depressurization of the process gas portion of the radioactive gaseous waste system. The fractions released were calculated 1l using the model developed for the Fast Flux Test Facility *.

Table 15.1.16-1 lists the maximum radioisotope inventory in the process gas charcoal bed adsorbers and associated pipings. It also lists the release f ractions used in the analysis for the release of noble gases from the process gas charcoa) bed adsorber due to depressurization of the process gas portion of the g

M radioactive gaseous waste system.

B ased on the conservative assumptions stated above and the atmospheric dif f usion f actor (CHI /Q) given in Figure 1 with building wake correction f actor given in Figure 2 of Regulatory Guide 1.4, the gamma and beta doses as a function of distance are shown in Fig. 15.1.16-2.

The s.te-indepencent beta and garna doses vs CHI /O for the process gas churcoal decay bed and adsorber line rupture are presented in Fig. 15.1.16-1.

It is concluded that the doses associated with the unlikely event of the process gas charcoal deca,y bed adsorber line rupture would be a small fraction of the recommended limits of 25 Rem to the whole body given in 10CFR100.

No credible mechanism exists that would cause the complete I

release of the noble gas inventory stored in the process cas charcoal decay bed. Therefore the radiological consequences of such an incident are not calculated. a e

I -

• Underhill, D. W., "Effect of Rupture in a Pressurized Inble Gay Absorption Bed," Nuclear Sc.fety, Vol .13, No. 6, pp 4 78-4 t: 1. ,

('~

h 15.1-10 . Amendment 7

, . ( b e, 2/28/75 e,

SWESSAR-P1

% 15.1.17 Failure of Air Eiector Liner This accident applies only to BWRs and is therefore not applic-

'le to the PWR Standard Plant.

15.1.18 Steam Generator Tube Rupture Steam generator tube rupture is described in Chapter 15 of the NSSS Vendor's SAR.

15.1-10A Amendment 7 2/28/75 hla OJ

SWESSAR-P1 15.1.18.1 Identification of Causes 1

Causes of the above accident are delineated in Chapter 15 of the S AR provided by the NSSS Vendor.

15.1.18.2 Analvsis of Effects and Consecuences lionradiological consequences of the above accident are described fil in Chapter 15 of the NSSS Vendor's SAR.

A conservative analysis of the radiological consequences resulting from s steam generator tube rupture is presented  :

a ssuming loss of off site power.

The assumptions used in the analysis of the steam generator tube rupture are given in Table 15.1.18-2.

Total iodines and noble gases released during the accident for steam releases and primary-to-secondary leakage in steam generators are given in Table 15.1.18-1.

The site-independent thyroid, beta, and gamma doses as a function of CHI /Q for the st emu uenerator tube rupture are shown in Fig. 15.1.18-1.

The thyroid, gamma, and beta doses for the steam generator tube rupture at various distances from point of release based on R egulatory Guide 1. 4 (Section 3 A.1-1.4) meteorology are shown in Fig, 15.1.18-2 and 15.1.18-3.

15.1.19 Failure of Onarcoal or Cryogenic System (BWh)

This accident applies only to BWRs and is therefore not applic-aDie to the PWR Reference Plant.

15.1.20 Rod Eiection Accident The rod ejection accident is described in Chapter 15 of the NSSS Vendor's SAR.

15.1.20.1 Identification of Causes Causes of the above accident are delineated in Chapter 15 of the NSSS Vendor's SAR.

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15.1-11/12 Amendment 11 5/30/75

SWESSAR-P1 15.1.20.2 Analysis of Effects and Consequences 1

Nonradiological consequences of the above accident are described in Chapter 15 of the NSSS Vendor's SAR.

The radiological consequences of this accident are calculated according to the methods described in Regulatory Guide 1.77 g (Section 3 A .1-1.7 7) . The parameters used in this analysi.s are presented in Table 15.1.20-2.

Total iodines and noble gases released during the accident from the containment structure and the steam relief valves are given in Table 15.1.20-1.

The site--independent thyroid, beta, and gamma doses as a function of CHI /Q at the EB and LPZ are shown in Fig. 15.1.20-1 and 15.1.20-2.

Thyroid, ganma , and beta doses for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 day duration for the rod ejection accident at various distances from point of release are shown in Fig. 15.1.20-3 and 15.1.20-4 respectively.

15.1-13 6 14 Amendment 8 3/28/75 n,7 QQ UJi

SWESSAR-P1 b.1.21 The Soectrum of Rod Drop Accidents

condition applies only to BWRs and is the.refore not pplicable to t.ne PWR Reference Plant.

15.1.22 Break in Instrument Line or Other Lines f rom Reactor Coolant Pressure Bourdary that Penetrate Containment W is condition is within the NSSS Vendor's scope and SAR.

15.1.23 Fuel Irandling Accident 15.1.23.1 Identification of Causes The fuel handling accidents include:

1. A dropped spent fuel assembly on to the fuel pool floor (fuel building)

2. A dropped spent fuel assembly into the refueling cavity 33 (containment)

Fuel handling accidents are defined as the dropping of a spent fuel assembly on to the pool or cavity floor and the damaging of that spent fuel assembly, despite the many administrative controls and physical limitations imposed on fuel handling operations. All refueling operations are conducted in accordance with prescribed procedures.

15.1.23.2 Analysis of Effects and Consequences me fuel handling accidents are analyzed as follows:

Fuel Handling Accident in the Fuel Building All fuel handling operations are conducted underwater by the fuel handling machine. The fuel assembly is assumed to be dropped within the pool with resulting damage and fission product release. It is assumed that the gap activity from the fuel rods in the assembly is released to the fuel pocl. 33 A fraction of the activity is released from the water.

Activity released to the atmosphere inside the fuel building is collected by the fuel building ventilation system and processed by the SLCRS (Section 9.4.6) .

The air exhausted f rom the filters is assumed to be released at grotmd level.

Fuel Irandling Accident in the Containment The sequence of events for a fuel handling accident in the containmen t is identical to that of a fuel handling accident in the fuel building. All fuel handling operations are 15.1-15 .

<3 eo Amendment 33 U' u, OJd 6/30/77

SWESSAR-P1 conducted underwater. The fuel assembly is assumed to be dropped within the refueling cavity with resulting damage and fission product release. It is assumed that the gap activity from the damage 2 fuel rods in the assembly is released to the reactor cavity. A f raction of the activity is released from the water. Activity released to the containment atmosphere is detected by safety related, redundant containment purge

, isolation monitors located upstream of the containment J atrosphere filters. These monitors provide a signal to accomplish containment isolation prior to release of radioactivity f rom the containment (See Section 9.4.5.2.3.)

The radiological consequences of a fuel handling accident in the fuel building have been evaluated in accordance with the methods described in Regulatory Guide 1.25 (Section 3A.1-1.25) . The assumption used for this analysis is presented in Table 15.1.23-2. The releases to atmosphere are given in Table 15.1.23-1. The site-independent thyroid, beta, and gamma doses as a function of CHI /O for the fuel handilng accident are shown in Fig. 15.1.23-1.

As can be seen from the radionuclide releases in Table 15.1.23-1, the analysis for 4,100 MWt is limiting. Thus Fig. 15.1.23-1 and 15 .1.23-2 are not provided for B&W.

Thyroid, gamma, and beta doses are given in Fig. 15.1.23-2 at various distances from the plant.

15.1.24 Small Spills or Leaks of Radioactive Material Outside Containment 15.1.24.1 Identification or Causes During normal plant operation radioactive fluids are contained in the reactor coolant system, chemical and volume control system, boron recovery system, and the radioactive waste systems. These systems are connected to the reactor plant vent and drain system (Section 9.3.3) to minimize any uncontrolled leakage. Therefore, the possibility of a radioactive leak exists from:

1. An a bnormal discharge from a valve, pump, etc in these systems, and

2. Leakage from auxiliary systems serving the above systems in which primary-to-secondary leakage has occurred.

15.1.24.2 Analysis of Ef fects and Consecuences Any leakage in the various buildings is directed toward the building sumps. The sumps collect and transfer the liquid to the 9

15.1-16 Amendment 35 10/6/77 b7 d [$7

SWESSAR-P1 high level or low level waste tanks for further processing by the liquid waste system (Section 11.2) .

Leakage of radioactive liquids from outdoor tanks will be col-lected in concrete Seismic Category I dikes.

15.1.25 Puel Cladding Failure Combined with Steam Generator Leak This condition is within the NSSS Vendor's scope and SAR.

15.1.26 Control Room Uninhabitability 15.1.26.1 Identification of Causes Tae control room is designed for safe continued occupancy during accident conditions. Section 12.1.2.10 discusses the shielding design which limits control room radiation levels during a DBA.

Control room operator doses associated with a DBA are given in Section 6.4. Section 9.4.1 describes the air conditioning, ventilation, and emergency air supply systems for the control room. Section 6.4 specifically addresses the habitability of the control room.

The contral room may becone uninhabitable in the event of a fire.

The potential severity of a fire in the control room is limited since the only flammable materials inside the control roon are logs, procedures, manuals, etc. A postulated fire is of such a small magnitude that it can be extinguished by using a hand fire extinguisher provided in the control room as a part of the fire protection system (Section 9.5.1) .

The resulting smoke is removed by the ventilation system.

However, for the purposes of this analysis it is assumed that the control room must be abandoned due to smoke.

15.1.26.2 Analysis of Effects and Consequences If the control room becomes uninhabitable, operators can maintain the reactor in the hot standby condition from the auxiliary shut-down panels in the emergency switchgear room. This is in compliance with AEC General Design Criterion 19 (Section 3.1.19) .

Provisions are also available for going to the cold shutdown condition from outside the control room.

The nuclear auxiliary shutdown panel contains instrumentation and controls to perform the functions described'in Section 7.4.3.

15.1.27 Failure or Overnressurization of Low Pressure Residual Heat Removal System This condition is within the NSSS Vendor's scope and SAR.

040 O ! '-r"U 15.1-17 ')

SWESSAR-P1 15.1.28 Loss of Condenser vacuum 15.1.28.1 Identification of Causes Loss of condenser vacuum could result from the following:

1. Failure of the condenser evacuation system.

2. Failure of the circulating water system.

15.1.28.2 Analysis of Effects and Consequences High condenser pressure initiates a turbine trip which, in turn, causes high main steam pressure, resulting in a ctuation of the main steam atmospheric dump and main steam safety valves. This results in the relieving of excess pressure and allows a safe shutdown of the plant. The plant can be maintained in the hot shutdown condition by use of the main steam atmospheric dump valves. The radiological consequences of this accident have as an upper bound the radiological consequences evaluated for a steam pipe rupture (Section 15.1.14).

15.1.29 Turbine Trip with Coincident Failure of Turbine Bypass Valves to Open 15.1.29.1 Identification of Causes Turbine trip could result from numerous causes such as: over-speed, low vacuum, thrust bearing wear, master trip solenoid C

a ction , low bearing oil pressure, low hydraulic oil pressure, high vibration, or by manual operator action.

Failure of the turbine bypass control valves to open could result from loss of instrument air or from control system failure.

15.1.29.2 Analysis of Effects and Consegueices Should a turbine trip occur coincident with a failure of the turbine bypass control valves to open, the pressure in the main steam system rises. This results in the actuation of .rst, the main steam atmospheric dump valves and then, the . min steam safety valves; thereby relieving the excessive main steam pressure created and allowing a safe shutdown of the plant. The plant can be maintained in the hot shutdown condition or cooled down to the point of changeover to the residual heat removal system by using the main steam atmospheric dump valves. The radiological consequences of this accident have as an upper bound the radiological consequences evaluated for a steam pipe rupture (Section 15.1.14) .

O 15.1-18 0/'-

SWESSAR-P1 15.1.30 Loss of Reactor Plant Service Water System 5

15.1.30.1 Identification of Causes Loss of all reactor plant service water is an extremely unlikely occurrence. Significant redundancy is built into the design of the reactor plant service water system (including the ultimate heat sink) and the electric power (the required onsite, offsite, and emergency a-c power systems) sources to operate the reactor plant service water system. Because of this redundancy, no single credible failure can reduce the reactor plant service water system capability below its minimum requirements.

The worst aypothetical situation that could occur would be a loss of all a-c power. This would shut down all the service water pumps as well as all other a-c motor driven components.

15.1.30.2 Analysis of Effects and Consequences The analysis of this worst case situation is presented in Section 15.1.9, Loss of All A-c Power to the Station Auxiliaries (Station Blackout). The conclusions reached are that the plant is safely shut down and no radiological hazard to plant personnel or to the public occurs since only secondary steam is discharced to the atmosphere.

15.1.31 Loss of One D-c System 15.1.31.1 Identification of Causes The d-c onsite power system is composed of four completely in-dep,endent, Class IE, d -c power supplie s . A d-c supply is connected to each diesel generator independently.

The loss of a single d-c system may be caused by:

1. Malfunction of the charger which forces the battery to take the full d-c load. This actuates an alarm in the control room to initiate appropriate corrective action.

2. Station batteries may deteriorate with time, but pre-cipitous failure is extremely unlikely. The surveillance specified is that which has been demonstrated by experience to provide an indication of a cell becoming unserviceable long before it fails.

Variour tests are set up to detect a failure. The specific gravity, electrolyte temperature, and cell voltage of each of the 60 cells are measured and recorded on a weekly and monthly basis. Twice a year, during normal operation, + 5e battery charger is turned off for approximately 3 min, and the battery voltage and current are recorded at the beginning and end of the test. Once a year during the normal refueling shutdown, 15.1-19 IL'

SWESSAR-P1 each battery is subjected to a simulated load test without the battery charger. '

Alarms are provided to indicate low battery voltage and low current from the chargers which make it extremely unlikely that deterioration will go undetected (see Section 16. 4. 6) .

3. A bus fault at the d-c distribution panel is a single event that can cause a complete loss of one of the four d-c power systems. Such a loss actuates an alarm, and a loss of voltage is indicated in the control room.

15.1.31.2 Analysis of Effects and Consequences For each of the causes above the following corrective action can be taken:

1. If the d-c system loss is due to a malfunctioning batterf charger, a spare battery charger is available and is substituted for the faulty one by merely opening and closing feeder circuit breakers as indicated on Fig.

8.3.1-2.

2. If the d-c system is lost because of a faulty battery, the faulty battery can be disconnected manually and repaired while the vital bus inverter and the d-c f<

distribution panel are fed from the battery t targer. (~

3. In the event a bus fault occurs at the d-c distribution panel, and one d -c system is lost completely, its associated vital bus would also be lost until it is switched manually to be fed from one of the four regulating transformers. The loss of any of three station d-c power supplies affects the respective diesel generator and the associated emergency bus.

Loss of one d-c s' apply introduces the loss of a-c and d-c power of one of the four protective channels of the reactor protection system (RPS) and one of three engineered safety features actuation systems (ESFAS) . This results in the trip cr actuation signal logic being diminished by one protective channel. The detailed effect of loss of power to the RPS and the ESFAS is described in Section 7.3.

In the event of the complete loss of one d-c system, one vital 120 V a-c bus and control of one section of switchgear are lost.

This condition is immediately annunciated in the control room and corrective action is undertaken. During the loss of the d-c system, two separate and redundant sections of emergency switchgear fed from two independent d-c systems are available for shutdown, since each section of emergency switchgear is fed from  ; .

a different d-c power supply. ~

.r. ne7 O!U USJ 15.1-20

SWESSAR-P1 g Because of the redundancy built into the d-c power systems, the i loss of up to one of the four complete d-c supplies can be tolerated by the plant's design without any degradation of plant safety. No release of activity from the plant is expected as a result of this postulated event.

Reference Sections Title No.

A-c Onsite Electric Power System 8.3.1 D-c Power System 8.3.2 15.1.32 Inadvertent Operation of ECCS During Power Operation This condition is within the NSSS Vendor's scope and SAR.

15.1.33 Turbine Trio With Failure of Generator Breaker 15.1.33.1 Identification of Causes Electric energy generated at J5 kV is raised to system voltage by the main transformer and delivered to the existing systems as shown in Fig. 8.3-1.

Station service transformers connected to the generatcr isolated phase bus leads from the main generator normally supply power to the plant auxiliaries at 13,800 V. Reserve station service power, for startup and emergency use, for plant auxiliaries is supplied from the preferred 69 kV source.

On a turbine trip, the generator breaker is interlocked to open via the anti-motoring circuit. Should the breaker malfunction and fail to open on the turbine trip, a breaker failure relay senses the malfunction and trips the adjacent switchyard breaker in the next protection zone. This action effectively isolates the plant from the system.

15.1.33.2 Analysis of Effects and Consecuences In the event of failure of the generator breaker on a turbine trip, the breaker failure relay associated with the generator breaker senses the malfunction and trips the ad jacent switchyard breakers in the next protection zone. These breaker actions leave the generator electrically isolated from its load and the rest of the system. However, the system loa.ds are not affected by the isolation of this particular generator, since they are supplied by the remaining power system without any adverse effect to system stability, as discussed in Section 8.2.2.

Reserve station service power is supplied from the preferred 69 kV source. Consequently, full offsite power is available after isolation of the generator breaker senses the malfunction 15.1-21 670 ~ n4'

SWESSAP-P1 and trips the adjacent switchyard breakers in the next protection [. . -

b one.

Peference Sections Title No.

Offsite Power System 8.2 A-c Power System 8.3.1 15.1.34 Loss of Instrument Air System 15.1.34.1 Identification of Causes Locs of instrument air could result f rom the following:

1. Los; of station oower to comore3 sors.

2. Mechanical or electrical failure of all air compressors.

3. Pupture of .ne instrument air header.

15.1.34.2 Analysis of Ef f ects and Consecuences The instrument air system is not safety related. The description of the instrument air system is aiven in Section 9.3.1. A discussion of the failure causes and their effects follows:

1. Loss of electric power to the compressors - Safety related air-operated valves are spring-loaded and "f ail-safe" upon loss of control air.

2. Mechanical or electric failure of air compressors -

Safety related air-operated valves are spring-loaded and

" fail-safe" upon loss of control air.

3. Rupture of the instrument air header -

Should an Latrument air header fail, the air-operated valves that are sa:fety related " fail safe."

On loss of instrument air, the reactor and turbine trip as described in Section 15.1.8, " Loss of Normal Feedwater."

As previously stated, the instrument air system is not safety related and its failure does not affect plant safety.

All the a bove ef f ect s and consequences are the same for the containment instrument air system.

B 4 c, .-

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15.1-22 Amendment 8 3/28/75

SWESSAR-P1 15.1.35A Malfunction of Turbine Gland Sealina for General Electric Turbine 3 15.1.35A.1 Identification of Causes The turbine gland sealing system is not safety related. A description of this system is given in Section 10.4.3.

Malfunction of the turbine gland sealing system can result from the following:

1. Failure of the steam seal feed valve

2. Failure of the steam packina unloading valve

3. Failure of the cland steam condenser exhauster 15.1.35A.2 Analysis of Effects and consequences a

1. The steam seal feed valve may stick in the open position. When the flow into the steam seal hoader begins to exceed that needed by the turbine seals, the automatic steam packing unloading valve opens , divertina the excess steam to the condenser. A pressure transmitter and motor operated bypass and shutoff valves permit manual operation.

2. The steam packing unloadina valve may fail in the closed position. If the failure happens at low loads, the glands continue to function properly. At higher load s, more steam is leaking from the pressure packings than is r3 quired by the vauum packings and the pressure in the steam seal header rises. A relief valve in the steam seal header protects aga inst overpressure. This condition causes steam leakace to the atmosphere. A pressure transmitter and motor operated unloadina valve allow manual operation.

3. Two full capacity cland steam condenser exhausters are provided. Upon failure of the operating exhauster, the standby exhauster can be utilized. Motor operated discharge valves isolate the exhausters.

The turbine gland sealing system is a reliable nonsafety related turbine plant system. The loss of the turbine seals causes the turbine to trip due to loss of vacuum. A full or partial loss of the turbine gland sealing system could also result in steam leakage from the turbine. Even with plant operation at the maximum allowable reactor coolunt activity and with steam cenerator tube leakace, the potential release of activity from this postulated accf. dent is significantly lower than suggested 10CFR100 limits.

15.1-23 Amendment 8 3/28/75 U/U Od6

SWESSAR-P1 15.1.35B Malfunction of Turbine Gland Sealina System f or Westinghouse Turbine -.

8 15.1.35B.1 Identification of Causes The turbine gland sealing system is not safety related.- 7.

description of this system is given in Section 10.4.3.

Malf un ction of the gland sealing system can result f rom the following:

1. Failure of the main steam seal supply valve

2. Failure of the high pressure spillover valve

3. Failure of the cland condesner exhauster 8 15.1.35B.2 Analvsis of Effects and Consequences

1. The main steam seal suoply valve may fail in the onen position. When the flow into the steam seal header excessively exceeds that needed by the turbina seals, a safety valve opens, causing steam to be released to atmosphere. A pressure transmitter and motor operated bypass and shutoff valves permit manual or remote correction of this situation. "

2. The high pressure spillover valve may fail in the closed position. If the failure happens at low loads, the glands continue to f unction properly. At higher loads, the glands will leak to atnosphere. A safety valve in the steam seal header protects against overpressure and releases steam to atmosphere. A pressure transmitter and motor operated shutoff and bypass valves allow manual or remote operation in the event of failure.

3. Two full sized gland condensar exhausters are provided.

Upon f ailure of the operatina exhauster, the standby exhauster can be utilized. Motor operated discharce valvec isolate the exhausters.

The turbine gland sealina system is a reliable nonsafety related turbine plant system. The less cf the turbine seals causes the turbine to trip due to loss of vacuum. A full or partial loss of the turbine gland seal system could also result in steam leakage from the turbine. Even with olant operation at the maximum allowable reactor coolant activity and with steam generator tube leakage, the potential release of activity from this postulated accident is significantly lower than suggested 10CFR100 linits.

6, / v t'

7 p<r'f J

15.1-24 Amendment 8 3/28/75

SWESSAR-P1 1

Table 15.1-1 is deleted 7 ou 0 '! 8 1 of 1 Amendment 7 2/28/75

SWESSAR-P1 q TABLE 15.1-2 IODINE CONCENTRATION IN STEAM GENERATOR LIQUID WITH 1,000 GPD PRIMARY-TO-SECONDARY LEAKAGE AND 1 PERCENT FAILED FUEL Concentration, uCi/qm Isotope B6W C-E W-41 W-3S I-131 Not 4.9 E-0 3 6.3 E-0 3 8.0 E-03 I-132 applicable 1.1 E-03 1.2 E-03 2.3 E-03 H I-133 7.3 E-03 9.5 E-03 1.1 E-02 I-13:4 4.4 E-04 4.3 E-04 5.0 E-0 4 I-135 3.5 E-03 4.4 E-03 5.1 E-03 b/U 0;9 1 of 1 Amendment 17 9/30/75

SWESSAR-P1 1

Table 15.1-3 is deleted i 1 of 1 Amendment 7 b[, gr, 2/M/M

SWESSAR-P1 TABLE 15.1.13-1 REACTOR CONTAINMENT N13CLIDE IINENTORY AVAILABLE FOR LEAKAGE IMMEDIATELY AFTER LOCA Inventory, Ci Isotopes 3,636 MWt 3,876 MWt 4,100 MWt I-131 2.3 E 07 2.5 E 07 2.6 E 07 I-132 3.3 E 07 3.5 E 07 3.7 E 07 I-133 5.0 E 07 5.3 E 07 5.8 E 07 I-134 6.0 E 07 6.6 E 07 6.7 E 07 I-135 4.8 E 07 5,1 E 07 5.3 E 07 Kr-83m 1.6 E 07 1.7 E 07 1.8 E 07 Kr-85m 4.0 E 07 4.3 E 07 4.5 E 07 Kr-85 8.4 E 05 9.0 E 05 9.4 E 05 is Kr-87 7.1 E 07 7.6 E 07 8.7 E 07 Kr-88 1.1 E 08 1.2 E 08 '

1.2 E 08 Kr-89 1.4 E 08 1.5 E 08 1.6 E 08 Xe-131m 8.0 E 04 8.5 E 04 9.0 E 04 Xe-133m 4.9 E 06 5.2 E 06 5.5 E 06 Xe-133 2.0 E 08 2.1 E 08 2.3 E 08 Xe-135m 5.5 E 07 5.9 E 07 6.2 E 07 Xe-135 4.1 E 07 4.4 E 07 2.9 E 07 Xe-137 1.8 E 08 1.9 E 08 2.1 E 08 Xe-138 1.8 E 08 1.9 E OP 2.0 E 08

/ 7 ,' flr 1

• 'J l 1 of 1 Amendment 19 12/12/75

SWESSAR-P1 TABLE 15.1.13-2 1

iwRAMETERS USED FG1 THE LOSS OF COOLANT ACCIDENT ANALYSIS *

1. Before the incident, the reactor was operating at 4,100 MWt ( C-E, W-41) 3,876 MWt (B&W) , or 3,636 MWt g (W-3S). g

2. Twenty-f ive percent of the equilibrium radioactive iodine inventory developed from maximum f ull power operation of the core is immediately available for leakage from the containment structure. Ninety-one percent of this 25 percent is in the form of elemental iodine , 5 percent in the form of particulate iodine, and 4 percent in the form of organic iodides.

3. One-hundred percent of the equilibrium radioactive noble gas inventory developed from maximum full power operation of the core is immediately available for leakage from the containment structure.

4. The iodine removal coef ficient by U.e cantainment spray system for elemental iodine is 10 hc-1 until elemental iodine has been reduced to 1 ,ercent of original inventory, then no credit is ttKen for additional reduction in elemental iodine. No credit is taken for reduction in particulate and organic iodine by the containment spray system.

5. The containment structure leaks at a rate of 0.2 percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1 percent per day for the duration of the accident.

6. The atmospheric diffusion factor (CHI /Q) including the volumetric building wake correction factor is based on values givt, in Section 2.3.

7. No credit is taken for depletion of the ef fluent plume of radioactive isotopes due to deposition on the ground or for the radioactive decay of iodine in transit.

8. Fifty percent of containment leakage is collected in the supplementary leak collection and release system (Section 6. 2. 3.1) and filtered through high efficency particulate air (HEPA) filters / charcoal absorbers with an over all efficiency for iodine of 95 percent before release to the environment.

9. One hundred percent of the containment leakage is released unfiltered until the cupplementary leak collection and release system becomes ef fective (Section

6. 2.3.1) .

• Based on assumptions outlined in Regulatory Guide 1. C 7 ' ' O"7 0tu aL 1 of 1 Amendment 19 1z/12/75

A ShESSAR-P1 TAhLE 15.1.14-1 SEAM PIPE BREAK ACCIDENT I<ELEASEE Tot al F ele.a se (Ci1 B&W C-E w_41 w_?;

Isotope 0-2 hr 0-8 hr 0-2 hr 0-8 br 0-2 hr 0-H hr 0-2 hr 0-8 hr I-131 5.8 E 00 hot 7.9 E-01 1.3 E 00 4.2 E 01 1. 3 E 0 2 1.8 E 01 9.3 E 01 I-132 1.8 E 00 Applicable 2.3 E-01 3.8 E-01 7.1 E 00 2.1 E 01 2.9 E 00 1.1 E 01 I-133 7.8 E 00 1.2 E 00 1.9 E 00 3.2 E 01 8.5 E 01 1.4 E 01 6.6 E 01 1 -134 8.4 E-01 1.3 E-01 2.2 E-01 6.7 E 00 7.1 E 00 2.7 E 00 1.0 E 01 1-135 3.6 E 00 6.2 E-01 1.0 E 00 1.6 E 01 3.5 E.01 6.8 E 00 3.1 E 01 Kr -8 3m 6.7 E-01 1.4 E-01 2.8 E-01 2.7 E 00 1.1 E 01 1.2 E 00 4.6 E 00 Kr-u5m 2.7 E 00 5.5 E-01 1.1 E 00 9.7 E 00 3.9 E 01 4.2 E 00 1.7 E 01 Kr-85 7.9 E-02 1.1 E-02 2. 2 E-02 2.0 E 01 8.1 E 01 1.2 E 01 4.7 E 01 Kr-87 1.H E 00 3.9 E-01 7.8 E-01 1.0 E 01 4.0 E 01 4.2 E 00 1.7 E 01 11 Kr-88 5.2 E 00 1.1 E 00 2.1 E 00 2.1 E 01 8.5 E 01 9.1 E 00 3.6 E 01 Kr-89 1.5 E 01 3.4 E-02 6.8 E-02 3.7 E 00 1.5 E 01 1.5 E 00 5.9 E 00 Xe-131m 2.5 E-02 3.5 E-03 7.0 E-0 3 1.2 E 00 4.9 E 00 5.1 E-01 2.0 E 00 Xe-133m 1.3 E 00 1. 9 E-01 3.9 E-01 4.4 E 00 1.8 E 01 1.9 E 00 7.5 L 00 Xe-133 5.7 E 01 8.3 E CO 1.7 E 01 2.6 E 02 1.0 E 03 1.1 E 02 4.4 E 02

""[* y ,

Xe-135m 1.b E 00 3.7 E -01 7.3 E-01 3.5 2.0 E

E 00 01 1.4 8.0 E 01 E 01 1.6 8.8 E 00 E 00 6.4 3.5 E 00 E 01 Xe-135 4.7 E 00 9.0 E-01 1.8 E 00 4'

-- g Xe-137 2.3 E-01 5.3 E-0 2 1.1 E-01 5.5 E-0 2 2.2 E-01 5.0 E-02 2.0 E-01 (a: .)

1 xe-138 8.4 E-01 1.9 E-01 3.8 E-01 1.1 E 01 4.4 E 01 4.5 E 00 1.8 E 01 pu J

( )

a. .

L  ::.3

g. 1_.)

U a

" :2 sco _ man

$ *D CIT

~%f C

CD

(:

(rJ 1 of 1 Peendaent 19 12/12/75

SWESSAR-P1 TABLE 15.1.14 -2 i

PARN4ETERS USED FOR T!!E STEAM PIPE BREAK ACCIDENT ANALYSIS

1. Prior to the accident, core themal power is 4,100 MWt (C-E , W-41) , 3,876 MWt (BGW) , or 3,636 MWt (W-3S). lg

2. Steam and power conversion system equilibrium activity prior to the accident is based on primary-to-seconda ry leakage of 1,000 gpd in the steam generators and is given in Table 15.1-2 P441, W-3S, C-E). g

3. Reactor coolant activity is based on operation with cladding defects in rods producing 1 percent of core power as given in Section 11.1.

4. The primary-to-secondary leakage is evenly distributed among the steam generators (W-41, W-3S , C-E). 19

5. Offsite power is lost, thua making the condensers unavailable for turbine bypass.

6. No steam jet air ejector release and no steam generator blowdown occur during the accident.

7. No noble gas is dissolved in the steam generator water.

8. The entire contents of the defective steam generator prior to the accident are released instantaneously at the time of the accident.

9. Prima ry-to-secondary leakage continues evenly distributed in the steam generators at the rate of 1,000 gpd after the initial steam release (W-41, W-3S, g C-E).

10. The iod ine partition factor in the steam generators before the accident is amount of iodine / unit mass of stear =

1.00 (BSW) amount of iodine / unit mass of water 0.01 (C-E) 0.01 (W-41) 0.01 (W-3S)

31. The iodine partition factor in the defective steam generator for the initial steam release and for continting leakage into that steam generator is amount of iodine / unit mass of steam = 1.0 amount or iodine / unit mass of water

12. The iodine partition f actor in,t,h,e ste,am , generator (s) ho UDz 1 of 4 Amendment 19 12/12/75

SWESSAR-P1 TABLE 15.1.14 -2 (CONr) 1 af ter the accident is amount of iaJine/ unit mass steam = 1.0 (B&W) amount of iodine / unit mass liquid 0.1 (C-E) 0.1 (W-41) 0.1 (W-3S)

13. During the occident, iodine carryover from the primary side in the nondef ective steam generator (s) is diluted in the incoming f eedwater.

14. Atmospheric dif fusion, including the volumetric building wake correction factor, is based on values given in Section 2.3.

15. No correction is made for depletion of the ef fluent clume of radioactive isotopes due to deposition on the ground or for the radioactive decay in transit.

16. The following assumptions and/or data are taken from Sections 15.1.7 and 15.4.2 of RESAR-41 and restated here for completeness:

a. Within the initial 30 min, 144,000 lb of steam and water are released from tha defective steam gen erator.

b. Steam released from three nondefective steam generators for the first 2 hr is 527,000 lb and thereafter (2-8 hr) is 1,330,000 lb.

c. Feedwater flow to three nondefective steam generators for the first 2 hr is 715,000 lb and the rea f ter (2-8 hr) is 1,420,000 lb.

d. Tha primary pressure remains constant at 2,235 psig for 0-2 hr and decreases linearly to atmospheric during the period of 2-8 hr.

e. Five percent of the total core fuel cladding is damaged which results in the release to the reactor coolant of 5 percent of the total gap inventory.

17. The f ollowing assumptions and/or data are taken from Section 15.1.14 of B-SAR 205 and restated here for complete ness .

19

a. Steam generator inventory = 49,940 lb.

/, ' ' O ;F G u/u L J 2 of.4 Amendment 19 12/12/75

SWESSAR-P1 TABLE 15.1.14 -2 (CO:E) i

b. Feedwater flow through feedwater control valves on j af f ected steam generator before closure = 30,240 '

19 I

lb.

c. Available mass in feedwater line between feedwater control valves and defective steam generator =

40,000 lb. 19

d. Mama release from nondefective steam generator "'

bef ore steamline isolation = 54,937 lb.

e. Steam released to atmosphere from defective steam generator = 115,830 lb. 19

f. Primary-to- secondary leakage during reactor coolant sys tem depr essurization = 3,000 lb.

18. The following ass uptions and/or data are taken from Sections 15.1.4 and 15.4.5 of CESSAR and restated here for completeness:

a. The water rass per steam generator is 152,000 lb.

b. Atmospaeric steam rt;3ase occurs until the reactor coolant temperature is reduced to 350 F, the point wb':re shutdown cooling can be initiated.

c. The maximum radioactivity release with loss of a-c power occurs with a break of the main steam pipe outside of containment and upstream of the main steam isolation valves.

d. Tha secondary system mass discharged from the defective steam generator is 265,300 lb in 800 sec; from the t'ndefective steam generator, 32,500 lb in the firat 800 sec and 151,000 lb thereafter.

19. The following assumptions and/or data are taken from Sections 15.1.7 and 15.4.2 of RESAR-3S and restated here for colupleteness:

a. Within the initial 30 min, 96,200 lb of steam and water are released from the def er+ ' ve steam genera tor .

b. Steam released from three nonde:: active steam generators for the first 2 hr is 479,000 lb and the reaf ter (2 -e hr) is 1,028,000 lb.

b/U pc UJ U 3 of 4 Amendment 19 12/12/75

SWESSAR-P 1 TABLE 15.1.14-2 (Cot 7r) 4

c. Feedwater flow to three nondefective steam generators for the first 2 hr is 550,000 lb and thereafter (2-8 br) is 1,100,000 lb.

d. Five percent of the total core fuel cladding is damaged which results in the release to the reactor coolant of 5 percent of the total gap inventory.

b [l} O"7 us/

4 of 4 Amendment 19 12/12/75

SWESSAR-P1 TABLE 15.1.16-1 MAXIMUM RADIOISOTOPE RELEASES FROM TIE PROCESS GAS CHARCOAL BED ADSORBER AND ASSOCIATED PIPING Inventory in Fraction 'Ib tal Released Release Isotope Piping (Ci) Charcoal Bed (Ci) from Bed (Ci)

Kr-83m 1.13 E 01 2.73 E 01 0.956 3.74 E 01 Kr-85m 4.17 E 01 2.38 E 02 0.774 2.26 E 02

?.88 E 02 0.0196 6.38 E 00 Kr -8 5 7.36 E-01 8.39 E 01 Kr-87 3.21 E 01 5.24 E 01 0.989 8.37 E 01 3.03 E 02 0.880 3.50 E 02 Kr-88 2.95 E 00 Kr-89 2.76 E 00 1.89 E41 1.00 Xe-131m 2.39 E-01 9.05 E 01 0.0207 2.11 E 00 II Xe-133m 1.34 E 01 9.42 E 02 0.103 1.10 E 02 Xe-133 5.66 E 02 9.39 E 04 0.0464 4.92 E 03 Xe-135m 2.98 E 01 1.01 E 01 1.00 3.99 E 01 Xe-135 1.00 E 02 1.20 E 03 0.484 6.81 E 02 Xe-137 4.45 E 00 3.68 E-01 1.00 4.82 E 00 Xe-138 1.83 E 01 5.63 E 00 1.00 2.39 E 01 I-131 5.66 E-03 1.42 E 00 ------ 5.66 E-03 1-132 2.24 E-03 6.62 E-03 ------ 2.24 E-03 I-133 9.43 E-03 2. 54 E-01 ------ 9.4 3 E--0 3 I-134 1.48 E-03 1.68 E-03 - 1.48 E-03 I-135 5.34 E-03 4 . 6 2 E-02 ------ 5.34 E-03 W 1 of 1 <7 ro Amendment 17 U!qU pdJu 9/30/75

SWESSAR-P1 TABLE 15.1.16-1 MAXIMUM RADIOISOTOPE RELEASES FROM THE PROCESS GAS CHARCDAL BED ADSORBER AND ASSOCIATED PIPING Inventory in Fraction Total Released Release Isotope Piping (Ci) Charcoal Bed (Ci) from Bed (Ci)

Kr-83m 1.1E 01 2.0E 01 0.9851 3.1E 01 Kr-85m 4.3E 01 1.8E 02 0.8661 2.0E 02 Kr-85 8.8 E-01 3.5E 02 0.0196 7.7E 00 Kr-87 3.0E 01 3.7E 01 0.9979 6.7E 01 Kr-88 8.3E 01 2.3E 02 0.9428 3.0E 02 Kr-89 2.6E 00 1.3E-01 1.0000 2.7E 00 g Xe-131m 2.8E-01 7.9E 01 0.0276 2.5E 00 Xe-133m 1.5E 01 8.1E 02 0.1361 1.3E 02 Xe-133 6.6E 02 8.2E 04 0.0618 5.8E 03 Xe-135m 2.8E 01 7.2E 00 1.0000 3.6E 01 Xe-135 1.0E 02 9.0E 02 0.5891 6.3E 02 Xe-137 4.1E 00 2.5E-01 1.0000 4.3E 00 Xe -138 1.5E 01 3.4E 00 1.0000 1.8E 01 I-131 6.7E-03 1.3E 00 --

6.7E-03 I-132 2.3E-03 5.1E-03 -- 2.3E-03 I-133 1.0E-0 2 2.1E-01 --

1. 0E-0 2 I-134 1.4E-03 1.2E-03 --

1.4E-03 1-135 5.5E-03 3.6E-02 --

5.5E-03 W-3S 1 of 1 Amendment 17 9/30/75 6/0 059

SWESSAR-P1 TABLE 15.1.16-1 MAXIMUM RADIOISOTOPE RELEASES FROM THE PROCESS GAS CHARCOAL BED ADSORBER AND ASSOCIATED PIPING Inventory in Fraction Total Released Release Isotope Piping . (Ci) Charcoal Bed (Ci) from Bed (Ci)

Kr-83m 6.74 00 1.40 01 .998 2.07 01 Kr-85m 2.78 01 1.36 02 .950 1.57 02 1 Kr-85 8.51-01 5.75 02 .0197 1.22 01 Kr-87 1.81 01 2.54 01 1.0 4.35 01 Kr-88 5.23 01 1.62 02 .986 2.12 02 Kr-89 1.44 00 8.78-02 1.0 1.57 00 Xe-131m 2.71-01 8.66 01 .0408 3.86 00 Xe-133m 1.36 01 8.21 02 0.197 1.75 02 Xe-133 6.18 02 8.80 04 .0907 8.60 03 Xe-135m 1.63 01 4.77 00 1.0 2.11 01 Xe-135 5.14 01 5.29 02 .735 4.40 02 Xe-137 2.33 00 1.65-01 1.0 2.50 00 Xe-138 8.44 00 2.22 00 1.0 1.07 01 I-131 6.32-03 1.36 00 6.32-03 I-132 1.62-03 4.12-03 1.62-03 I-133 8.16-03 1.89-01 8.16-03 I-134 8.39-04 8.14-04 8.39-04 I-135 3.74-03 2.78-02 3.74-03 BSW 1 of 1 Amendment 1

^

7/'30/74

.,r

() j ,j orU Uu

SWESSAR-P1 TABLE 15.1.16-1 MAXIMUM RADIOISOTOPE RELEASES FROM THE PROCESS GAS CHARCOAL EED ADSORBER AND ASSOCIATED PIPItG Fra ction Tbtal Inventory in Released Release Isotope Piping (Ci) Charcoal Bed (Ci) from Eed (Ci)

Kr-83m 6.68 E 00 2.34 E 01 0.9760 2.95 E 01 Kr-85m 2.56 E 01 2.11 E 02 0.8316 2.01 E 02 Kr-85 5.06 E-01 3.43 E 02 0.01976 7.28 E 00 Kr-87 1.84 E 01 4.33 E 01 0.9958 6.15 E 01 Kr-88 5.04 E 01 2.63 E 02 0.9211 2.93 E 02 3 Kr-89 1.59 E 00 1.58 E-01 1.0 1.75 E 00 Xe-131m 1.65 E-01 9.00 E 01 0.02471 2.39 E 00 Xe-133m 9.07 E 00 9.20 E 02 0.1223 1.22 E 02 Xe-133 3.89 E 02 9.30 E 04 0.0526 5.28 E 03 Xe-135m 1.72 E 01 8.43 E 00 1.0 2.56 E 01 Xe-135 4.21 E 01 7.30 E 02 0.5471 4.41 E 02 Xe-137 2.49 E 00 2.97 E-01 1.0 2.79 E 00 Xe-138 8.92 E 00 3.95 E 00 1.0 1.29 E 01 I-131 3. 9 2 E-0 3 1.42 E 00 -

3.92 E-03 1-132 1.38 E-03 5.89 E-03 -

1.3 8 E-03 I-133 6.17 E-03 2.40 E-01 -

6.17 E-03 1-134 8.66 E-04 1.41 E-03 -

8.66 E-04 I-135 3.31 E-03 4.14 E-02 -

3.31 E-03 C-E 1 of 1 g j !j' O b i. Amen dment 3 10/15/74

SWESSAR-P1 TABit 15.1.18-1 STEAM GENER ATOR TUBE RUi"MTRE RELEASES 7btal Release (Cl)

Isot<4w? B&W C-E W-41 W -3S 0-2 hr 0-8 hr 0-2 hr 0-8 hr 0-2 hr 0-8 hr 0-2 hr 0-8 hr I-1?1 3.1 E 02 Not 6.0 E 00 6.3 E 00 1.7 E 00 2.4 E 00 3.9 E 00 5.3 E 00 I-132 8.4 E 01 Applicable 1.7 E 00 1.8 E 00 6.5 E-01 8.5 E-01 1.2 E 00 1.3 E 00 I-133 4.2 E 02 -

9.2 E 00 9.6 E 00 2.8 E 00 3.7 E 00 6.1 E 00 7.8 E 00 I-134 4.5 E 01 -

7.9 E-01 7.9 E-01 3.2 E-01 3.3 E-OS 6.2 E-01 6.3 E-01 1-135 2.0 E 02 -

4.7 E 00 4.9 E 00 1.5 E 00 1.9 E 00 3.1 E 00 3.1 E 00 Kr-83m 3.6 E 01 - 1.7 E 01 1.7 E 01 2.5 E 01 2.5 E 01 2.5 E 01 2.5 E 01 Kr-85m 1.5 E 02 -

6.3 E 01 6.4 E 01 9.1 E 01 9.1 E 01 9.6 E 01 9.8 E 01 Kr-85 4.2 E 00 -

1.3 E 00 1.3 E 00 1.6 E 00 1.6 E 00 2.0 E 00 2.0 E 00 Kr-87 9.7 E 01 -

4.6 E 01 4.6 E 01 '.0 E 01 7.0 E 01 6.8 E 01 7.0 E 01 ig Kr-88 2.8 E 02 -

1.3 E O2 1.3 E 02 1.8 E O2 1.8 E 02 1.9 E 02 1.9 E 02 Kr-89 8.0 E 02 -

3.9 E 00 4.0 E 00 6.0 E 00 6.0 E 00 5.9 E 00 6.0 E 00 Xe-13tm 1.3 E 00 -

4.1 E-01 4.1 E-01 5.2 E-01 5.2 E 11 6.4 E-01 6.5 E-01 Xe-133m 6.8 E 01 -

2.3 E 01 2.3 E 01 2.9 E 01 2.9 E 01 3.5 E 01 3.5 E 01 Xe-133 3.0 E 03 -

9.6 E 02 9.7 E O2 1.2 E 03 1.2 E 03 1.5 E 03 1.5 E 03 Xe-135m 8.8 E 01 -

4.3 E 01 4.3 E 01 6.5 E 01 6.5 E 01 6.4 E 01 6.5 E 01 Xe-135 2.5 E 02 -

1.0 E 02 1.1 E 02 2.2 E 02 2.2 E 02 2.3 E 02 2.3 E 02 Xe-137 1.2 E 01 -

6.2 E 00 6.2 E 00 9.7 E 00 9.7 E 00 9.2 E 00 9.4 E 00 Xe-138 4.4 E 01 -

2.2 E 01 2.2 E 01 4.0 E 01 4.0 E 01 3.3 E 01 3.4 E 01 C"O

(:;:))

er pc ,

-; :)

• p;.m (AD

y. .

W ;P 3 pf.a===

7 c; #g"

m

,h "C 1 of 1 Amenrtment 19 12/12/75

SNESSAR-?1 TABLE 15.1.18-2 I'O? METERS USED FOR THE STEAM GENERATOR TUBE RUPTURE ACCIDENT ANALYSIS 1 Prior to the accident, core thermal power is 4,100 MWt (C-E , W-41) 3,876 MWt (BSW) , or 3,636 MWt (W-3S) . 19

2. Steam and power conversion system equilibrium liquid activity prio.r to the accident is based on primary-to-secondary leakage of 1,000 gal per day in the steam generators and is given in Table 15.1-2 F J41, W-3S, C-E) .

19 Prior to the accident the reactor was operating with a 1 gpm steam generator tube leak (B&W) .

3. Reactor coolunt activity is based on 1.0 percent tailed fuel as given in Section 11.1.

4. The primary-t o-second a ry leakage is evenly distributed among steam generators (W-41, W-3S, C-E). 19

5. Offsite power is lost, thu s making the condensers unavailable f or turbine bypass .

6. No steam jet air ejector release and no steam generator blowdown occur during the accident.

7. No noble gases are dissolved in the secondary side steam generator water.

8. Primary-to-secondary leakage continues in the nondef ective steam generator (s) .

9. During the postulated accident iodine carryover f rom the primary side in thc nordefective steam generator (s) is diluted in the incoming f eedwater.

10. The iodine partition factor in the steam generators before the accident is amount or iodine / unit mass of steam = 1.00 ESW amount of iodine / unit mass of water .01 C-E

.01 W-41

.01 W-3S and af ter the accident is = 1.00 BSW

.1 C-E

.1 W-41

.1 W-3S Qb 6 I Us 1 of 3 Amendment 19 12/12/75

SWESSAR-P1 TABLE 15.1.18-2 (CONT)

11. Atmospheric diffusion, including the volumetric building wake correction factor, is based on values given in Section 2.3.

12. No correction is made for depletion or the effluent plume of radioactive isotopes due to deposition on the ground or tor the radioactive decay in transit.

13. The following assumptions and/or data are taken from Section 15.4.3 of RESAR-41 and restated here for completeness of assumptions:

a. Thirty minutes after the accident the pressure between the defective steam generator and the reactor coolant system is equalized. The defective unit is isolated. No steam and associated fission product activities are released from the defective steam generator thereafter.

b. Steam release from the defective steam generator for the first 30 min is 30,000 lb.

c. Steam release from three nondetective steam generators for first 2 hr is 460,000 lb and the reaf ter (2-8 hr) is 1,300,000 lb.

d. Fee dwater flow to three nondefective steam generators for the first 2 hr is 461,000 lb and the reaf ter (2-8 hr) is 1,281,000 lb.

e. Rea ctor coolant released to the defective steam generator is 121,000 lb.

14. The following assumptions and/or data are taken from Section 15.1.17 of BSAR-205 provided by Babcock & Wilcox l19 and restated here for connleteness of assumptions:

a. Final isolation of the def ective steam generator is achieved at 25 min .

b. Reactor coolant leakage to the defective steam generator before isolation is 97,000 lb.

19

c. Steam vented to atmosphere from the defective steam generator is 252,000 lb.

d. Ste am venting time to the atmosphere from the defective steam generator is 19 minutes.

I'-

2 of 3 ,

O(U b endment 19

((1v 12/12/75

SWESSAR-P1 TABLE 15.1.18-2 (CONT)

e. Steam vented to atmosphere from the nondetective steam generator is 215,200 lb.

13

f. Steam venting tine to the atmosphere from the nondefective steem generator is 19 minutes.

15. The following assumptions and/or data are taken from Section 15.4.3 of CESSAR and restated here for completeness of assumptions:

a. The operator isolates the steam generator with the ruptured tube within 60 min. During the 60 min, 80,000 lb of reactor coolant is transported to the main steam system.

b. Cooldown is completed via the abnospheric steam dump valves.

c. At the end of approximately 4.0 hr following initiation of the transient reactor coolant system, heat removal may be transferred to the shutdown cooling system and the nondefective steam generator can be isolateG from the main steam system.

16. The tollowing assumptions and/or data are taken from Section 15.4.3 of RESAR-3S and restated here for completeness of assumptions:

a. Thirty minutes after the accident the pressure between the defective steam generator and the reactor coolant system is equalized. The defective unit is isolated. No steam and associated fission pro duct activities are released from the defective steam generator thereafter.

b. Steam release from the def ective steam generator for the first 30 min is 47,000 lb.

c. Steam release from three nondefective steam generators for first 2 hr is 454,000 lb and thereafter (2-8 hr) is 1,200,000 lb.

d. Feedwater flow to three nondefective steam generators for the first 2 hr is 441,000 lb and thereafter (2-8 hr) is 1,253,000 lb.

e. Rea ctor coolant released to the defective steam generator is 125,000 lb.

,G J

(J 3 of 3 ,y Amendment 19

^!

12/12/75

SWESSAR-P1 TABLE 15.1.20-1 ROD EJECTION ACCIDENT RELEASES Total Releases [Ci)

Isotope BSW C-E* W-41 W-3S I-131 2.7 E 03 2.1 E 03 6.7 E 03 5.9 E 03 I-132 1.2 E 02 7.6 E 01 2.2 E 02 2.0 r 02 I-133 1.1 E 03 8.2 E 02 2.5 E 03 2.2 E 03 I-134 1.2 E 02 5.2 E 01 1.6 E 02 1.4 E 02 1-135 4.2 E 02 3.0 E 02 9.1 E 02 8.1 E 02 Kr-83m 6.1 E 02 2.9 E 01 7.0 E 01 5.3 E 01 Kr-85m 1.0 E 03 1.6 E 02 3.4 E 02 2.8 E 02 Kr-85 1.5 E 03 2.0 E 02 3.6 E 02 9.6 E 02 Kr-87 5.9 E 02 9.5 E 01 2.6 E 02 1.9 E 02 Kr-88 1.8 E 03 2.8 E 02 6.4 E 02 5.0 E 02 Kr-89 4.4 E 01 6.3 E 00 1.8 E 04 7.4 E 01 p Xe-131m 6.9 E 02 9.4 E 01 1.3 E 02 1.1 E 02 Xe-133m 1.8 E 03 2.7 E 02 4.2 E 02 3.7 E 02 Xe-133 1.1 E 05 1.7 E 04 2.8 E 04 2.5 E 04 Xe-135m 7.2 E 03 2.6 E 03 7.0 E 03 6.2 E 03 Xe-135 1.3 E 04 1.8 E 03 2.5 E 03 2.2 E 03 Xe-137 7.1 E-01 1.2 E 01 2.4 E 02 1.0 E 02 Xe-138 3.0 E-02 4.8 E 01 3.0 E 02 1.6 E 02

• Containment release only Il0 1 of 1 qn Amendment 24 U 4/23/76 b\

SWESSAR-P1 TABLE 15.1.20-2 PARAMETERS USED FOR THE ROD EJECTION ACCIDENT ANALYSIS *

1. Prior to the accident, the core thermal power is 4,100 MWt (C-E , W-41) 3,876 MWt CBSW) , or 3,636 MWt (W-3S) .

2. The amount of activity accunulated in the fuel-clad gap is 10 percent of the iodines and 10 percent of the noble ,

gases, Table 11.1.1-1. 24

3. No credit is taken for activity decay prior to accident initiation.

4. The fraction of fuel rods in the core experiencing clad damage is taken from the NSSS Vendor's SAR and restated here for convenience.

BSW Section 15 .365 24 C-E Section 15 .07 W-41 Section 15 .1 W-3S Section 15 .1

5. The fraction of fuel rods in the core that melts is taken f rom the NSSS Vendor's SAR and restated here for convenience.

BSW Section 15 0.0 C-E Section 15 0.0 W-41 Section 15 .0025 24 W-3S Section 15 .0025

6. All the activity f rom the fuel pellet-clad gap in the failed fuel rods is available for release from the containment.

7. One-hundred percent of the noble gas activity and 25 percent of the iodine activity from the melted fuel is available for release from the containment.

8. One-hundred percent of the noble gas and iodine activity in the reactor coolant based on prior operation with 1.0 percent failed fuel is available for release from the containment.

9. No credit is taken for removal or iodine in the containment due to containment sprays.

10. The containment leaks for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design leak rate of 0.2 percent per day. Thereafter the containment leak rate is 0.1 percent per day.

b0 C??

Jvl 1 of 3 Amendment 24 4/23/76

SWESSAR-P1 TABLE 15.1.20-2 (CONT)

11. Fifty percent of the containment leakage is collected in the supplanentary leak collection and release system (Section 6.2.3.1) and filtered through high efficiency particulate air (EEPA) filters and charcoal adsorbers with an 'overall efficiency for iodine of 95 percent before release to the environment.

12. Steam and power conversion system equilibrium activity prior to the accident is based on primary-to-secondary leakage of 1,000 gpd in the steam generators (C-E , W-41, W-3S) . Prior to the accident the reactor was operating 19 with a 1 gpm steam generator tube leak (B&W). I

13. The primary-to-secondary leakage is even3y distributed among the steam generators.

14. Offsite power is lost, thus making the condensers unavailable for turbine bypass.

15. No steam jet air ejector release and no steam generator blowdown occur during the accident.

16. The model for the activity available for release to the atmosphere f rom relier valves assumes that the release consists of the activity in the steam and power conversion system prior to the accident plus that fraction of the activity leaking from the reactor coolant through the steam generator tubes following the accident. The leakat3 of reactor coolant to the secondary side of the t ..eam generator is assumed to continue at its initial rate, which is the same rate as the leakage prior to the accident, until the pressures in the primary and secondary are equalized. No mass transfer from the reactor coolant system to the steam and power conversion system is assumed thereafter.

17. No noble gas is dissolved in the steam generator water.

18. The iodir e partition f a ctor in the steam generators before the accident is:

1.00 B&W amount of iodine / unit mass of steam = 0.01 C-E amount of iodine / unit mass of water 0.01 W '41 0.01 W-3S and during the accident is: 19 1.0 BSW amount of iodine / unit mass of steam = 0.1 C-E amount of iodine / unit mass of water 0.1 W-41 0.1 W-3S i 670 06B 2 of 3 Amendment 19 12/12/75

SWESSAR-P1 TABLE 15.1.20-2 (COh'r)

19. Primary to secondary leakage termination time is taken from the NSSS Vendor's SAR and restated here for convenience.

BSW Section 15 180 sec l19 C-E Section 15 (later)

W-41 Section 15 2,400 sec W-3S Section 15 300 sec

20. Mass of steam discharged from the steam and power conversion system through the relief valves is taken from the NSSS Vendor 's SAR and restated here for convenience.

BSW Not applicable (later) l19 C-E Section 15 (later)

W-41 Section 15 76,000 lb W-3S Section 15 59,000 lb

21. Termination of relief valve steam discharge is taken from the NSSS Vendor's SAR and restated here for convenience.

Not applicable 19 ESW Section 15 ,

C-E Section 15 (later)

W-41 Section 15 500 sec W-3S Section 15 140 sec

• Based on assumptions outlined in Regulatory Guide 1.77.

3 of 3

$~Il Amendm[nt e

O -

19' U9 12/12/75

SW ESS AF -P 1 TAliLE 15.1.23-1 FUEL llANDLING ACCIDFNT RELEASES TO A'INO!'PffERE Released At?tivity (Ci)

• I sot one B6 W C-E W-41 W-3S F'uel Eb el Fuel Fuel Buildinq BuildinrT BuiIdtnq Building Kr-83m 4.4 -09 5.2 -05 1.6 02 1.8 -08 Kr-85m 4.1 -03 3.6 -01 1.6 03 4.8 -03 Kr-85 6.2 02 2.0 03 2.5 03 2.1 03 Kr-87 - - - -

Kt-88 2.7 -06 1.4 -03 7.4 02 1.5 -06 Xe-131m 3.2 03 1.0 04 4.1 03 1.4 04 Xe-133m 5.2 02 2.4 03 4.6 03 3,-

2.0 03 Xe-133 3.0 04 1.2 05 1.9 05 1.2 05 Xe-135m 8.8 -01 2.6 01 7.2 03 1.6 00 Xe-135 9.8 01 1.6 03 5.6 04 2.3 02 1-131 6.8 00 2.8 01 4.1 01 2.8 01 1-132 5.6 00 2.7 01 5.4 01 2.3 01 I-133 9.4 -01 7.2 00 5.1 01 3.1 00 I-134 - - - -

1-135 1.4 -03 4.1 -02 1.2 01 2.5 -03

• a" denotes <1.0 -10 O

sj C;

C:3

- g 1 of 1 Amend ment 35 c -) 10/6/77

SWESSAR-P1 TABLE 15.1.23-2 PARAMETERS USED FOR THE FUEL HANDLING ACCIDENT ANALYSIS *

1. Prior to the accident, the core thermal power is 4,100 MWt (C-E, W-41) , 3,876 MWt (B&W), or 3,636 MWt (W-3S) . The core fission product inventory at shutdown for this analysis is given in Table 11.1.1-1.

2. The accident is assumed to occur at the earliest time af ter shutdown that fuel handling cperations can begin. The post shutdown decay time is given in the NSSS Vendor's SAR and restated here for convenience.

B&W Section 15.1.20 (72 hr)

C-E Section 15.4.6 (72 hr)

W-41 Section 15.4.5 (20 hr)

W-3S Section 15.4.5 (100 hr)

3. The maximum fuel rod pressurization is $1,200 psig as given in the NSSS Vendor's SAR and restated here for convenience.

B&W Section (later)

C-E Section 15.4.6 W-41 Section 15.4.5 W-3S Section 15.4.5

4. The minimum depth of water between the top of the damaged fuel rods and the fuel pool surface is >23 feet, Fig. 1.2-B 35 and 1.2-3.

5. All the gap activity in the damaged rods is released. The gap activity as a fraction of the core inventory is presented in Table 11.1.1-1.

6. The number of fuel assemblies per core / fuel rods per fuel assembly is given in the NSSS Vendor's SAR and restated here for convenience.

BSW Section 15.1.20 (205/264) of which only the outer 64 rods of one assembly are damaged.

C-E Section 4.1, 15.4.6 (241/236)

W-41 Section 4.1 (193/264)

W-3S Section 4.3 (193/264)

7. The radial peaking factor is 1.65.

• (Based on Assumptions Outlined in Regulatory Guide 1.25) 1 of 2 Amendment 35 10/6/77 bl0 0,? j

SWESSAR-P1 TABLE 15.1.23-2 (CONT)

8. The iodine gap inventory is composed of inorganic species (99.75 percent) and organic species ( .25 percent) .

9. The overall decontamination factor for the fuel pool is 100.

35

10. Noble gases are not held up in the fuel pool.

11. The fuel building exhaust is collected by the supplementary leak collec_ ion and release system (Section 6.2.3.1) and filtered through high efficiency particulate air (IEPA) filters /charcocl adsorbers with an overall efficiency for iodine of 95 percent before release to the environment.

35

12. Atmospheric diffusion, including the volumetric building wake correction factor, is based on values given in Section 2.3.

13. No correction is made for depletion of the effluent plume of radioactive isotopes due to deposition on the ground or for the radioactive decay of iodine in transit.

b g j !j 2 of 2 Amendment 35 10/6/77

SWESSAR-P1 Table 15.1-23-3 is deleted. 35 1 of 1 Amendment 35 10/6/77 670 073

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FIG. 15.1.23 - 2 FUEL HANDLING ACCIDENT DOSE VS. DISTANCE PWR REFERENCE PL ANT SAFETY AN ALYSIS REPORT SW ESS AR- P I E

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F IG. 15. l.23 - 2 FLEL HANDLING ACCIDENT DOSE VS. DISTANCE PWR REFERENCE PLANT SAFETY ANALWIS REPORT SWESSAR-PI W-3S [70 l '] =i AMENDMENT 17 9/3CV75

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AMENDMENT 9 4/30/75

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SWESSAR-P1 15.2 BALANCE OF PLANT ASSUMPTIONS UTILIZED IN THE NSSS VENDOR ACCIDENT ANALYSES g The balance of plant assumptions utilized by the NSSS Vendor in Chapter 15 of his SAR are presented in Table 15.2-1.

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r, g 670 15.2-1 Amendment 21 2/20/76

r SWESSAR-P1 TABLE 15.2-1 BAIANCE OF PLANT ASSUMPTIONS USED IN WESTINGlk)USE ACCIDEPrr ANALYSIS RESAR-41 SWESSAR-P1 Asstumtion Reference Peterence

1. Reactor trips (underpower, Table 15.4-21 Section 7.8 underfrequency, and turbine) Section 15.3.4.1 Underpower limit of 701 Table 15.1-2 Underpower chang A to under-Underpower delay of 1.2 see Table 15.1-2 frequency by NSSS Vendor.

Section 8.4 Turbine delay of 1.0 see Table 15.1-2 Not saf ety related

2. Refueling water storage tank Table 15.4-21 Table 6.2.2-1 Table 15.4-22

3. Diesel generators and essergency Table 15.4-21 Section 8.3.1.1.3 21 power distribution equipment Table 15.4-22 Section 15.2.9.1

4. Essential service water system Table 15.4-21 MSSS systents are interf aced to ccunponent cooling water. Section 9.2.2.6

5. Essential corronent cooling Table 15.4-21 Section 9.2.2.6 water Table 15.4-21 Provide cooling water to RHR heat

6. Containment safeguards cooling exchangers, Section 9.2.2.6 equipment Provide containment air recirculation syst s , Section 9.4.5.1.6.

O 7. Feedwater system Section 10.4.7.6

a. Isolation Table 15.4-21 N Section 15.2.10.2 Section 15.1.14.3 a Section 15.2.13.1 Section 15.3.1 Section 15.4.2.1.1

-* Section 15.4.2.2 C3 Two isolation valves provided.

c) b. Control and isolation valve Section 15.4.2.1.2 closure time = 5 sec

c. Maximum step increase of Section 15.2.10.2 Valve will te sized to limit maximtun finw 250% in feedwater flow to to less than 350% of nominal.

any steam generator 1 of 4 Arrendment 21 W 2/20/76

SWESSAR-P1 TABLE 15.2-1 (CONT)

PESAR-41 SWFSSAR-P 1 Asstunpti on Reference Peterence

d. Miniatum f eedwater tenperature Section 15.2.10.2 SWESSAR-P1 design will satisfy this of 70 F a ssuntpt ion .

8. Auxiliary f eedwater systan Table 15.4-21 Section 10.4.10.6 Table 15.4-22 Section 15.3.1 Section 15.4.2.2 Section 15.4.3.2.2

a. Delay time 560 see Section 15.4.2.2 Table 10.4.10-1 Table 15.2-5 Table 15.3-4

b. cold water to unaffected Section 15.4.2.2 Table 10.4.10-1 steam generator at 240 sec Section 15.4.2.2

c. Auxiliary f eedwater ten.pera- Section 10.1.2 Table 10.4.10-1 ture 5120 F Table 15.2-5 Table 15.3-4

d. Flow rate 2450 gpsa Section 15.2.8.2 Table 10.4.10-1 Section 15.2-5 Section 15.3-4

e. Maxistas f eedwater flow of Section 15.4.2.1.2 Section 10.4.7.1 751,000 lb (0-2 hr) and Table 15.2-4 1,514,000 lb (2-8 12)

f. Minimura number of steam Section 15.2.8.2 Section 10.4.10.3 generators supplied = 1 Oection 15.2-5 Section 15.3-4 CN N g. Single f ailure can be either Section 15.4.2.2 Section 10.4.10.3 a the turbine driven pump or one of the three diesel generators.

9. Steam system Section 15.2.8.2 Section 10.3 CN Section 15.2.9.1 a Section 15.4.1.2 Table 15.4-21 Table 15.4-22 W 2 of 4 Amendnn nt 21 2/20/16

SWESSAR-P1 1ABLE 15.2-1 (CONT)

RESAR-41 SWESSAR-P1 Assumption Fef erence Feference

a. Isolatiou valves closure Section 15.4.2.1.1 Table 10.3-1 time = 10 see

b. Stop valves closure time Section 15.2.13.1 SWESSAR-P1 design will satisfy tiris

= 5 see Section 15.4.2.1.1 a ssurnption .

c. Blowdown limited to one Section 15.4.2.1.1 Section 10.3.3 steam generator with failure of one isolation valve

d. Maximum capacity of any Section 15.2.13.2 Table 10.3-1 g single steam dtnp, relief, or saiety valve is 292 lb/sec at 1,300 psia.

e. Turbine valve limited to Section 15.2.11.2.1 Maximum valve capacity is 1051.

1.2 times ncuminal steam flow at nominal pressure rigure 10.2-2, 10.2-3

10. Steam generator blowdown Table 15.4-21 Figure 10.4.8-1A autcusatic isolation valves provided

11. Sample reactor coolant system Table 15.4-21 Sc + ion 9.3.2

12. Sample steam generator shell Table 15.4-22 Section 9.3.2 side fluid

13. Class IE batteries Table 15.4-21 Section 8.3.2 Section 15.2.9.1

14. Control room ventilation Table 15.4-21 Sect ion 9.4.1.6 m

15. Control room protected Table 15.4-21 Section 6.4

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16. Emergency lighting Table 15.4-21 Sect ion 9.5. 3

17. Cbntainment pressure monitor Table 15.4-21 Section 7.3.3.9 CN N

W 3 of 4 Amendment 21 2// 0/7t,

SWESSAR-P1 TABLE 15.2-1 (CONT)

RESAR-41 SWESSAR-P1 Assumption Feferen& Feference

18. Maintain connection of reactor Section 15.2.5.1 Section 8.3.1.1.1 coolant pumps to generator if Section 15.3.4.1 no electrical f aults for 30 sec.

19. Saf ety injection syntesa 21

a. Delay time 525 see Section 15.4.2.1.2 Table 8.4-1 Table 15.3-4 Table 15.4-23

b. Safety injection water Table 15.3-4 Table 6.1-2 temperature 5120 F Table 15.4-23

c. Accumulator water tempera- Table 15.3-4 Table 6.1-2 ture 5120 F Table 15.4-23 Later

d. Accumulator line resistance Table 15.3-4 250 5 L/D 5 500 Table 15.4-23

20. Contairmnent pressure Table 15.4-23 Section 6.2.1.1.2 52.1 psia f or blowdown Section 6.3.3 47.9 psia for reflood O

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